ML20010G486
| ML20010G486 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 09/18/1981 |
| From: | ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | |
| Shared Package | |
| ML20010G423 | List: |
| References | |
| NUDOCS 8109210251 | |
| Download: ML20010G486 (400) | |
Text
{{#Wiki_filter:.. -. __ . = . . - - .- ._- PAGE REPLACEMENT GUIDE FOR ! AMENDMENT 61 ! CLINCH RIVER BREEDER REACTOR PLANT
- PRELIMINARY SAFETY ANALYSIS REPORT (DOCKET NO. 50-537)
Transmitted herein is Amendment 61 to Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537. Amendment 61 consists of new and replacement pages for the PSAR text and Question / Response supplement pages. Vertical lines on the right hand side of the page are used to identify questien resoonse information and lines on the left hand side f) d are used to identify new cr changed design infonnation. The following attached sheets list Amend.nent 61 pages and instructions for their incorporation into the Preliminary Safety Analysis Report. I i 8109210251 810918 DR ADOCK 050 y
AMENDMENT 61 PAGE REPLACEMENT GUIDE INSERT THESE PAGES REMOVE THESE PAGES Chapter 1 1.1-21, 22 1.1-21, 22 1.2-7, 8 1.2-7, 8 1.4-6a, 7, 8 1.4-6a, 7, 8, 8a 1.4-30 thru 33 1.4-30 thru 33 1.5-26, 27 1.5-26, 26a, 27 Chapter 3 3-ix, x 3-ix, x 3-xi, xii 3-xi, xii 3-xviii, xix 3-xviii, xix 3.1-5 thru 8 3.1-5 thru 8 3.1-11, 12 3.1-11, 12 3.1-13, 14 3.1-13, 14 - 3.1-17, 18 3.1-17. 18 3.1-33, 34 3.1-33, 34 3.1-69 thru 76 3.1-69 thru 76 3.1-79, 80 3.1-79, 80 3.1-83 thru 86 3.1-83 thru 86 3.2-8, 9, 10, 10a, 10b, 11 3.2-8, 9, 10, 10a, 10b, 11, 11a thru 11e, 12 1 3.2-15a, 15b 3.2-15a, 15b 3.4-1, la, 2, 2a, 3 thru 6 3.4-1, la, 2, 2a, 3 thru 6 3.5-5, 6 3.5-5, 6 3.5-9, 10 3.5-9, 10 3.5-13, 13a 3.5-13, 13a 3.6-1, la 3.6-1, la 3.6-9, 9a 3.6-9, 9a 3.7-3, 3a, 3b, 3c 3.7-3, 3a, 3b, 3c 3.7-5, 6, 6a 3.7-5, 6, 6a 3.7-10a, 10b 3.7-10a, 10b 3.7-19, 20 327-19, 20 3.7-23, 24 3.7-23, 24 3.7-45, 46 3.7-45, 46 3.7-A.11a, lib 3.7-A.11a, lib 3.7-A. 16b, 16c, 16d, 3.7-A. 16b, 16c, 16d, 17, 18, 19 17, 18, 19 3.7-A-C1, Cl-a 3.7-A-C1, Cl-a 3.7-A-CS, C5-a 3.7-A-C5, C5-a 3.8-1, la 3.8-1, la 3.8-2a, 2b, 3, 3a 3.8-2a, 2b, 3, 3a 3.8-5, 6, 7, 7a 3.8-5, 6, 7, 7a O A
r REMOVE THESE PAGES INSERT THESE PAGES Chapter 3 (Cont'd.) 3.8-7d, 8, 9, 9a 3.8-7d, 8, 9, 9a 3.8-9b, 9c 3.8-10, 10a, 11, lla, 3.8-10, 10a, 11, lla, 12 thru 15, 15a 12 thru 15, 15a 3.8-18a, 19 3.8-18a, 19 3.8-22, 22a, 23, 23a, 24, 24a, 3.8-22, 22a, 23, 23a, 24, 24a, 25, 25a, 26, 26a, 27, 28, 28a, 25, 25a, 26, 26a, 27, 28, 28a, 29 thru 32, 32a, 33 thru 37, 29, 30, 31, 31a, 32, 32a, 37a 33 thru 37, 37a 3.8-39, 40 3.8-39, 40 3.8-47b, 48, 49 3.8-47b, 48, 49 3.8-52 thru 56 3.8-52 thru 56 3.8A-iii 3.8A-iii 3.8A-11, 12 3.8A-11, 12 3.10-1 thru 5 3.10-1 thru 4 3.11-1, 2 3.11-1, 2 3A.2-3 3A.2-3 3A.4-1 thru 8 3A.4-1 thru 8 3A.5-1 thru 4 3A.5-1 thru 4 3A.6-1 thru 5 3A.6-1 thru 5 Chapter 4 () 4.2-174, 175 4.2-214, 215 4.2-174, 175 4.2-214, 215 4.2-222 thru 225 4.2-222 thru 225 4.2-238,-239 4.2-238, 239 4.2-368 4.2-368 4.2-534, 535 4.2-534, 535 4.3-9, 10 4.3-9, 10 4.3-144, 145 4.3-144, 145 Chapter 5 5.1-17d, 17e 5.1-17d, 17e 5.2-4a, 4b 5.2-4a, 4b 5.2-6, 6a 5.2-6, 6a 5.2-12, 12a 5.2-12, 12a 5.4-10, 11 5.4-10, 11 5.5-48, 49 5.5-48, 49 5.5-52, 53 5.5-52, 53 Chapter 6
-6.1-2 6.1-2 '
6.2-10a, 11 6.2-10a, 11 6.2-14b, 14c 6.2-14b, 14c 6.2-27, 27a 6.2-27, 27a 6.2-27d thru 279 6.2-27d thru 27g s. 6.2-37, 38 6.2-37, 38 B
c REMOVE THESE PAGES INSERT THESE PAGES Chapter 7 7-v, 7-va 7-v, 7-va 7-1-3, 4 7.1-3, 4 7.4-2a 7.4-2a 7.4-3, 4 7.4-3, 4 7.5-29, 30, 31 7.5-29, 30, 31 Chapter 9 9.1-42, 43 9.1-42, 43 9.1-64, 65 9.1-64, 65 9.1-92 9.1-92 9.1-94, 95 9.1-94, 95 9.5-5, 6 9.5-5, 6 9.5-8, 8a 9.5-8, 8a 9.5-19 thru 28 9.5-19 thru 28 9.5-30 thru 41, 41a, 42 9.5-30 thru 41, 41at 42 9.14-1 thru 10 9.14-1 thru 10 9.15-1, 2 9.15-1, 2 9.16-1, 2 9.16-1, 2 Chapter 11 11.3-43, 44, 44a, 45, 11.3-43, 44, 44t, 45, 0 45a, 46, 47 Chapter 13 45a, 46, 47 13-1, ii, iii, iiia, 'v, 13-1 thru 13-ix v, vi, vii 13.1-1 thru 21 13.1-1 thru 24 13.2-1 thru 7 13.2-1 thru 7 13.4-1 13.4-1 13.5-1 thru 5 13.5-1 thru 5 13.6-1 13.6-1 13.7-1 thru 10, 10a, 10b, 13.7-1 thru 13 11, 12, 13 Chapter 14 14-i, 11 14-1, ii 14.1-1 thru 7, 7a thru 7q, 8 14.1-1 thru 23 14.2-1 14.2-1 Chapter 15 T 15.1-50, 50a, 50b, 50c, 51, 52, 15.1-50 thru 54 53, 53a, 54, 54a, 55, 56, 57, 57a, 57b, 57c 15.1-58 thru 64, L4a, 64b thru 64e O. 15.1-65 thru 77 - 15.1-79 thru 89, 89a, 89b, 89c - 15.1-90, 91, 92, 92a thru 921 C
REMOVE THESE PAGES INSERT THESE PAGES Chapter 15 (Cont'd.) 15.2-1, 2 15.2-1, 2 15.2-3, 4 15.2-3, 4 15.2-43, 44 15.2-43, 44 15.2-77, 78 15.2-77, 78 15.3-1, 2 15.3-1, 2 K,3-7a 15.3-7a 16.7-20d, 20e 15.7-20d, 20e Chapter 16 16.6-1 thru 13 16.6-1 thru 18 i O 1 l l l i i. i I I l !O f D i i !
. ~ . , . . _ . . . _ _ _ _ _ . . _ . _ . . _ _ _ . . _ . _ _ _ - _ _ _ _ -
l AMENDMENT 61 QUESTION / RESPONSE SUPPLEMENT This Question / Response Supplement contair.s an Amendnent 61 tab sheet to be inserted following Qi page, Amensnent 60, February 1981. Page Qi Amendment 61, is to follow the Amendment 61 tab. Replacement pages for the Question / Response Supplement are listed below: REPLACEMENT PAGES . REMOVE THESE PAGES INSERT THESE PAGES Q001.268-1 Q001.268-1 Q001.451-1, 2 Q001.451-1, 2 0001.541-1 0001.541-1 Q001.550-1, 2 Q001.550-1.2 0001.554-1 Q001.554-1 Q001.555-1 Q001.555-1 0001.608-1 Q001.608 I Q001.610-1 0001.610-1 0001.612-1 0001.612-1 Q001.690-1, 2 0001.690-1, 2 O Q011.25-3, 4 Q421.1-1,2 Q011.25-3,4 Q421.1-1, 2
- Q421.2-1, 2 Q421.2-1, 2 Q421.6-1 Q421.6-1 Q421.10-1 Q421.10-1 O
E
- _ - . _ . _ . . _ . __ - . - _ . _ _ . - _ ~ . . _ . _ _ . _ _ _ . . , _ _ . . _
V TABLE I (Continued) 7m ( _-
)
3 Discussed Further in No. Title Rev. PSARSection(s) 1.93 Availability of Electric Power 0 . 16.3.9 Sources (12/74) 1.94 Quality Assurance Requirements for - 17.1, Question 411.2 Installation, Inspection, and Testing of Structural Concrete and Structural Steel during the Construction Phase of Nuclear Power Pltnts (4/75) 1.95 Protection of Nuclear Power Plant - NA: Chlorine will not be used Control Room Operators Against an for treatment of circulating Accidental Chlorine Release (2/75) water and will therefore not be stored on site. Possible Off-site Release will be treated in Reg. Guide 1.78. 1.96 Design of Main Steam Isolation Valve - NA: This Guide is intended Leakage Control Systems for Boiling for BWR's only. Water Reactor Nuclear Power Plants (5/75) j f} NJ Amend. 44 l.1-21 April 1978
( TABLE I ((,ontinued) NOTES: 61l 1. (Deleted) 61l 2. This Guide is applicable to the CRBRP. Actual use of this Guide, however, is expected to be very limited, if any due to its Iimited scope. 61l 3. This Gulde is applicable to the CRBRP, as appropriate. 61l 4 This Golde is cone.idered applicabio to ASME Section Ill, Class 1 components. The provisions of this Guide are considered appiIcable to the CRBRP with the following exception: Regulatory Position C.2 requires that the pre-heat temperature for production welds be maintained until a postweld heat treatment has been performed. This will be complied with whenever practicable or required by ROT E15-2NB-T, unless the need and acceptability of an alternate procedure has been demonstrated. 61l 5. This Guide developed and intended primarily for application to tubular products used for ASE-lil Code Class I components on LWRs. The corresponding CRBRP components will be of austenitic steel. The state-of-the-art of the UT examination, as specified by the Guide, has not been capable of producing meaningful results. The CRBRP, however, is anticipated to meet the requirements as set forth in NB-2550 of ASME-Ill for the examination addressed by the Guide. 61l 6. The CRBRP PSAR was prepared in accordance wIth Regulatory Guide 1.70 (LMFBR Edition, 2/74) where practicable. 61 l 44 7. This Guide is considered generally applicable to CRBRP. l l l 1 t 1 O
.1-22 Amend. 61 Sept. 1981
the secondary shutdwn system is arranged using a general coincidence a logic. These logics are described in Section 7.2.1. Primary and secondary systems are electrically and mechanically isolated. Sufficient redundancy is included within each system to assure that single random failures will not degrade protection by either system. 1.2.7 Auxiliary Systems The Auxiliary Liquid Metal System provides the facilities for receipt, storage and purification of all liquid metal used in the CRBRP. It also provides the capability for controlling reactor sodium level variations, accommodates primary sodium volumetric changes, provides cooling for the core components stored in the Ex-Vessel Storage Tank (EVST),and by means of the Direct Heat Removal Service (DHRS) gives a means of long term reactor decay heat removal that is independent of the 26 intermediate heat transport system and steam generator system loops. The Compressed Gas System processes ambient air to provide compressed dry air for pneumatic instruments, maintenance systems, unloading devices, tooling, and miscellaneous cleaning and inspection services. This system provides for sodium removal systems and as required for plant usage. The Recirculating Gas Cooling System provides cooling service to cells and equipment located in the Reactor Containment Building and the Reactor Service Building. p 15 The Chilled Water Systems provide heat removal v capability from certain equipment and areas in the Reactor Containment Building and the Reactor Service Building. The Inert Gas Receiving and Processing System (IGRPS) provides inert gases as required by other systems of the CRERP, including cover gas, cell inerting atmosphere, valve actuation gas in inerted cells, cooling gas, gas for certain seals, for component 59l cleaning and other services, and vacuum for out-gassing and purposes. In addition, the IGRP System provides for the control of reactor cover gas radioactivity and for the processing of gases to be released from the system to remove their contained radioactivity. The Impurity Monitoring and Analysis System provides for the sampling, monitoring, and analysis of the soaium, NaK, and argon cover gas systems in the plant, and acceptance sampling and analysis of incoming 56 sodium, NaK, argon, and nitrogen. The Treated Water System includes the domestic ( system,theclosedcoolingwatersystem, water (makeup)potablo) water treatment system and the cooling water makeup system. The River Water Service System handles and treats river water for the plant. The system includes the river water pumps and piping, intake filtration equipment and the plant service water system. O Amend. 59 1.2-7 Dec. 1980
-m.. . _ , ,_,, -
, - --- .- m.
The Heat Rejection System provides the heat sink using the main cooling tower for weste heat loads from the turbino condensers, and from the various plant 4l auxillary and service systems such as sodium pump oil coolers, air conditions, 44 air cc.npressors, pump coolers and the turbine oil coolers. The Emergency Plant g Service Water System emergency cooling tower structure provides the heat sink , for the safety related components listed in Table 9.9-3. Details of the 33 auxillary system are given in Chapter 9. 1.2.6 Refuelina Svstem The Reactor Core is designed to be refueled annually. Under equilibrium conditions, all fusi and inner blanket assemblies are replaced as a batch every i two years, with a planned mid-term Interchange of 6 inner blanket assemblies 4120 1 for 6 fresh fuel assemblles designed to add suf'Iclent excess reactivity to the system to completa the (550 fpd) burnup. The rad al blanket assemblies in the 59 first and second rows are replaced as a batch at 4 and 5 year intervals, 51 respectively. The in-Vessel Handling Subsyctem (IVHS) provides for the transfer of core 44 assembiles In the reactor vessel, between their normal positions in the reactor 41l core and the storage positions outside the core accessib!e by the Ex-Vessel Transfer Machine. The major equipment comprising the IVHS are the in-Vessel Transfer Machine (IVTM), Auxiliary Handling Machine (AHM), AHM Floor Valves (FV), IVTM Port Adaptors, and associated maintenance and storage facilities and equipment. The IVTN is Installed in the small rotating plug In the reactor 44 l head af ter reactor shutdown. The machine raises or lowers core assemblies by means of a grapple. Translation to a new position is by rotation of the 4Il 44 l: eactor head rotatable plugs. The AHM Is used to Install and remove the control rod crivellnes, port plugs, and in-vessel section of the IVTM in the reactor. The port adaptors and floor valves provide a means for closure of the 61] reactor and storage ports during the transfer of refueling equipment in Sg preparation for refueling operailon. The Ex-Vessel Handling Subsystem (EVHS) provides for the transfer of core assemblles between the reactor, the Ex-Vessel Storage Tank (EVST), and tne Fuel 41 44 Handling Ccil (FHC) located in the Reactor Service Building (RSB). The system consists of the Ex-Vessel Transfor Machine (EVTM) mounted on a Gantry-Trolley 441 (G-T), EVTM Floor Valves (FV), Core Component Pots (CCP), port plugs and adaptors, and associated maintenance and storage equipment and facilItles. 59 The Ex-Vessel Storage Subsystem (EVSS) consists of the Ex-Vessel Storage Tank (EVST). The EVST is a sodium-filled tank used to store and cool spent fuel 4d prior to shipment of fsite, and preheat new core assemblies. The capacity of 4 11 the EVST is about 650 assemblies. 41l 44
'20 ,
O 1.2-8 Amend. 61 Sept. 1981
bursement of utility funds and exercising the various contractual rights ! designed to protect the utilities' interests, including approving any l proposed changes in Project scope or deviation from the approved Reference Design or specifications, maintaining access to information and data, either in the possession of the Government or any of the Project contractors, seeing that the conditions for the disbursement of utility funds are met, and exercising the rights of termination of the Project in the' event a contractually based tennination occasion arises. 1.4.2.3 DOE ORGANIZATION 45 The overall DOE organization is shown in Figure 1.4-2. Prime responsibility for the CRBRP Project is assigned to the Director, CRBRP Project. The line of authority is from the Secretary of Energy to the Assistant Secretary for Nuclear Energy to the Deputy Assistant Secretary 61 for Nuclear Reactor Programs, and then to the Project Director. The Director for Reactor Research and Technology, in consultation with the Project Director, manages the Base Technology program which 45 contributes support to the CRBRP. 25 1.4.2.4 TVA ORGANIZATION The organization of TVA is sho.vn in Figure 1.4-4. The responsibi-lity for TVA's activities will be met by or throt.gh the Office of Power, shown in Figure 1.4-3. The staff and 6ivisions that will carry out, support, or have the potential to support TVA's role as operator are discussed in the following paragraphs: 1.4-66 Amend. 61 O Sept. 1981
1.4.2.4.1 TFICE OF POWER (Floure 1.4-3) l25 We Division of Energy Denonstrations and Technology is Ox responsible for research activities associated with the broad field of electric power supply for the Office of Power. We Division can provide assistance in such areas as environmental, nuclear, and computer technology. It can also provide assistance in non-nuclear energy research and technology. We Division of Energy Demonstrat. ions and Technology has been involved in INBR technology since the carly phases of the Project 61 Definition Phase of the Project. The Nuclear Systes Group, a part of the Division of Energy Demonstrations and Technology, assists the Manager of Power in coordinating WA's Project activities and provides the interface of WA with all Project carticipants for the CPSRP. De Group will coordinate ancVor provide WA's activities and the technical direction for WA's contractual 61 responsibilities. Division of Macipar Power
. The Division of Nuclear Power has the responsibility for the operation and maintenance of all WA nuclear electric generating plants and will have this responsibility for the CRBRP. Additional information about the responsibilities of the Division of Nuclear Power for the CRBRP is 61 included in Section 13.1.
Di'rision of Powr System Ooerations This division provides the services of its central electrical, instrumentation, and chmical laboratory and technical staff. In addition, field test engineers are provided for chmical and laboratory tests and for solution of special electrical engineering and chmical probles. Engineers and technicians from this division are responsible for the maintenance and testing of the relaying associated with the transmission systs and the inter-WA communications system. Division of Transminnion Plannina and Enaineerina Advice and assistance regarding the engineering and design of the electrical tranmission lines, sttstations, and cmmunication facilities are available from the Division of Transmission Planning and Engineering. 61 Office of Power k ality Ananrance and And_it Staff
%e Quality Assurance and Audit Staff of the Office of Power ascures that periodic audits of the activities of all nuclear plants owned ancVor operated by WA are performed and that effective operational quality assurance progrms are developed and inplemented.
Amend. 61 s 1.4-7 Sept. 1981
1.4.2.4.2 BMRCE OF 'lVA ORGANIZATION (Ficure 1.4-4) 1.4.2.4.2.1 DIVISIONS /TFICE/ STAFF OUTSIDE 'IEE &FICE OF PCWER '1 EAT PROVIDE A DIRECT WRVICEr Division of Occuoational Health & Safety
%e Division of Occupational Health & Safety is responsible for furnishing special services in the ,_ield of industrial safety and hygiene, environmental hygiene, radiological protection, pollution control, and other related areas. The Radiological Hygiene Branch within the Division of Occupational Health and Safety will provide administrative 25 supervision for the Health Physics Unit at the CRBRP. The Radiological Hygiene Branch is responsible for the preparation and review of radiological protection standards and for establishing and conducting all phases of the offsite radiological monitoring progrm. We Inclistrial Hygiene Branch within the Division of Occupational Health and Safety will perform independent industrial safety and hygiene hazard control reviews l25 and audits at the nuclear plants, including the CRBRP. '1hese reviews and audits are conducted to verify that the plant is operated to assure control of industrial safety and hygiene hazards and to assure ccanpliance with the
'lVA Hazard Control Plan. It also appraises and reccx:rnends corrective action on potentially harmful factors in the working environment other than radiation, all in relaticn to identified and approved standards as given in 61 the 'IVA Hazard Control Manual.
Division of Medical Services The Division of Medical Services is responsible for 'lVA's l overall health program. This will include mployee health services for the l CRBRP. 1 Division of ProDerties g Services The Division of Properties and Services will share industrial-radiological security responsibilities for the CRBRP with the l Power Security Section of the Management Services staff and the Division of Nuclear Power in the Office of h er. The functional relations between these three groups and how they share industrial-radiological security responsibilties are discussed in Section 13.7 under Radiological Security 61 Progrm. Office of Narnral Resources This office through its Air Resources Program, Data Services Branch, Laboratory Branch, Water Quality Branch, Fisheries and Ecology Branch, and Water Systes Developnent Branch provides environmental technical guidance, assistance, and services as needed to assure activities 61 are in ccxnpliance with Federal Environmental Regulations and Legislation. Amend. 61 O Sept. 1981
]
Nucinar Rafety Review Staff O The Nuclear Safety Review Staff is a top-management level group which acts independently of WA organizations concerned with the design, construction, operation, and support of nuclear plants to mulitor, review, and audit WA's nuclear activities and advise the Board on nuclear safety policy. 1.4.2.4.2.2 h Oroanizarinns In addition to the organizations listed in Section 1.4.2.4.2.1, any other WA Organization is available to provide service for i 61 the CRBRP. O l l O V Amend. 61 1.4-8a Sept. 1981
- O O O l
l l CLINCH RIVER BREEDER REACTOR PLANT PROJECT OFFICE i 0FFICE OF THE j DIRECTOR l INFORMATION DIVISION h OL DIVISION M NAGE DIVISION T 8 VI 10
~
i 1 1 l } b AUDIT ADMINISTRATIVE AUTOMATIC DATA QUALITY LABOR SITE j O c DIVISION SERVICES PROCESSING ASSURANCE RELATIONS REPRESENTATIVE ! DIVISION DIVISION DIVISION DIVISION t
- I
- CONSTRUCTION ENGINEERING PROCUREMENT PUBLIC SAFETY OPERATIONS j DIVISION DIVISION DIVISION DIVISION DIVISION t
i EF I EE l 1 I $8 i 3-77-P09217-2 'l , Figure 1.4-1 CPSRP Project Office Organization I
.i 4
i 3
SECRETARY OF ENERGY ASSISTANT SECRETARY FOR NUCLEAR ENERGY I __ 0FFICE OF THE DEPUTY ASSISTANT SECRETARY FOR NUCLEAR REACTOR PROGRAMS OFFICE OF 0FFICE OF SAFETY, QUALITY ASSURANCE PLANS AND RESOURCE AND SAFEGUARDS liANAGEMENT F ? S! i 1 0FFICE OF ADVANCED NUCLEAR OFFICE OF LIGHT OFFICE OF REACTOR SYSTEFS AND PROJECTS WATER REACTORS RESEARCH AND TECHNOLOGY FAST FLUX TEST FACILITY CLINCH RIVER BREEDER REACTOR PROJECT OFFICE PROJECT OFFICE WE EB
", f' , FIGURE 1.4-2 ORGANIZATION OF DEPARTMENT OF ENERGY (DOE) 8D 0 0 0 ;
i TENNES Manager of Deputy Ma Power ops of Powi Ass't M'gr Poder ops
.j Division of Division of Division of Division of 1 PWR Sys ops Fossil & Hydro Nuclear Power Energy Cons & RT J
Nuclear Plants Including CRBRP Asst. Mi of Poi Distri< 1 ( w
'l iEE VALLEY AUTHOR!TY Manager of Power l
hager Ass't Manager Ass't M'gr Manager of 1 of Power of Power Power Engin'g fr Finance, Regulatory, Budget, Control, Personnel Power Planning Safety Review, Staff Quality Assurance and Audit l )ivision of Division of TR Division of Division of Planning & Power Fuels Power Constr. Engineering Jtilization i inager Division of Energy (er Demonstrations & Tech. _ FIGURE 1.4-3 Office of Power
- ts Organization (TVA)
Amend. 61 1.4-32 Sept. 1981 J
P TENN] b Offici Information I Equal Emp' Office of Office of 01 General Council Agricuitural and Chemical Development Divi Divisions: I Agricultural Development Office of Chemical Development Chemical Operations Natural Resources Division / Staff Land and Forest Resources l Water Resources Office of Powe Natural Resources Services Divisions: Environmental Quality Staff Land Between the Lakes Power Utilization Power System Operation 2 Fossil and Hydro Power Nuclear Power Energy Conservation and Power Construction Transmission Planning E l Energy Demonstrations c Fuels 1 9 1
i iSEE VALLEY AUTHORITY
)ard of Directors e of Ger.eral Manager Budget Staff Nuclear Safety iffice Washington Office e ew aff oyment Opportunity Staff I 'fice of Conmunity Office of Engineering Design Office of Development and Construction Management Services sions: Divisions: Divisions / Staff / Office @onnerce Engineering Design Labor Relations Staff @ommunity Services Construction Management Systems Personnel l
Finance Purchasing Property and Services Office of Health and Safety r Divisions: l Medical Services Occupational Health l and Safety iRates td Engineering Rd Technology l l FIGURE 1.4-4 Organization of TVA 1.4-33 Amend. 61 Sept. 1981 e 1 l
h 41 l 1.5.1.3.5 Fallback Position 57 l In the event that operating with f ailed radial blanket cannot be shown to be satisfactory from a public safety viewpoint, the reactor may be required to shutdown when the blanket material is exposed to the sodium. 1.5.1.4 Sodlum.-Water Reaction Pressure Rellef Test 1.5.1.4.1 Purnose The prine' pal concern associated with the large water to sodlum leak in steam generators is potential system damage, principally to tho lHX by propagation of transient pressure waves through the Intermediate Heat Transport System (IHTS). The objective of the Sodium Water Reaction Pressure Rellef Subsystem (SWRPRS) is to relieve pressures from the IHTS and thereby protect the primary coolant boundary from damage in the region of the primary sodium to intermediate sodium heat transfer interf ace. 61 l The approach to design of CRBRP SWRPRS is to assume a conservative design basis 59 water to sodium leak and to use a validated calculational method to predict pressure loads on the IHX. It is a design requirement that the lHX be able to withstand the sodium-water reaction pressures without compromising the primary coolant boundary. A survey of available existing analytical methods was completed to select the best method f or improvements consistent with CRBRP requirements. Tne TRANSWRAP computer program (Ref. 5) was selected for use in the CRBRP analysis. An O improved version of this code was used to establish loads on the IHX for the ' reference design fHTS piping and component arrangement and the reference design SWRPRS. A design basis leak was assumed to consist of an Equivalent Double-61 Ended Guillotine (EDEG) failure of one steam generator tube followed by the equivalent of two additiont.! EDEG tube f ailures. The two additional failures occur as follows: one EDEG failure occurs one second after the initial EDEG failure. one additional EDEG failure occurs two seconds after initial EDEG failure This sequence is superimposed on a system which has been pressurized by an undetected moderate-sized leak to just below the rupture disk burst pressure. 59 The three tube DEG failure is not Intended to represent a realistic event, but , rather it provides a basis for calculating conservatively large pressure loads for the desig1 of IHX and the pressure relief system. Results of analyses using this basis are repor+ed in Section 5.5.3.6. To increase confidence in assuring integrity of the primary coolant boundary even during a large sodium-water reaction, the development program will provide technical Information which is not available for inclusion in the PSAR. The safety related objectives of the development program are: 1.5-26 Amend. 61 Sept. 1981 _ . - ~ _. _
a) to validate the computer program used to predict pressures in the IHX dering a postulated sodium water reaction, and b) to confirm that effects of the design basis leak assumed for determining pressures in the IHX are conservative. O l l l l \ O 1.5-26a Amend. 61 Sept. 1981. l
- m
Program 41 l 1.5.1.4.2 As part of the Steam Generator Development Program, AI (V) has construt.+.ed the Large Leak Test Rig (LLTR). The test programs 41 included pulling apart a notched tubular specimen in the sodium filled test article to simulate a DEG failure. A steam / water mixture was forced through 44 i 41l the burst tube into the sodium. For most tests, surrounding tubes contained stagnant, pressurized steam / water mixtures. In general, the development 441 effort provided technical information regarding the design of pressure relief systems to handle unexpected water-to-sodium leaks. Measured values of pressure at various locations in the test rig are being compared with calculated pressures obtained using the modified TRANSWRAP computer program to analyze the test rig and test article. It now appears that the computer code predicts values of pressure that are either in agreement with measurements or are conservatively large relative to measured pressures for the test rig and test article. Thus, it appears that the analysis of CRBRP for 41l sodium-water reaction pressures using this code are being conservatively 44 accomplished. This conclusion is still under review and evaluation and there-fore subject to adjustment as the remaining test. data are examined. Examination of the test article following intentional" bursting of 44 l a single tube gives some indication of the nature and extent of damage propagation to other tubes. It is expected that the tests will demonstrate 47 that the calculated loadings from sodium-water reactions are conservative. C')' 1. 5.1.4. 3 Schedule
- 4) l 44,41 CY 73 74 75 76 77 78 79 80 81 LLTR-Module Steam l Generator (MSG) test 47l 4j l data available A
Modified TRANSWRAP validated by test 41 l results g i Extent of damage in MSG evaluated l 3 44 4jl kJ Amend. 47 Nov. 1978 1.5,27
TABLE OF CONTENTS (Continued) Page No. 3.8.3.3.5 Sodium Fire Load 3.8-14 3.8.3.3.6 Hot Sodium Spill Effect 3.8-14 2.8.3.3.7 A:cident Temperature Load 3.8-14 3.8.3.3.8 Negative Pressure on the Linere 3.8-14 3.8.3.3.9 Hot Spots 3.8-15 3.8.3.3.10 Loading Combinations 3.8-lSa 49 3.8.3.3.10.1 Load Combinations for Concrete Structures 3.3-15a 3.8.3.3.10.2 Loading Combinations for Steel Structures 3.8-16 3.8.3.4 Design and Analysis Procedures 3.8-18 3.8.3.4.I General Analysis Procedures 3.8-18 3.8.3.4.2 Analysis for Seismic Loads 3.8 18a 3.8.3.4.3 Analysis for Openings 3.8-18 a 34 3.8.3.4.4 Liner Analysis 3.8-18a 3.8.3.4.5 Radiation Generated Heat Effect 3.8-19 3.8.3.4.6 Reinforcement Design 3.8-19 3.8.3.4.7 Structural Steel Design 3.8-19 3.8.3.5 Structural Acceptance Criteria 3.8-20 3.8.3.5.1 Stress 3.8-20 3.8.3.5.2 Strain 3.8-20 3.8.3.5.3 Gross Deformation 3.8-20 3.8.3.5.4 Factor of Safety 3.8-20 3.8.3.5.5 Shear Response 3.8-20 3.8.3.6 Materials, Quality Control and Special Construction 3.8-21 Techniques (3 V 3 - ix Amend. 49 Apr. 1979
TABLE OF CONTENTS (Continued) 3.8.3.6.1 Materials 3.d-21 3.8.3.6.1.1 Concrete 3.8-21 ! 3.8.3.6.1.2 Reinforcing Steel 3.8-21 3.8.3.6.1.3 Liner 3.8-21 i 3.8.3.6.1.4 Structural Stol 3.8-21 l l 3.8.3.6.2 Construction Techniques 3.8-21 l l 3.8.3.6.3 Quality Control 3.8-21 l 3.8.3.6.4 Testing Requirements for Reinforcing Steel 3.8-21 3.8.3.6.4.1 Mil! Tests 3.8-21 61 l 3.8.3.6.4.2 Owner's Surveillance 3.8-22 39l l 3.8.3.7 Testing and in-Service Survelllance Requirements 3.8-22 3.8.4 Other Seismic Category Structures 3.8-22a l 1 3.8.4.1 Description of the Structures 3.8-23 3.8.4.1.1 Reactor Servico Building 3.8-23 3.8.4.1.2 Control Building 3.8-24 3.8.4.1.3 Steam Generator Buildings 3.8-24a 49 3.8.4.1,4 Diesel Generator Building 3.8-23s 44l33 3.8.4.1.5 Emergency Cooling Tower Basin 3.8-26 3.8.4.1.6 Diesel Fuel Storage Tank Foundation 3.8-26 34 3.8.4.1.7 Electric Manholes 3.8-26 3.8.4.1.8 Confinement Structure 3.8-26 33 3.8.4.1.9 Interconnection of All Nuclear Island Seismic 3.8-26a Category i Structures to the Reactor Containment 44 Building O 3-x Amend. 61 Sept. 1981
TABLE OF CONTENTS (Continued) Pace No. 3.8.4.3.2 Creep, Shrinkage and Local Strees 3.8-29 3.8.4.3.3 Sodium Fire Load 3.8-29 3.8.4.3.4 Hot Sedium Spill Effect 3.8-29 3.8.4.3.5 Loading Combinatic.ns 3.8-29 3.8.4.4 Design and Analys's Procedures '.8-29 3.8.4.4.1 Analysis Procedures 3.8-29 3.8.4.4.2 Design Procedures 3.8-30 3.8.4.4.3 Reactor Service Building 3.6-31 3.8.4.4.3.1 Reactor Service Area 3.8-31 61 3.8.4.4.3.: Radwaste Area 3.8-31 3.8.4.4.4 Control Building 3.8-31 3.8.4.4.5 Steam Generator Butiding 3.8 3.8.4.4.6 Diesel Generator Building 3.8-32 3.8.4.4.7 Emergency Cooling Tower Basin 3.8-32 3.8.4.4.8 Diesel Fuel Storage Tank Foundation 3.8-F2 34 3.4.4.9 Confinement Structure 3.8-32 3.8.4.5 Structural Acceptance Criteria 3.8-32 3.8.4.6 Materiais, Quality Control, and Special Construction 3.8-32 44l3.8.4.6.1 Materiels 3.8-32a 3.8.4.6.2 Quality Control 3.8-33 3.8.4.6.2.1 Concrete 3.8-33 3.8.4.6.2.2 Reinforcing Steel 3.8-34 3.8.4.6.2.3 Structural and Miscellaneous Steel 3.8-35 3.8.4.6.3 Special Construction Techniques 3.8-35 3.8.4.7 Testing and in-Service Survelile- 9 Requirements 3.8-35 34 3.8.5 Foundations and Concrete Supports 3.8-35 Amend. 61 3-xi Sept. 1981 i.__..__ - ,_ _ . _ -- ~ _ . _ _ , _ _ . _ . - _ .
TABLE OF CONTENTS (Continued) Page No. 3.8.5.1 Description of the Foundations and Supports 3.8-35 3.8.5.1.1 General Description 3.8-35 3.8.5.1.2 Design Features 3.8-36 3.8.5.1.3 Load Transfer 3.8-36 3.8.5.1.4 Large Equipment Supports 3.8-36 3.8.5.1.5 Reinforcing Pattern 3.8-37 3.8.5.2 Applicable Codes, Standards and Specifications 3.8-37 3.8.5.3 Loads and Loading Combinations 3.8-37a 49 3.8.5.3.1 Loads and Loading Combinations for the Portion of the Mat Below RCB 3.8-37a 3.8.5.3.2 Loads and Loading Combinations for the Mat Below Ci.her Seismic Category I Structures 3.8-38 1 3.8.5.3.3 Loads and Loading Combinations for Other Concrete Supports 3.8-38 g 3.8.5.3.4 Loads and Loading Combinations for Foundation Stability 3.8-38 3.8.5.4 Design and Analysis Procedure 3.8-38a 1 3.8.5.4.1 Combined Mat 3.8-38a l 3.8.5.4.2 Reactor Concrete Support Structure 3.8-39 3.8.5.5 Structural Acceptance Criteria 3.8-39 3.8.5.5;l Stress 3.8-39 i 3.8.5.5.2 Strain 3.8-39 3.8.5.5.3 Gross Deformation 3.8-39 1 3.8.5.5.4 Differertial Settlement 3.8-39 3.8.5.6 Materials, Quality Control and Special Construction 3.8-39 l 3.8.5.7 Testing and Inservice Surveillance Requirements 3.8-39 l l 3-xii Amend. 49 Apr. 1979 1
LIST OF FIGURES (Continued) Page No.
" 3.7-16I 0.250g SSE Rotational N-S Response Spectrum for Common Base Mat at El. 733' Node 48 (3% Damping) 3.7-42g s 3.'7-16J 0.250g SSE Horizontal E-W Response Spectrum i
for Reactor Containment Building at El. 800' Node 10 (3% Damping) 3.7-42h
- 3. 7-16 K 0.250g SSE Torsional E-W Response Spectrum for Reactor Containment Building at El. 800' Node 10 (3% Damping) 3.7-421
- 3.7-16L 0.250g SSE Horizontal N-S Response Spectrum for Reactor Containment Building at El. 800' Node 10 (3% Damping) 3.7-42j 3.7-16M 0.250g SSE Torsional N-S Response Spectrum for Reactor Containment Building at El. 800' Node 10 (3% Damping) 3.7-42k 3.7-16N 0.250g SSE Vertical Response Spectrum for Reactor Containment Building at El. 800' O Node 11 (3% Damping) 3.7-421 3.7-160 0.250g SSE Rotational E-W Response Spectrum for Reactor Containment Building at El. 800'
] Node 10 (3% Damping) 3.7-42m 3.7-16P 0.250g SSE Rotational N-S Response Spectrum for Reactor Containment Building at El. 800' Node 10 (3% Damping) 3.7-42n 3.7-16Q 0.250g SSE Horizontal E-W Response Spectrum for Reactor Service Building at El. 816' Node 29 (3% Damping) 3.7-42o 3.7-16R 0. 250g SSE Vertical Response Spectrum for Reactor Building at El. 816' Node 30 (3% Damping) 3.7-42p 3.7-16S 0.250g SSE "orizontal E-W Response Spectrum for Steam Generator Building at El. 806' , Node 42 (3% Damping) 3.7-42q l
- 3. 7-16T 0.250g SSE Vertical Response Spectrum for Steam Generator Building at El. 806' Node 43
- j. 49 o (3% Damping) 3.7-42r
~.
3- xviii Amend. 49 Apr.1979 l l
. _ . _ . _ _ _ ~ _ . . _ _ _ _ _ . _ . _ _ . _ . . _ . . _ _ . .
LIST OF FIGURES (Continued) Page No. 3.7-17A Reactor Configuration Normal Operation 3.7-43 3.7-17B Equipment Arrangement Assumed in Refueling Model 3.7-43a 3.7-17C Equipment Arrangement Assumed in Preparation for Refueling Model 3.7-43b 3.7-17D Reactor System Computer Model 3.7-43c 49 3.7-18 Reactor System Computer Model 3.7-44 3.7-18A CRBRP Control Rod Driveline Seismic Model - Element Identification 3.7-44a 49 3.7-188 Contact Force on Translating Assembly Versus Time During 0.25g SSE 3.7-44b 3.7-19 Plan 3.7-45 3.7-20 Plan Elevation 3.7-46 3.7-21 Finite Element Model - N.S. Direction N.S. Horizontal Soil Spring Ratio 3.7-47 3.7-48 h 3.7-22 3.7-23 Finite Element Model - E.W. Direction 3.7-49 44 3.7-24 Finite Element Model - Torsional Spring 3.7-50 34 3.8-1 Diesel Fuel Oil Storage Tank Foundation 3.8-48 3.8-2 Nuclear Island Structures - Connon Mat Plan 3.8-49 i 3.8-3 DELETED 3.8-50 61 3.8-4 DELETED 3.8-5 Typical Vertical Cross-Section of PHTS Cell 3.8-52 3.8-6 Horizontal Section A-A of PHTS Cell 3.8-53 34 ,3.8-7 Horizontal Section B-B of PHTS Cell 3.8-54 l l 3-xix Amend. 61 l Sept. 1981 l
(a) The steam generators, which form a passive barrier for the Intermediate system and provide the heat removal interface to the ultimate sink.
.O (b) The IHTS main loop piping and flowmeter, which also form part of the intermediate coolant boundary.
(c) The IHTS dump valves - These form part of the Intermediate coolant boundary, but do not have any safety function other than maintaining the boundary. (d) The IHTS dump piping - the piping upstream of the first dump valve is also part of the intermediate coolant boundary. (e) The IHTS sodium pump - This pump must be capable of operating at pony motor speed. In addition it forms part of the intermediate coolant boundary. (f) The IHTS expansion tank and connecting piping - This is part of the intermediate coolant boundary, some nozzles on the upper head of the expansion tank are connected to the auxillary systems, which are not required for the IHTS to perf orm its saf ety function. (g) The IHTS vent Iines from the steam generators to the expanulon tank - These lines are normally full of sodium and the flow is determined by the pressure dif ferential between the steam generc$or sodium and the expansion tank cover gas. These lInes form part of the intermediate coolant boundary. (h) The SGS leak detection systems are attached to the IHTS pipes at various locations and the piping to the detectors is an extension of the Intermediate coolant boundary. Isolation valves can separate the leak detectors themselves from the Intermediate system. The piping up to the i3rst isolation valve and that valve are part of the intermediate coolant boundary. g (!) Rupture discs which protect lHTS and lHX from sodium water reactions in Steam Generators are part of the intermediate coolant boundary and provide the safety function of ilmiting the pressure transients during i a sodium water reaction, l ! (J) Piping to the sodium service system upstream of the isolation valves. (k) Service system isolation valves. 16 l O Amend. 61 l 3.1-5 Sept. 1981
(1) IHTS Control and instrumentation - The control functions associated with IHTS pony motor operation must remain functional and the IHTS dump valves must not receive falso operating signals. 16 NORMAL OPERATION Normal operation means steady state operation and those departures from steady state operation whica are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant, it includes conditions such as startup, normal shutdown, standby, load following, limited fuel rod leakage, operation with specific equipment out of service as permitted by Technical Specifications, and routine Inspection, testing and maintenance of components and systems during any of these conditions, if it is consistent with the Technical Specifications. OFF-NORMAL CONDITIONS Of f-normal conditions mean those steady state and transient conditions not part of normal operation which (1) individually may be expected to occur once or more during the plant lifetime and include but are not limited to an inadvertent control rod withdrawal, tripping of sodium circulating pumps, failure of all offsite power, and tripping of the turbine generator set or (2) which Individually are not expected to occur during the plant lifetime; however, when Integrated over all plant components and systems, events in this category may be expected to occur a number of times. Events in (1) are termed 61 Anticipated Faults and events in (2) are termed Unlikely Faults. EXTREMELY UNLIKELY FAULTS Events of such extremely low probability that no events in this category are expected to occur during the plant lifetime, but which neverthe ess represent extreme or limiting cases of f ailures which are identified as possible. These extremely unlikely events, which are design bases, shall encompass a spectrum of events appropriate to the design. These may include, for example, a large sodium fire, a large sodium-water reaction, and a rupture of a radwaste system tank. Inert Atmosohere. Inert atmosphere means a gas or gaseous mixt'ure iImited in oxygen and other substances that are chemically reactive with sodium so that chemical reactions wIlI not signifIcantly increase the consequences of contact with sodium. 32 Heat Transoort System. The heat transport system is the aggregate of systems and/or components containing the heat transport fluids and used for extracting heat from the reactor and transporting it to the equipment used for electrical power conversion during normal operation or, af ter plant shutdown, to an ultimate heat sink. It does not include systems whose prime function is the cooling w structures or equipment. l \ O i 3.1-6 Amend. 61 Sept. 1981 l 1
Reactor Residual Heat Extraction' System. The reactor residual heat
, extraction system is the portion of the heat transport system which, s
after plant shutdown, extracts from another portion of the heat trans-port system heat originating in the reactor and transports this heat to the ultimate heat sink. Ultimate Heat Sink. The ultimate heat sink is that heat tink including necessary retaining structures (e.g., a river with its dam, or a pond with its dam) to which reactor decay heat and essential cooling system heat loads are dissipated following normal reactor shutdown or shutdown 32 after an accident. Fuel Design Limits. Fuel design limits means those limits such as temper-ature, burnup, fluence, and cladding strain which are specified by the designer for normal operation and anticipated operational occurrences. 3.1. 2.1 Comparison of Plant Conditions with 10CFR50 The text of Section 3.1.2 gives clear definitions of five plant conditions: Nonnal Operation, Anticipated Faults, Off-Normal Conditions, Unlikely Faults, and Extremely Unlikely Faults, which are consistently 32 used throughout Chapter 15 of the PSAR. In 10CFR50, Appendix A and PSAR Section 3.1, a total of three categories of operational conditions and events are used, namely: Nonnal Operation (defined previously ). Anticipated Operational Occurrences, and Postulated Accidents, defined 25 33 below. Anticipated Operational Occurrences. Anticipated operational occurrences means those steady state and transient conditions not part of normal oper-ation which might occur one or more times during the life of the nuclear power unit and include, but are not limited to, an inadvertent control rod withdrawal, tripping of sodium circulating pumps, failure of all offsite power, and tripping of the turbine generator set. , Postulated Accidents. Postulated accidants means those events which, . although not expected to occur, are selected, in addition to normal and anticipated operational occurrences for establishing design bases of systems, components and structures and/or selection of Exclusion Distance and Low Populction Zone for the reactor site. The accident sequences may be postulated to be the natural consequences of l possible accidental events, or, alternatively may be hypothesized for j purposes of safety system evaluation and site analysis. These may in-clude, for example, reactivit/ trmients causing core damage, fuel handling accidents, a large sodi; '1, a large sodium-water reaction, rupture of a radwaste system tank, a reactor malfunction resulting in ( a . loss of core flow or reactivity insertion combined with a potential l commc.. mode failure of the protection system, and a non-mechanistic ! 32 release of fission products for purposes of site analysis. ' Amend. 33 3.i-/ Jan. 1977
-._, ..r, . . - - - _ _ . ,_ _ . , ~ . , _ . , . - ,. .- . , . . _,-_ m _ . , _ , - ~ . - . , , , -,. , . -
For clarifica'tlon, a one-to-one correspondence of the conditions is shown in Tabl e 3.1-3. 25 3.1.3 Conformance with CRBRP General Deslan Criteria 3.1.3.1 Overall Recuirements Criterion 1 OUALITY STANDARDS AND RECORDS Structures, systems, and components important to saf ety shalI be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and suf fIclency and shalI be supplanented or modified as necessary to assure a quality prod ct in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components wilI satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shalI be maintained by or under the control of the nuclear power unit iIcensee throughout the iIfe of the unit.
RESPONSE
The design of this plant conforms to the Intent of this criterion. The design criteria for structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The CRBRP structures, systems and components have been analyzed in accordance with the basic intent of 10CFR50, Section 50.55a and Regulatory Guide 1.26, and have been classified as Safety Class 1 (SC-1), Safety Class 2 (SC-2), or Safety 61 Class 3 (SC-3), commensurate with the Importence cf the safety functions to be perf orraed. The safety class assignment is to be considered in the design, f abrication, construction, erection, test and operation of the plant. Further details are provided in PSAR Section 3.2. Codes and Standards to be employed in the design, f abrication, erection and testing of the plant are Identified and evaluated for applicability, adequacy
- and suf ficiency, and as necessary are supplanented or modified to assure a quality product in keeping with the required safety function.
A quality assurance program has been estabiished and irrplanented in order to provide adequate assurance that the structures, systems and cunponents wili satisf actorily perf orm their Intended service. The progran complies with the requirements of the contracts and the execution of the program complies with the requirements of 10CFR50, Appendix B. The quality assurance program controls the quality-related activities throughout the life of the project and is documented. Appropriate records of the design, fabrication, erection, and testing of structures, system and components important to safety are maintained under too control of the nuclear power plant licensee throughout the life of the plant. Procedures define those records which are necessary to document the quality of the structures, 3.1-8 Amend. 61 Sept. 1981
O G Criterion 3 FIRE PROTECTION Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety require-ments, the probability and effect of fires and explosions. Noncombus-tible and heat resistant materials shall be used wherever practical' throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components impor-i tant to safety. Fire fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of structures, systems, and components.
RESPONSE
The Non-Sodium Fire Protection System provides the plant with equipment, piping, valves, detectors, instrumentation and controls to prevent or mitigate the consequences of a non-sodium fire. It consists of the following: Water Supply System Wet Sprinkler System Preaction Sprinkler System Water Spray System h V Halon 1301 Gas Blanketing System Standpipe System Portable Fire Extinguisher System Fire Detection System 54l Fixed Dry Chemical System The general description of the above systems is provided in Section 9.13.1 and Table 9.13-4. The fire prevention and protection systems to be provided for all the areas associated'with the safety related structures, systems and components are listed in Table 9.13-3. In areas with safety related structures, systems and components, l the Non-Sodium Fire Protection System piping and components (such as sprinkler heads) will be designed so that neither piping failures nor inadvertent operation of the :ystem fire protection components due to a , seismic event will result in the loss of function ~of safety related l structures, systems and components. This is accomplished through the l use of seismically qualified pipe supports, and dry pipe preaction 32 sprinklers within areas containing safety related equipment. Standpipes l 48 l l 3.1-11
)
d
- l. Amend. 54 May 1980 y .c.-,y%ya u. - p- ., .c y -wa. ,- y,- .
9 .,p-, - , , ,, y y ,. . , _,
serving a Seismic safety-related equipment Category I water are supply Seismic system Category I and if necessary. will be Building supplied by l1 48 isolation valves will be specified as Seismic Category 1. Electrical power for the Fire Protection System will be provided frem the normal plant AC pover distribution system. If normal AC power is unavailable, the water supply system pressure will be maintained by two diesel-driven fire pumps, and the fire detection system will be energized by a non-Class IE 4-hour DC battery / Inverter system that has the capability of being connected to an emergency diesel generator through qualified Isolation devices. The Non-Sodium Fire Protection System will be designed in accordance with applicable codes and standards. 48 Five barriers will provide Isolation between areas such as: g Steam Generator Building, Stean Generator Bay from intermediate Bay, Maintenance Bay, Auxiliary Bay end Diesel Generator Building. Access to alI bulldings, other than the Reactor Containment Bullding, wilI be designed such that there will be multiple means of access for operating personneI and there wiII be multiple means of access for fire fighting personnel. The largest potential source of fire from fuel oil is in the vicinity of the standby diesel generator f uel oil storage tanks, located below grade adjacent to the Diesel Generator Bullding. As these tanks are located below grade, the chance of an accident is reduced. Physical separation provided between the two tanks limits the spreading of fire from one tank to the other. Since either tank is capable of f ulfilling the emergency fuel oil requirements, a safe shutdown of the plant will not be jeopardized by a fire in either tark. Charcoal filters will be bounded and separated by fire barriers, and the filter units will be made redundant, so that safe shutdown of the plant will not be 48 Jeopardized by a fire in either filter. Table 9.13-3 lists the safety related areas of the plant containing combustible materials. The burning characteristics of these materials such as maximum fire intensity, flame spread, smoke generation and toxicity of combustion products are listed in Table 9.13-2. A detailed fire hazards analysis will be provided in the FSAR and will evaluate the potential fire hazards throughout the plant and the ef fect of postulated design basis fires relative to maintaining the abil ity to perf orm saf ety shutdown f unctions and minimizing radioactive releases to the environment. This analysis wilI serve to confirm the adequacy 32 of 18 0 3.1-12 Amend. 61 Sept. 1981
Criterion 4 PROTECTION AGAINST S0DIUM REACTIONS Systems, components and structures containing sodium shall be designed h v to limit the consequences of sodium chemical reactions resulting from a sodium spill. Special features such as inerted vaults shall be provided as appropriate for the reactor coolant system. Means to detect sodium or sodium reaction products and fire control systems shall be provided to limit and control LSe extent of such reactions to assure that the functions of components important to safety are maintained. Means shall be provided to limit the release of sodium reaction products to the environment as necessary to protect plant personnel and to avoid undue risk to the public health and safety. Materials which might come in contact with sodium shall be chosen to minimize the adverse effects of possible chemical reactions. In areas where sodium chemical reactions are possible, structures, components and systems, important to safety, including electrical wiring and components, shall be designed and located so that the potential for damage by sodium chemical reactions is minimized. Means shall be provided as appropriate to minimize possible contact between sodium and water. The effects of possible interactions between sodium and concrete shall be considered in the design. The sodium-steam generator system shall be designed to detect sodium-water reactions and limit the effects of the energy and reaction products released by such reactions so as to prevent loss of safety functions of the heat transport system.
RESPONSE
/~
Protection against sodium reactions is provided by:
- 1. The use of stainless or carbon steel.for tanks; components and piping containing sodium or NaK;
- 2. The use of steel cell liners and drip pans in concrete cells to prevent any concrete-sodium reaction in the event of a spill;
- 3. The use of ine.lation approved for sodium systems with an inner and outer sheath of stainless steel to minimize absorption in the insulation;
- 4. The use in NaK coolers of air or an auxillary coolant fluid which will not produce an exothermic reaction.
l 5. The use of suitable instrumentation to detect any sodium reactions . The instrumentation to detect sodium reactions and to control the reaction suppressant dispensing system is described in Section 9.13.2. The cells are either inerted or are provided with fire control capability, electrical equipment is above the normal expected depth of any sodium spill, and the electrical wiring is so located as to minimize damage f rom sodium fires. The Steam Generator System is provided with subsystems to detect sodium-water leakage and to limit any reaction effects. These are discussed in D Sections 7.5 and 5.5, respectively. 32 Amend.~32 3.1-13 Dec. 1976
Criterion 5 - Environmental
- and Missile Deslan Bases Structures, systems, and components important to saf ety shall be designed to O cccommodate the ef fects of, and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, anticipated operational occurrences, and postulated accidents. These structures, systems, and components shalI be appropriately protected against dynamic of tects, such as the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment falIures and from events and conditions outsIde the nuclear power unit.
Resoonse: All plant locations containing safety related control and electrical equipment, that need a controlled environment to maintain the required operability, are to be provided with redundant air conditioning and/or ventilation facilities for the needed environmental control. 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. The safety-related systems which are required to function during and following any identified accident are Identified in Section 7.1. Worst case environmental conditions will be defined for each location. Where possible, the equipment comprising the safety-related l&C systems is located in controlled atmospheres (e.g., control room). For this equipment, the worst case environments are tt.ose resulting from malfunctions of H&V or power source systems. Safety-related equipment located in the Containment, the Steam Generator Building, the Reactor Service Building, the Control Building 61 cnd the Diesel Generator Building will be designed to operate through, or be protected from, the worst environmental conditions for wM ch the equipment must perform. Environmental conditions which will be conoideren. ~n design include temperature, humidity, atmosphere and shock and vibration. Design
- considerations will also be given to typical environmental conditions for which l protection will be provided such as products of sodium fire, high radiation, or l steam / water atmosphere due to postulated 1!re. Protection will include location in a separately ventilated room, onclosure of cabinets that prevented entry of reaction products, distance from radiation sources, etc.
l I 32 CNatural phencoena are covered by Cr;terion 2. 1 0 3.1-14 Amend. 61 Sept. 1981
Criterion 8 Reactor Design Cl The reactor and . associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel desigr. limits are not ex': %d during any condition of normal operation including the effects of ant'-- , '. operational occurrences.
RESPONSE
This criterion is satisfied by the following two design bases. 5
- 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 S1 l 115,000 mwd /T with an average burnup of 80,000 mwd /T. These 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 equilibrium core conditions. Mainten-
/] ance of fuel rod structural integrity is a design basis should an V Unlikely Fault occur during the fuel residence time.
- 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 design parameters, assure that incipient fuel melting does not 51 l ccur in the fuel pellet with peak power. The superimposed effects l 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 the highest radial and axial power factors,15% overpower conditions and 3a nuclear and engineering uncertair. ties, is less than that which results in fuel melting.
l l 3.1-17 Amend. 51 Sept. 1979 ! V
Criterion 9 Reactor inherent Protection The reactor and associated coolant systems shall be designed so that in the power operating range the not effect of the prompt inherent nuclear feedback characteristics tent to compensate for a rapid increase in reactivity.
RESPONSE
The following design basis satisfies this criterion: The Doppler offect provides the prompt negative reactivity feedback which is required to mitigate the effects of reactivity transients (rapid power increases). Therefore, the fuel temperature (Doppler) coefficient shall be strongly negative when the reactor is critical. The negative Doppler coef ficient is obtained through the inherent use of fuel with a large propcrtion of U-238. The Doppler coefficients for each major fueled reactor region have been calculated at the beginning and end of cycle for both the first and equilibrium cores 51l with FFTF-grade (Iow Pu-240) plutonium f uel (See Table 4.?-16) . In all cases, the Doppler coefficients are strongly negative. The analysis of accideat conditions, presented in Chapter 15, uses conservative values of the Doppler reactivity feedback coef ficient (nominal value less 3 uncertainty). At low power / flow ratio operating conditions as during the reactor startup, positive bowing reactivity effects are predicted. The not 61 reactivity feedback during this power-to-flow ratio range is evaluated to conservatively envelope all possible combinations of bowing and compensating nogative reactivity offects. For certain assumptions on assembly bowing behavior a net positive reactivity feedback is predictad over a portion of the line power-to-flow ratio range. The PPS can safely accommodate all design basis events initiated in the start m range when the above worst case offects are considered. Studies have been performed for a range of startup overpower transients. These have demonstrated that, even ne.c,lecting the offects of the plant protection system, the integrated r%ctivity feedback from the point of the initiation of the transient up to full power temperatures is always negative. Consequently, reactor power and temperatures are bounded even when worst case reactor feedback 61, characteristics are utilized. Maximum temperature values fall well below values which are expected for normal power operation l demonstrating satisfactory reactor inherent protection. 61l As the power-to-flow ratio approaches 1.0 and at higher power-to-flow ratios (>1.0), reactor assembly bowing reactivity is negative and 42 enhances the effect of negative Doppler which is discussed above. O Amend. 61 3.1-18 Sept. 1981
After being placed in service, the standby diesel generators and their respective associated supply systems will be inspected and tested O periodically' to detect any degradation of the system. (See Section 3.3.1.1.1) Initial pre-operational tests will be performed with equipment and components installed and connected to demonstrate that the equipment _is witnin~ design limits and the system meets performnce specifications. This test will also demonstrate that loss of the Plant Power Supply and offsite AC power supplies can be detected. Periodic equipment tests will be performed to detect cny degradation of the system and to demonstrate the capability. of equipment which is normally de-energized. The test me thods utilized are detailed in Sect, ion 8.3.1.1.2. Periodic tests of the transfer of power between the Plant Power Supply and offsite AC power supplies are performed during prolonged plant shutdown or during refueling to demonstrate that:
- s. Sensors can properly detect loss of the Plant Power Supply and-the offsite AC power supplies.
- b. Components required to acconplish the transfer from the Plant Power Supply to the Preferred AC Power Supply are operable.
- c. Components required to accomplish the transfer from the Preferred AC Power Supply to the Reserve AC Power Supply are operable.
- d. Components required to accomplish the transfer from the Reserve AC Power Supply to the Standby AC Power Supply are operable,
- e. Components required to accomplish the transfer from the Plant Power Supply (simulating the unavailability of the offsite AC power supplies) to the Standby AC Power Supply are operable.
- f. Instruments and protective relays are properly set and operating correctly.
The 120 V Vital AC System components are inspected and tested at!the 53 Gendor's facilities. The system is- also inspected during instal _lation. When the i installation is complete, preoperational equipment tests and inspections are
' performed to demonstrate that:
A. Components are correct and properly nounted. B. Connections are correct and the circuits are continuous. C. Components are operational. D. Instruments and protective devices are properly calibrated and adjusted. . hi The initial system t'sts e will also demonstrate 'that while su plied by the DC power systems or the'480-120/208 V Instrument AC Regulating ? transformer, 53l-
- 32 the 120V Vital AC Power System can supply power. to the design load as required.
' Amend.-53 3.1-33 Jan. 1980
Periodic tests are performed to detect any deterioration of the equipment and to demonstrate the capability of equipment which is normally energized. Provision is included in the design for testing the transfer of power between the unit station service transformers and the reserve transformers. These tests are perf ormed during prolonged plant shutdows; periods by simulating loss of the AC power supply from ti e unit station service transformers as described in Section 8.3.1.1.2. Provisions are also included in the design for testing the operability and performance of equipment. The tests include a preoperational eqt ipment test, initial system iest, and periodic equipment and system tests as described in Sections 8.2, 8.3.1.1.1 and 8.3.1.1.2. Criterion 17 - Control Room A control room shalI be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions (including those conditions addressed in NRC Criterion 4 - Protection Against Sodium Reactions). Adequate radiation protection shalI be provided to permit access and occupancy of tne control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and e'th a design capability for suosequent control of the reactor at any coolant temperature lower than the hot shutdown conditions. Resoonse: 49 The control room design is based on proven power plant design philosophy. All control stations, switches, controllers, and Indicators necessary to operate and scutdown the plant and to maintain safe control of the reactor will be 32 located in tre control room. The design cf the control room will permit safe occupancy during abnormal condit!cns. The doses to personnel during accident conditions from containment buildino shine, radioactive clouds and Ingress / egress to the exclusion boundary are less than 5 rem wiiole body, or its equivalent to any part of the bcdy. These doses and criteria are detailed in Section 6.3. 49 3.1-34 Amend. 61 Sept. 1981
, levels for all stressed elements of the containment boundary. Details of the j containment design are given in Sections 3.8.2.and 6.2. ! Criterion 43 CAPABILITY FOR CONTAIPMENT LEAKAGE RATE TESTING 4 ! The reactor containment and other equipment which may be subjected to ! containment test conditions shall be designed so that periodic integrated leakage
- rate testing can be conducted at containment design pressure.
I
Response
, The reactor containment hsign will permit overpressure strength testing
- during construction and permit preoperational integrated leakage rate testing
- in accordance with Appendix s of 10CFR50. All equipment which may be subjected ,
- to the test pressure will be designed or arranged with suitable provisions so i
^ that periodic integrated leakage rate testing can be conducted. Further details are provided in Section 3.8.2 and 6.2. l Criterion 44 PROVISIONS FOR CONTAINMENT TESTING AND INSPECTION l The reactor containment shall be designed to permit 1 appropriate periodic i L inspection of all important areas, such as penetrations, 2 an appropriate sur-ve111ance program, and (3) periodic testing at containment design pressure of { the leak tightness of penetrations which have resilient seals and expansion' bellows. Respons_e: The reactor containment and the containment isolation system will be designed so that appropriate periodic inspection of all important areas such as penetrations can be made. The design will also be such that an appropriate surveillance program- . can be maintained. The design will permit periodic testing at containment design . pressure of the leak tightness of isolation valves and penetrations having resi-lient seals and expansion bellows. It will also permit demonstrating periodically
- the operability of the containment isolation system. The containment will be
- inspected and tested to heed the requirements of.10CFR50 Appendix J. Further 32 infornation is given in Section 6.2 1L i
Amend. 32. Dec. 1976 j , 3.1-69
Criterion 45 PIPING SYSTEMS PENETRATING CONTAINMENT Piping systems penetrating reactor containment shall be provided with leak detection, Isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating the piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the Isolation valves and associated apparatus and to determine if valve leakage is within acceptable iimits. Resoonse: The design of the piping systems penetrating reactor containment conforms to tho erIterion In accordance wIth CrIterie 46 (Reactor Coolant Boundary Penetrating Containment), 47 (Primary Containment isolation) and 48 (Closed Systems Penetrating Containment) as shown h Table 6.2-5. The containment isolation features of the a;, sign of lines penetrating ) containment provide the necessary assurance that the containment system will provide the barrier to release or spread of radioactive gas or particulate matter. For lines of closed systems penetrating containment, one isolation valve located outside of containment as close as practical to containment is 61l provided. A single valve meets the criteria and provides the necessary l capability to limit the release of activity. The valves and associated actuators are located in protected areas and are testable. Manual Initiatio', of Isolation is provided. l For the lines connected to the reactor coolant boundary, or containment 61 atmosphere, two valves, either manually or automatically actuated as appropriate, provide the necessary protection. 1
- 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. Ranote and local manual Initiations are provided.
The nitrogen exhaust line to CAPS has two automatically actuated valves. The ex-containment portions of the system are protected against the ef fects of severe natural phencinena. Valve closure signal is initiated by the plant protective system. The valves provide two barriers following closure. The valves and associated actuators are located in protected areas and are testable. The two valves in the argon exhaust to RAPS and the two valves in the Gas Sampling Line close automatically on signal from the Plant Protective System. 32 All valves are protected and testable. 1 O Amend. 61 3.1-70 Sept. 1981 l
Automatic Isciation of the lines for containment air ventilation and those for containment vacuum breakers is provided by two isolation valves for the containment vacuum breakers, and three isolation valves for the containment air ventilation, with independent actuating trains. One valve is inside 61 containment; and one and two outside as close as practical for the containment vacuum breakers, and for the containment air ventilation, respectively. 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 areas which are protected from tornado generated missiles and which are designed to withstand the seismic forces. The IHTS piping within containment out to the end of the penetration seal is protected from inadvertent accidents and natural phenomena by being totally enclosed with reinforced concrete cells (248, 251, and 252) which serve as radiation shields within the Intermediate bay of the steam generator building (Figures 1.2-13 and 20) . The IHTS piping is designated Safety Class 2, Seismic Category 1, and classified as ASE Section 111, Class 2, designed and constructed to Class 1 requirements. Since the entire IHTS is a closed system and is neither part of nor directly connected to either the containment atmosphere or the primary coolant boundary, and is protected as described 32 above, Isolation valves are not required (Tables 3.2-2, 4, and 5). O i l l O V 3.1-71 Amend. 61 Sept. 1981 t t g g +- e n v nv -
~w .--,
-,-p -
Criterion 46 - Reactor Coolant Boundary Per ating Containment Each line that is part of or directly cc ected to the reactor coolant boundary and that penetrates reactor c' .ainment shall be provided with containment isolation valves as tollows, unless it can be demon-strated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: (1) One locked closed isolation valve inside and cne locked closed isolition valve outside containment, er (2) One automatic isolation valve inside and one locked closed isolation valve outside containment, or
- 3) One ic:ked closed isolation valve inside and one auto-matic ischtion 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.
Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, auto-matic isolation valves shall be designed to take the position that provides greater safety. Other appropriate requirements to minimize the probability or conse-quences of an accidental rupture of these lines or of lines con-nected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these require-ments, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.
Response
Those lines which are part of or directly connected to the reactor coolant boundary and penetrating the containment are the argon supply line, the argon exhaust to RAPS, and the gas sampling line. 3.1-72 Amend. 32 O Dec. 1976
J
~
Each line will be provided with one automatic isolation valve in side containment and one automatic isolation valve outside contain-ment. (See Tabie 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 to the containment as practical and the automatic isolation valves are designed to take the position that provides greater safety upon loss of actuating power. Appropriate measures will be taken to minimize the probability cr consequences of an accidental rupture of these lines or lines connected to them to assure adequate safety. (More details are provided in section 6.2.4.2.)
32 4 O l 2 () 3.1-73
,me,c. 3, Dec. 1976-i f
l l
. _ . . . . . _ - _ . - , . _ . ~ . . . _ . . _ , - . . . , - . . . _ . _ _ _ _ _ . . - . . . - - . . ~ . _ . . . _ . , . . . . . . - . . . _ _ , ~ . . . . . . , . . .
Criterion 47 - Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment Isolation valves as folIows, uniess It can be demonstrated that the containment Isolation provisions for a specific cless of lines, such as instrument lines, are acceptable on some other defined 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 valve outside containment, 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 automat!c Isolatica valve outsIde contalnment, 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. Isolation valves outsIde containment shalI be Iocatod as elose 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 following lines penetrate the reactor containment and are directly connected to the contairment atmosphere: Containment Ventiiatton Air Supply I.ine Containment Ventilation Air Exhaust Line Containment Vacuum Breakers Each of these lines, except the containment vacuum breakers will be provided 61 with three confinement / containment isolation valves, one immediately outside l the confinement and one inside the containment, with Independent actuating , trains. ' The va!ves and associated actuators will close on loss of air or electrical power. Because the system operating pressures are low and the closure times required for the containment isolation valves are four seconds, the dynamic j forces resulting from the inadvertent closure under operating conditions wil! l not challenge the Integrity of the valves or connecting piping. However, a l quick acting automatic relief damper will be provided in a branch duct between , the Air Supply Line containment isolation valves and the supply fans in order l to relleve any excess pressure on the ductwork originated by the activation of the containment isolation valves. A relief damper is provided in i-he exhaust 32 air line between the Isolation valves and the exhaust fans, to relieve the i vacuum in the exhaust duct af ter the isolation valves close. It, addition, upon l containment isolation, the containment ventilation supply and exhaust f ans are 61 automatically stopped. 3.1-74 Amead. 61 Sept. 1981 1
I Criterion 48 - Closed Systen Penetrating Containment Each line that penetrates primary reactor containment and is neither part of nor directly connected to the reactor coolant boundary, nor connected directly to the containment atmosphere shall have at least one containm:nt isolation valve, unless it can be demonstrated that containment isolation provisions for a specific class of lines are
-acceptable on some other defined basis. The isolation valve, if re-quired, shall be either automatic or locked closed, or capable of remote manual op2 ration. This valve shall be outside containment and located as close to the containment as. practical. A simple check valve may not Le used as the automatic isolation valve.
Response
Each of the following lines of closed systems penetrates the reactor containment and is neither part of the reactor coolant boundary nor connected directly to the contairinent atmosphere: Sodium Transfer Line Between Storage Tanks (Section 9.3) Sodium Transfer Line from EVST (Section9.3) DHRS NaK Line to Containment (Section9.3) DHRS NaK Line from Containment Section9.3) Normal Chilled Water to Containment Section 9.7) Normal Chilled Water from Containment Section 9.7 Emergency Chilled Water Supply (Section 9.7 ' Emergency Chilled Water Return (Section 9.7 Each of these lines has at least one c.ontainment isolation valve capable of remote manual operation and located outside and as close to containment + ' as practical. These lines and the associated containment isolation valve designs are discussed in Section 6.2.4. The IHTS has been judged to be an accer table isolation boundary without ! the inclusion of isolation valves because of (1) the precautions taken to i ' protect the IMTS boundary against accidents, extreme environmental condi-tions, and natural phenomena, (2) the ability to monitor the integrity of the boundary and (3) upon the acceptability of the' radiological consequences
- which would result from a failure of the boundary. The basis for.this 32" judgenent is discussed in more detail in Section 6.2.4.1 and 15.6.1.5.2.
l i 3.1-75 Amend. 32-l Dec.1976 l i. 719P8WQeg***_
Criterion 49 - Containment Atmosohere Cleanuo Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shalI be provided as necessary to reduce, consistent with the f unctioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained. The necessity of such systems should consider the ef fects of sodium leakage and its potential reaction with oxygen and its potential for hydrogen generation when in contact with concrete. Each system shalI have suitable redundancy in components and features, and suitabl e interconnections, leak detection, Isolation, and con +ainment capabilities to assure that for onsite electric power system operation (assuming of fsite power is not available) and for of fsite electric power system operation (assuming onsite power is not available) Its safety function can be accomplished, assuming a single failure. Resoonse: 3, During normal operation, the confinement / containment annulus will be maintained 4 at a minimum of 1/4" water gauge negative p essure with respect to the outside atmosphere by exhausting 3000 CFM of filtered air through one of two ESF annulus pressure maintenance fans. Upon a containment isolation signal, both the annulus pressure maintenance fan O and the annulus filter fan with its associated filter unit will operate. During this condition, only a portion of the total air flow (3000 CFM) is exhausted to the outside atmosphere and the remainder of the total air flow (11000 CFM) is returned back to the annulus space below the 816'-0" elevation 61 where it is relieved to the upper annulus through equally spaced openings at elevation 816'-0". The filter system will be designed as an ESF system, and wIlI comply wIth Regulatory Gulde 1.52. The fIlter system wIl! be designed to 36 achieve 1 minimum of 99% particulate and 95% absorbent offIclency. Radiation monitoring equipment associated with the annulus filtration system is described in Section (same notation) 12.2 of the PSAR. minimum of 1/4" water gauge negative pressureBy maintaining with respect tothe the annulus outside at a l36 atmosphere, the bypass leakage (that fraction of annulus radioactivity which leaks from the confinement building without being filtered) can be maintained at less than 1%. The annulus filtration system features of the design provide the necessary assurance that the radioactivity released as a result of the design basis 32 accident will not exceed the guidelines of 10CFR100. 36 3.1-76 Amend. 61 Sept. 1981
Criterion 53 FUEL STORAGE Am HAELING AE RAD 10ACTlVlTY CONTROL l The fuel storage and handling, radioactive wasie, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accideat conditions. These systems shall be designed (1) with a capability to permit appropriate perledic inspection and testing of components important to safety, (2) with suitable sh!elding for radiation protection, (3) with~ appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat remove!, and (5) to provent significant reduction in fuel storage coolant inventory under accident conditions.
RESPONSE
Fuel storage far!litles and fuel handling equipment important to safety are designed to provide accessibility for performing inspection, maintenance and testing activities. All fuel storage facilities and fuel handling equipment will be shielded for radiation protecticn to meet the requirements specified in 10CFR20, 50 and 100, and Regulatory Guide 8.8. Containment, confinement, and filtering are provided as required for all fuel storage f acilities and fuel handling equipment containing radioactive material to Iimit any radioactive releases below those radiation doses specified in 10CFx20 and 100 as appropriate. Adequate cooling capability is provided for spent f uel storage and spent fuel handling equipment to assure decay heat removal with enough reliability, independence and redundancy to accommodate all plant conditions. A significant reduction of sodium coolant Inventory in the spent fuel storage v f acilities under accident conditions will be prevented by employing high quality design and construction standards to the spent fuel storage vessels, by guard Jackets surrounding the storage vessels and by anti-syphon features. The des!gn measures necessary to meet this criterion are described in Section 9.1 for the fuel storage and handling system. Criterion 54 PREVENTION OF CRITICALITY IN FUEL STORAGE AND HAELING l Criticality in the fuel storage and handling system shalI be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
RESPONSE
Geometrically safe configurations and fixed neutron absorbers (in the Ex-Vessel 61 storage tank) are employed to preclude criticality in new and spent fuel storage facilities and in fuel handling equipment. The appropriate safety measures and the design features necessary to meet this criterion are described 32 in Section 9.1 for the fuel storage and handling system. 1 O Amend. 61 i 3.1-79 Sept. 1981
----.-e. ~ , , - - - - . +.w., -m r-, n - ,w,,- ,~ ,.__.,--# ., _--...-n...w.--. ,p - , . -, - -.--,-y,,-,.-,~,w-. - - - . . , . . , - .
Criterion 55 MONITORING FUEL AND WASTE STORAGE Appropriate systems sbalI be provided in f uel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
RESPONSE
Monitoring systems are provided to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels. Appropriate locai clarms will be set of f and ennunciated in the control room to sarn personnel of potential safety problems. The number, sensitivities, ranges, and locations of the radiation detectors will be determined by requirements of the specific monitored process during normal and postulated abnormal (accident) conditions. All monitors will be designed se that saturation of detectors during a severe accident condition will not cause erroneously low readings. Monitoring during severe post accident conditions will be accomplished by the high range gamma area monitors discussed in Section 12.1.4, in conjunction with the sampling lines described in Section 11.4.2.2.1. Excessive gamma radiation levels will
- rip an alarm locally, and annunciate it in the control room thus permitting operators to take corrective action. Monitoring Instrumentation will be provided fcr the EVST and its associated areas for conditions that might result in a loss of the capability to remove decay heat and to detect excessive radiation levels. The RSB has radioactivity monitors above the EVST to detect accidental releases and to sound alarms. Monitoring instrumentation will also be provided for the FHC for conditions that might result in a loss of the capability to remove decay heat, and to detect excessive radiation levels.
Ternperature Instrumentation and sodium level sensing probes will monitor cooling capability of the EVST. Too high sodium temperatures, and too high or too low sodium levels will sound an alarm. Other monitors will be provided in 61 the two primary EVST cooling systems and the backup cooling loop. Sodium leak detectors will monitor the space between the storage tank vessel and the guard vessel. An argon gas activity monitor will be provided. An area monitor will measure the gamma radiation activity in the operating gallery of the FHC. 61l Instrumentation is provided for the EVST cooling system to monitor and alarm of f-normal conditions in both the sodium and NaK systems, including high temperature low flow, and external leak detection. The operating pressure of the NaK system is maintained higher than that of the sodium system. Leakage of NaK to sodium is monitored, and alarmed, by abnormal level indication in the NaK system expansion tank, in conjunction with the level in the EVST. Most of the gas processed in CAPS is the Inerted celi nitrogen which is periodically purged to control Itt oxygen content. Additional gas from air atmosphere cells may also be processed in CAPS If it contains radioactivity, in order to reduce the gas-processing load in CAPS the nitrogen and the air atmosphere in the cells are monitored for radioactivity. When a cell shows 32 high levels of radioactivity, the atmosphere can be passed 3.1 -80 Amend. 61 Sept. 1981
The planned sampling frequencies will ensure that significant f_e,') changes in the environmental radioactivity can be detected. The vectors , s_ which would first indicate increases in radioactivity are sampled most frequently. Those which are less ' effected by transient changes but show long-term accumulations are sampled less frequently. However, specific sampling dates are not crucial and adverse weather conditions or equipment failure on occasion prevent collection of specific samples. 2 The capability of the environmental monitorinn program to detect design-level releases from plant effluents is uncertain because of the
- insignificant quantities which will be released. The orogram will however provide the capability of detecting any significant buildup of radioactive material in the environment above and beyond that which is already present. Tliose vectors which are most sensitive to reconcentration of specific isotopes are sampled. If any increase in radioactivity levels is detected in these vectors, the program will be evaluated and broadened if deemed necessary.
2 From the data obtained from the radioanalytical and radiochemical analyses of the vectors sampled, dose estimates can be made for an individual or the populate living near the plant site. 32 O J i I ' (N (,,) Amend. 32 Dec.1976 3.1-83
TABLE 3.1-1 COMPONENTS WHICH COMPRISE THE REACTOR COOLANT BOUNDARY The Iist of Components or Parts of Componen+s which comprise the Reactor Coolant Boundary per the definitions of PSAh Section 3.1.2 is as follows: Primary Heat Transport System (PHTS) Piping and Apourtenances PHTS Pump Tank PHTS Pump Tank Drain Line Up To and including the Second Isolation Valve PHTS Pump Shaft Seal PHTS Pump instrument Penetrations PHTS Check Valve Body PiiTS Check Valve Freeze Vent PHTS Hot Leg Freeze Vent intermediate Heat Exchanler (IHX) Shell lHX Sheli Freeze Vent IHX Tube Bundle (including Tube Sheets) IHX Bellows Seal IHX Downcomer IHX Vent Line IHX Vent Line Freeze Vent IHX Cold Leg Pipe Draln Up To and IneludIng the Second Isolation Valve Reactor Vessel Closure Head 4l! Large Rotating Plug (LRP) , Intermediate Rotating Plug (IRP) Small Rotating Plug (SRP) LRP Riser Assembly IRP Riser Assembly SRP Riser Assembly in-Vessel Transfer Machine Port Plug 41l Rotating Guide Tube Control Rod Drive Mechanism Nozzle Extensions Control Rod Drive Mechanism Motortubes Upper internals Structure Jacking Mechanism Column Supports Upper Internals Structure Jacking Mechanism Seals Liquid Level Monitor Port Plugs 41 Maintenance Port Plugs 25 9 3 .1 -84 Amend. 62 Sept. 1981
TABLE 3.1-2 COMPONENTS WHICH COMPRISE THE INTERMEDIATE COOLANT BOUNDARY
. Intemediate Heat Exchanger
. Inlet nozzle
. Downcomer and bellows
. Lower tubesheet
. Hemispherical head
. Tubes (intermediate system is inside tubes)
. Upper tubesheet
. Intermediate channel
. Out?et nozzle
. Sta: tup vent nozzles
. Superheater (1)
. Sodium inlet nozzle
. Vessel shell
. Tubes (Intennediate system is outside tubes)
. Sodium bleed vent
. Sodium outlet nozzles (2)
. Evaporators - (2)
. Sodium inlet nozzle
. Vessel shell
. Tubes (Intermediate system is outside tubes)
. Sodium bleed vent
. Sodium outlet nozzle
. Intermediate Sodium Pump
. Inlet nozzle
. Discharge nozzle
. Pump tank
. Gas Equalization line
. Instrument penetrations
. Shaft Seal
. Intermediate Expansion Tank
. Expans % T...k Shell
. Nozzles (10) 25 i
l
. Amend. 32 Dec. 1976 3.1-85 - . _ - - , .~ . _ . _ . _ ... _ ._. __ _ _ ,_-- _ _ _ .__ ..__,.. .--_._ ___._ . __ _ .._ _ . _ _ .. ._. ._
TABLE 3.1-2 (Continued)
. Sodium Venturies (Loop 2 only)
. Instrumentation Bosses
. Intermediate Sodium Dump Valves
. Pump outlet
. Evaporators
. Superheater
. Expansion tank vent
. Piping
. 24" Hot leg
. 18" between steam generator to mixing tee
. 18" x 36" pump Inlet mixing tee
. 24" cold leg
. Reducers
. Elbows
. Tees
. 4" drai n l ines
. 36" pump suction piping
. 8" expansion tank return iIne
. 2" lHX vent i .e 61 l . 2" Expansicn tank, pump tank equalization line
. 6" Expansion tank vent iIne
. Miscellaneous Components (not actually parts of the IHTS)
. Sodium and Gas Rupture Discs
. Hydrogen Detector Valves i Plso included is revised Table 5.4-5 on page 5.4-33.
25 9 Amend. 61 3.1-86 Sept. 1981
I i l i
- O TABLE 3.2-1 SEISMIC CATEGORY l STRUCTURES
- 1. Containment Building 61l 18 l 2. Confinement structure 61 l 3. Reactor Service Area of the Reactor Service Building
- 4. Control Building
; 5. Stoon Generator BulIdIng
- 6. Diesel-Generator Building 43l 33l 7. Emcrgency Cooling Tower Basin-1
- 8. Diesel Fuel Storage Tank Foundation 33l9. Electric Manholes 43 32l 4
't , O 3.2-8 Amend. 61 Sept. 1981 n ,,r-. ,w - . ,--.,.-,~,n,- +...n,n.-._,n,_ . , . - . , . .nn,,.+--.-.-_,v_., ,,--,,-,-,-n--,n.,-,-n~
TABLE 3.2-2(Continued) PRELIMINARY LIST OF SEISMIC CATEGORY I MECHANICA SYSTEM gl COMP 0NENTS AND ASSIGNED SAFETY CLASSES 20 Safety Quality Components Class II) Group (") Location (2) Reactor Vessel & Primary Heat Transport System Reactor Vessel & Closure Head 1 A RCB A l20 Primary Sodium Pump 1 RCB Intermediate Heat Exchanger (IHX) 1 A RCB Piping 1 A RCB Reactor Guard Vessel 2 B RCB Pump and IHX Guard Vessels 2 B RCB Auxiliary Liquid Metal System Primary Sodium Overflow Tank 1 A RCB Primary Sodium Makeup Pumps 1 A RCB Overflow and Primary Sodium Makeup Piping and Valves (6) 1 A RCB Overflow Heat Exchanger 1 A RCB 26 Airblast Heat Exchangers 2 B RSB EVST Sodium and NaK Forced Convection Loop Components, Piping and Valves 2 B RSB EVST Natural Convection Sodium Loop Componeints and Piping 2 B RSB EVST Natural Convection NaK Loop Components, Valve, and Piping 3 C RSB 44 Natural Draft Heat Exchanger 3 C RSB h imary Loop Drain Line (6) 1 A RCB Primary Cold Traps (7) 3 C RCB In-Containment Pri Na Storage Vessel 3 C RCB Ex-Cont. Pri Na Storage Vessel 3 C SGB EVST N h NaK Drain Piping (8) 3 C RSB PHTS Drain Lines (9) 3 C RCB IHTS Na Processing System 3 C SGB 36 EVST Cold Trap 3 C RSB 47' Intermediate Heat Tra:. aport System IHTS Piping Extending from IHX 2 B RCB,IB,go SGB Intermediate Sodium Pumps 2 B SGB 2 R SGB Dump Valves 20 Expansion Tanks 2 B SGB IHTS Drain Lines (6) 2 B SGB 36 49 IHTS Drain Lines (9) 3 C SGB O Amend. 49 3.2-9 April 1979
TABLE 3.2-2 (Continued) PRELIMINARY LIST OF SEISMIC CATEGORY I MECHANICAL SYSTEM COMPONENTSANDASSIGNEDSAFETYCLASSES(3) Safety Quality Components Class (I) . Group (II) Location (2) Steam Generator System Evaporators 2 B SGB Superheaters 2 B SGB Steam Drums 3 C SGB Sodium-Water Reaction Pressure Relief 53 l Systems (internal to steam gen. b1dg.) 3 C SGB L 36 IHTS Na Dump Tank 3 C SGB SWRP Rupture Disk Assemblies (4) 2 B SGB 36 35 l S.G. Water and Steam Components, Piping and Valves 3 C SGB Steam Generator Auxiliary Heat Removal System Air-Cooled Condensers 3 C SGB '20 Auxiliary Feedwater Pumps (w/o motor drives) 3 C SGB Os, , Protected Water Storage Tank (PWST) 2 B SGB Connecting Piping & Valves (Extending from PWST to and including the First Valve) 2 B SGB Turbine Drive 3 C SC3 Connection Piping and Valves (except piping from PWST to and including the first valve) 3 C SGB 20 Containment Isolation Valves l (Within their associated fluid systems) 2 B RCB, IB i Containment Cleanup System Note (12) - RSB Containment Annulus Air Cooling System Note (12) - RSB Containment Annulus Filtration System 3 C RSB 36 Refueling System Ex-Vessel Storage Tank (EVST) 2 B RSB l 36 EVST Guard Vessel 3 C RSB 44 49 EVTM Containment Pressure Boundary 3 C RSB l 43 3.2-10 Amend. 53 Jan. 1980
TABLE 3.2-2 (Continued) PRELIMINARY LIST OF SEISMIC CATEGORY I MECHANICAL SYSTEM COMPONENTS AND ASSIGNED SAFETY CLASSES (3) Safety Quailty Components Class (l) Gre,up(I I) Location (2) Inert Gas Receiving and Prxessing System Primary Cover Gas Lines (Recycle Argon) 2 B RCB Equallz.- Line Between Reactor Vessel Primary Pump and Oveiflow Vessel 2 B RCB RAPS (Outside Containment) 3 C RSB 36 RAPS (Inside Containment) 3 C RC8 CAPS (Outside Containment) 3 C RSB Control Building Yentilation Fan 3 C CB Filters 3 C CB Air Conditioning Unit 3 C CB 3 C SGB,DGB, 20 Emergency Plant Service Water System (5) Emergency Cooling Tower Emergency Chilled Water System (5) 3 C SGB, CB, DGB l20 RSB, RCB 49 Auxiliary Mechanical Systems for Diesel Generatces .- 3 C DGB Fuel Oil Storage and Transfer System including: Diesel Fuel Oil Storage Tanks 3 C YARD Fuel Oil Transfer Pumps 3 C DGB Fuel Oil Day Tanks 3 C DGB Cooling Wcter System including: Water Expansion Tank 3 C DGB JacketA Cooling Heat Exchanger 3 C DGB Water Temperature Regulating Valve 3 C DGB Starting Air System including: Air Storage Tanks 3 C DGB Lubrication System including: Lubricating Oil Heat Exchanger 3 C DGB Lube Oil Filters and Strainers 3 C DGB g 3.2-10a Amend. 61 Sept. 1981
i
=
TABLE 3.2-2 (Continued) PRELIMINARY LIST OF SEISMIC CATEGORY l MECHANICAL SYSTEM COMPONENTS AND ASSIGNED SAFETY CLASSES (3) Safety Quality Components Class (1) Group (11) Location (2) 3 C RSB,.RCB 49l Recirculating Gas Cooling System (Subsystems Serving: Na makeup pump cold trap pipeways, Na . makeup pum;;, and vessels, EVS pump and cold trap, EVS pumps ' 45 and pipeways) ) ! 49l . Non-Sodium Fire Protection System SeismicalIy Qualifled Water Supply 3 C DGB l48 i 54 Piping, Valves, and Valves I&C RCB Penetration, Valves, and Valves l&C 2 B SGB, RCB , Standpipe System (Nuclear Island) Note (t0) SGB, CB, DGB RSB, RCB Piping and Valves Standpipe System Seismic Category I i Pumps Note (10) DGB 48 ! . Notest j (1) Safety Classes.are defined in Sections 3.2.2.1 through 3.2.2.3 (2) RCB - Reactor Containment Building fB - Intermodlate Bay of the SGB SGB - Steam Generator Building
- RSB - Reactor Service Building
- CB - Control Building i DGB - Diesel Generator Building (3) All components will be seismically qualified by analysis unless otherwise noted; motors are included with the mechanical components they d-Ive.
(4) The SWAPRS rupture disc assemblles wIlI be seismically'quali4Ied by analysis based on rupture data obtained during dynamic testirg. (5) Control panel attached to chillers will be quellfled by test. (6) Out to First isolation Valve (7) Within Dual isolation Valves (8) Downstream of Iso!ation Valve (9) Downstream of First isolaticn Valve (10) Non-Safcty Related, Seismic Category I. (11) Based on Regulatory Guide.1.26, as Interpreted for an LWBR
'(12) The containment annulus cooling system and containment cleanup system shall meet the safety class 3 requirements. However, these systems are provided for the mit!gation of an accident beyond the design basis. Therefore, they are not classified as SC-3.
3.2-10b Amend. 61 Sept. 1981
l f3 TABLE 3.2-3 V
- PRELIMINARY LIST OF SYSTEM COWONENTS CLASSIFIED 1E Bullding Electric Power System-Motor Control Centers
, Unit Substations 125V DC Distribution Panel Battery Charges inverters Vital Reg. Transformers Batteries 4.16 KV Switchgear Diesel Generator Sets Diesel Generator XFMR and Resistors
' Fue1 011 Transfer Pumps Na Pump Drive Breaker Trips Connectors and Terminations Cables Emergency Chilled Water System Isolation Valve Operators Pressure Control Valve Operators Pressure Relief Valves i Emergency Chilled Water Pump l { Coe. trol and instrumentation l Cat s
*The equipment contained in this list are generically identified. The specific, detailed listing of all Class IE equipment, along with the environmental qualification program to which they are subjected, is provided
- 61< in Reference 13 of PSAR Section 1.6.
3.2-11 Amend. 61 Sept. 1981
TABLE 3.2-3 (Continued) Heating, Ventilating, and Air Conditioning Svstem Unit Coolers Emergency Chiilers Supply Fans Exhaust Fans Air Handling Units Control Room Fiiter Fans Control Room Filter Units Control Room HVAC Monitors Control Roore Air Conditioning Units Aux. Building HX Air Conditioning Units Control and Instrumentation Cables Reactor Containment System R W TMBDB instrumentation Panels Containment Instrumentation Panels Cables Recirculating Gas Cooling System Local Control PaneIs Solenoid Yalves Moisture Switches Liquid Level Switches Cold Trap Flow Switches 61 O 3.2-11a Amend. 61 Sept. 1981
TABLE 3.2-3 (Continued) Cold Trap Temperature Switches Pump Temp. Switch Fan Motor , Cables
^
Nuclear Island General Purpose Maintenance Equipment rtem Containment isolation Valve Operators Cables Steam Generator Auxiliary Heat Removal System AuxIIIary Feedwater Pump Motors Auxiilary Feedwater FIon Meter Protected Air Cooled Coodenser Condensate Flow Meter Aux!Ilary Feedwater Valve Operators Vent Control Valve Operators L ter Storage Tank Fill Valve Operator AFW Turbine Steam Supply Valve Operator AFW Turbine Pressure Control Valve Operator Cables Steam Generator System Feedwater Velve Operators Superheater Outlet Valve Operators Steam Drum Drain Valve Operators 61 Cables
.i c O
Amend. 61 3.2-11b Sept. 1981
TABLE 3.2-3 (Continued) Reactor Heat Transport instrumentation System Reactor inlet Pressure Transmitter Reactor Vessel Na Level fransmitter Na Flow Sensor PRI IHX OutIet Na Temp Sensor Signal Conditioning PHTS Pony Motors Primary Pump Tachometcr Instrument Racks Instrumentation Reaction Products Dump Line Pressure Switches IHTS PM FIow Sensor Superheater and Steam Drum Vent Control Valve Operator Temperature / Leak Detection Instrumentation Steam FIow Meter Superheated Steam Temperature Sensor Superheeter Steam Pressure Sensor Feedwater' Flow Meter Feedwater Flow Meter Feedwater Tm p. Sensor Evap. Outlet Sodium Te p. Sensors 61 Cables O Amend. 61 3.2-11c Sept. 1981
TABLE 3.2-3 (Continued) O Emergency Plant Service Water Syste:n 4 Pump Motcrs 1 ECT Fans Makeup Pump Motors Temp. Control Valve Operators Temp. Transmitters Temp. Indicator ControlIers Pressure Dif*orential Switches Level Switches Cables Auxiliary Liquid Metal System Valve Operators OverfIow Thermocoupies 4 Local Paneis Pump Motors Control Cabinets Capacitor Cabinets j Transformer Cabinets 1 ! EVST Thermocouples Control Room PaneIs l 61 Cables n v Amend. 61 3.2-11d ~ Sept.'1981
- . - - . - . . _ . . - . - - . . - - . . . - - - . . - - - . - . . - . - . . . . . - . - . . - - . - . . . . . . . - , , . . . - ~ _ . -
l l TABLE 3.2-3 (Continued) Inert Gas Receiving and Processing Systems l Containment isolation Valve Gperators Expansion rank Equalization Valve Operator Cables ! Plant Control System Main Control Panel SCRAM Breaker Cubicle Cables l Reactor and Vessel Instrumentation System
- Reactor Ccolant Operating Level Instrumentation i
Cables Flux Monitoring System l Flux Monitoring instrumentation ! Cabinets i
- Instrument Drawers l
Cables i PIant Protection System Containment isolation System Cabinets Reactor Shutdown System Cabinets Plant Protection System Cabinets 61 Cables l ( l 3.2-11e Amend. 61 l Sept. 1981 1
i 1 O l i TABLE 3.2-4 HAS BEEN DELETED 61 . f O l
-O 3.2-12 (Next page is 3.2-14) Amend. 61 Sept. 1981
TABLE 3.2-5 (Continued) s PRELIMINARY LIST OF ASE CODE CLASSIFICATIONS FOR SEISMIC CATEGORY l MECHANICAL SYSTEM COMPONENTS Component Code / Code Class (1) Location (2) Emergency Plant Service Water System ASE-I l l/3 SGB,DGB Emergency Chilled Water System ASE-I l l/3 SGB,CB,DGB, RSB,RCB Normal Chilled Water System ASME-Ill/3 , RCB 61l15 Auxiliary Mechanical Systems for Dies t ASE-I l l/3 DGB , Generators Fuel Oil Storage and Transfer System including: Diesel Fuel Oil Storage Tanks ASE-l i l/3 YARD Fuel Oil Transfer Pumps ASE-l l l/3 DGB Fue1 01l Day Tanks ASE-1 I I/3 DGB Cooling Water System including: Water Expansion Tank ASME-1II/3 DGB O Jacket Coo 1Ing Heat Exchanger Water Temperature Regulating Valve ASE-1 I I/3 ASE-I l l/3 DGB DGB Starting Air System including: Air Storage Tanks ASE-I l l/3 DGB Lubrication System including: Lubricating Oil Heat Exchanger ASE-I l l/3 DGB 61 Lube 01i Fiiters and Strainers ASE-l l I/3 DGB Control Room Heating, VentiIating, and ASE-l l I/3 C8 32 Air Condition System isolation Valves 48 Non-Sodium Fire Protection System SGB,CB,DGB 48 Seismically Qualifled Water Supply ASE-l i 1/3 DGB c Piping, Valves, and Valves l&C RCB Penetration, Valves, and Valves l&C ASME-Ill/2 SGB,RCB ! Standpipe System (Nuclear Island) Note (9) RSB,RCB 54 Piping and Valves 48 48 3.2-15a Amend. 61 Sept. 1981
Tf8LE 3.2-5-(Continued) PREllMINARY LIST OF ASME CODE CLASSIFICATIONS FOR SEISMIC CATEGORY l MECHANICAL SYSTEM COMPONENTS Component Code / Code Class (l) Location (2) Standpipe Sysicvn Seismic Category i Note (9) DGB l48 48l Pumps Notes: (1) including applicable code cases. (2) RCB - Reactor Containment Building IB - Intermediate Bay of the SGB l SGB - Steam Generator Area of the RSB 61 i RSB - Reactor Service Area of the RSB CB - Control Building DGB - Diesel Generator Building 61 l (3) Only piping from containment isolation valves to the filter intake; filters and discharge ductwork per Reg. Guide 1.52. (4) System wilI meet the requirements of Reg. Guide 1.52 (5) Out to First isolation Valve (6) Within Dual isolation Valves (7) Downstream of isolation Yalve (8) Downstream of First isolation Valve (9) Non-Safety Related, Seismic Category I l l l l l 1 i 9 3.2-15b Amend. 61 Sept. 1981 L
3.4 WATER LEVEL (FLOOD) DESIGN O The flood elevations used in the design of the Seismic Category I structures are discussed in Section 2.4. 3.4.1 Flood Protection
- 1. Seismic Category I systems and equipment requiring protection against flooding are delineated in Table 3.4-1.
- 2. By design consideration, there wilI be no exterior accesses or entrances below the plant grade at El. 815, to all Seismic Category i Structures, thereby completely eliminating the potential ingress of groundwater or flood water. 6
- 3. All Seismic Category I systems and equipment will be located on floors above maximum flood level (MFL) for flood protection, or will be protected by the following measures.
Water stops will be provided in all construction joints of foundations and walis below grade level. Watarproofing will be provided on the outside faces of all exterior walls below grade and on the underside of all foundation mats. Watertight doors and equipment hatches will be provided at entries to O noncritical areas when such accesses are required and located below the MFL. There are no exterior penetrations into Reactor Containment Structure below grade level; all penetrations into containment below the maximum flood or groundwater level are routed through the Interior of other Seistr.ic Category I structures which are also designed as watertight reinforced concrete structures. Exterior penetrations into other Category I structures below grade. level will be designed to be watertight by providing steel pipe sleeves embedded in the concrete walls. Each penetration assembly will be continuously seal-welded to the pipe sleeve at each end directly or through steel seal plates attached at each end of the pipe sleeve depending upon the number and configuration of the penetrations. Wherever seismic differential movement and/or thermal growth are a design conc deration, sealed flexible joints will be provided at such connections. 6 61l There are no exterior accesses or entrances to any of the Seismic Category I structures below the plant grade, and no exterior penetrations into Reactor Containment Building ard other Category I 61I structures below the Maximum Flood Level. Typical details for Installation of waterstop and waterproofing membrane are shown in Figure 3.4-1. V 3.4-1 Amend. 61 Sept. 1981
- 4. Flood warning systems will be developed and installed in con-junction with TVA flood control network. Administrative pro-cedures will be adopted to ensure that all watertight doors, normally open for internal access, will be closed and locked in the event of a flood warning affecting the plant.
3.4.2 Analysis Procedures Seismic Category I structures and component parts will be designed for the hydrostatic forces due to the MFL. O Amend. 6 & 3.4-ia Oct. 1975 W
I 1 A triangular hydrostatic pressure diagram will be used in the wall design and a c)s (
'" unif orm hydrostatic load application will be used in the base and cover slab designs. Hydrostatic pressures on structures or their comoonents will be treated as if they were dead loads. The ef fect of hydrosiatic pressures on structures will be considered in accordance with the load f actors and loading combinations stated in Section 3.8.
Uplif t f orces during the MFL will be accounted for in the design of structures. Suf ficient margin of safety will be provided for seismic Category I structures subjected to the ef fect of buoyancy during this maximum flood condition. The wind waves producing the dynamic water forces are discussed in Section 2.4.3.1. A conservatively high wind velocity of 40 miles per hour overland from the cost adverse direction has been applied at the time of the PMF to conf orm with the Intent of Regu! ate: y Guide 1.59. Wind waves were computed using procedures of the Corps of Engineers (Ref.1). Furthermore, it has been established that, by analyses, an operational basis earthquake (OBE), imposed concurrently with the one-half PMF resulting in postulated Norris Dam failure, would be the controlling design condition. This condition, as discussed in Section 2.4.4 plus the effect of wave runup, would produce the maximum plant flood level as stipulated by the Regulatory Guide 1.59. The peak flow at the plant site for the controlling, OBE-one-half PMF, Norris Dam failure situation would be 941,000 cfs. The maximum crest still reservoir level will probably reach an elevation of 804.3 feet at Mlle 18 and El. 798.2 O- feet at Mile 16. Combining a coincidental 40-mile-per-hour overland wind in a critical direction, the runup would be 3.6 feet on a 3:1 smooth slope and 4.9 feet on a vertical wall. Based on the conditions described above and by the use of the Minikin Metnod of computing wave f orces as detailed in the Corps of Engineer's Technical Report No. 4, " Shore Protection Planning and Design", (Ref. 2), the maximum wave f orce exerted on the plant structures is computed to be relatively insignificant. With the plant grade established at elevation 815 thereby providing a freeboard of at least 6 feet in excess of the maximum design flord, any additional l hydrodynamic force on the walls of the Seismic Category I structures from wave action will be eliminated. It is concluded from the above that the maximum differential pressure resulting from the application of the maximum design flood on adjacent walls of the 6d Nuclear Island Buildings will be minimal and will have virtually no influence on the f actors of safety against sliding and overturning inherently established 6d by the approximately 100' of embedment below grade. 1 l ('~\
\_-) .
3.4-2 Amend. 61 Sept. 1981
l 9 References to Section 3.4
- 1. U. S. Army Corps of Engineers, " Computation of Freeboard Allcni6nces for Waves in Reservoirs", Engineering Technical Letter No. 1110-7-8, August 1966.
- 2. U. S. Army Corps of Engineers, " Shore Protection, Planning and Design", Coastal Engineering Research Center Technical Report No. 4, Third Edition,1966.
O l l 9 3.4-2a
l l TABLE 3.4-1 Os PRELIMINARY LIST OF SEISMIC CATEGORY I SYSTEMS AND COMP 0NENTS REQUIRING FLOOD PROTECTION Component LocationIII Intermediate Heat Transport System IHTS Piping Extending from IHX IB & SGB Intermediate Sodium Pumps SGB Steam Generator System Evaporators SGB Superheaters SGB Steam Generator Auxiliary Heat Removal System Auxiliary Feedwater Pumps SGB Connecting Piping & Valves (Extending from SGB protected water storage tank to and including the second valve) Refueling System Ex-Vessel Storage Tank RSB Ex-Vessel Storage Tank Gu:rd Vessel
- RSB Inter Gas Receiving & Processing System (IGRPS)
Primary Cover Gas System RSB & SGB Emergency Plant Service Water Sy, stem (EPSW) RSB, IB, DGB & CB Emergency Chilled Water System DGB Portions of Direct. Heat. Removal Service RSB Auxiliary Mechanical Systems for Diesel DGB Generators 4.16 KV Auxiliary Power System 4.16 KV Class 1E Switchgear DGB 36 4.16 KV/480V Class IE Unit Substation DGB Amend. 36 s 3.4-3 March 1977
TABLE 3.4-1 (Continued) C:enonent Location (1) 480 Y Motor Control DGB, SGB, RSB 120 V AC Vital Power System inverter System Components 120 V AC Vital CB instrument Power Boards CB 125 V DC Pows. Systans 125 V DC Batteries CB 125 V DC Vital Battery Charge CB 125 V DC Battery Distribution Boards CB 36 250 Y DC Diverse Power System 250 V DC Battery CB 250 Y DC Battery Charger CB 37 250 V DC/480 V AC Inverter CB 61l lB = Intermediate Bay of SGB SGB = Steam Generator Building RSB = Reactor Service Building DGB = Diesel Generator Buli *!ng CB = Control Building l O 3.4-4 Amend. 61 Sept. 1981
_ . . - _ _ ._ _ _ _ _ . _ _ _ . . . . _ . _ _ _ . . - _ ~ .- .. ._ _ _ _. . _ _ . _ _ _ . _ _ _ _ _ _ _ _ - _ 4 I l' l l t i i il i i. .i
- I 1
TABLE 3.4-2 1 1 1
- I j 61 HAS BEEN DELETED I
I i i 1 3.4-5 Amend 61 Sept. 1981 f W
- w. e- ~ - y--, w m e, w ww . - . - __ w _ _ _ - , _m_ --wv*
e CATEGORY I STRUCTURES EXTERIOR FACE OF WALL q*
~.
.s
. Y ..?
i
- C04TINUuus y, WATER STOP
-TOP OF FOUNDATICN
' ., - J.- MAT
.' /
m- D* % /
, ~'-. . . * -
L ' ,%(**.
.' .., i
- Ep-
' NUCLEAR ISLAND bulLDINGS WATERPROOFING FOUNDATION M AT MEMBRANE 7 3" MEMBRAME &
W PROTECTION SLAB
$.; '. . -4 '
$$*l+*:;
.x. .
. . . .....q e Y
. a w. :
. . x 3" MUD MAT
\
B" MIN. CRUSHED ROCK DRAIN AGE LAYER ROCK SURFACE 8" APPRo%. LEVELING MAT FIGURE 3.4-1. TYPLCAL FLOOD (GROUhlDWATER) PRorECTtoN DD. TAIL O Amend. 61 3.4-6 Sept. 1981
Missile barriers will be provided only at locations where necessary for protecting vital equipment af fected. Typical details of such missile barriers O are shown in Figure 3.5-3. 3.5.2.*, 1 Turbine Fallure Missiles The selected orientation of tho turbine generator shaft with respect to the seismic Category I structures is such that the probability of damage to these structures from turbine failure missiler is essentially eliminated. The analysis of turbine failure missiles is indicated in Section 10.2.3. I 3.5.2.1.2 Primarv Pumn Internals Failure Missiles in the event of primary pump overspeed and if it is postulated that the rotating components of the pump internals could generate missiles this missile source will be extremely unlikely since proper design and inspection will prevent its occurrence. Nevertheless, pump Internal missiles will be postulated and the necessary analysis conducted to assure that the guard vessel surrounding the pump cannot be adversely impacted. The design objective will be to assure that the missiles are contained within the pump hydraulics and pump tank (see Table 3.5-1). 3.5.2.1.3 Intermediate Heat Transnort System (IHTS) Pumn Internals Failure Missiles it is postulated that the rotating components of the Intermediate pump Internals and the motor drive could generate missiles. These missiles are extremely unlikely, since the proper design and inspection should prevent their occurrence. In addition, the pump and drive are not expected to be subject to significant overspeed. Nevertheless, pump and drive internal missiles will be postulated and necessary analysis performed to assura they are contained within the pump and the motor drive casings. 3.5.2.1.4 Rotating Cnmnonent Missile Selection from the Steam Generator Svstem (SGS) Missiles associated with the recirculation pump are postulated to occur as a result of pump overspeed caused by a pipe break in the recirculation system, t Amend. 61 3.5-5 Sept. 1981 l
It is assumed that the resulting high-velocity fluid will cau m the pump to reach an overspeed condition which will result in rotating component failures. Based on normal design m&rgin used in LWR pumps (
Reference:
GED0-10677, " Analysis of Recirculation Pump Overspeed in a Typical Gereral Electric BWR", October 1972), the impeller failure is expected to occur at less than approximately 300% rated speed. It is expected that the resulting fragments would be contained within the pump casing and will travel through the straight run of piping, impacting the pipe at the first bend. The energy of the fragments will be established following the selection of the recirculation pump. 3.5.2.1.5 Rotating Component Missile from Steam Generator Auxiliary Heat Removal System (SGAHRS) The auxiliary FW pumps and the turbine drive are not expected to generate high-energy missiles, since they are not expected to overspeed. The motor drive are constant-speed synchronous motors and the turbine will not overspeed even in the event the control valve fails to fully open, since the pressure-reducing station and the inlet nozzle designs result in nearly 100% rated steam flow for this condition. 3.5.2.2 Pressurized Component Failure Missiles 3.5.2.2.1 Missile Selection From intermediate Heat Transport System (IHTS) The IHTS is composed of three independent and physically s(parated low pressure (approximately within a pressure range from 98 psi to 210 psi) coolant loops. Since the pressure is low, the energy state of the contained fluid is correspondingly 100, no potential sources of high-energy missiles have been identified. 3.5.2.2.2 Missile Selection From Steam Generator Auxiliary Heat Removal System (SGAHRS)_ The selection of potential missiles and their characteristics is provided in Table 3.5-2. This listing of missiles represent an envelope in that the most energetic of the potential missiles in a given cell or area was chosen. 3.5.2.2.3 Missile Selection from the Steam Generator System (SGS) Potential missiles associated with the SGS are cissiles generated by the recirculation pump as a result of pump overspeed, steam generator module isolation valve, and safety valves. O 3.5-6
i
- 3. Class III missiles - Resulting from contained fluid energy, jet-propelled missiles
- a. Equation for velocity (Ref. 1):
1- - In 1 -
=K) rg + tans V. -
V. K K) = 1- - 1 r. 1 - + where Am PS b " AoMr (tans) V = missile velocity at distance X, fps 4 Vf = jet velocity at the throat, fps r = radius of throat, ft x = distance traveled,'ft 8 = angle of jet expansion, degrees from normal
^ V = initial velocity of missile, fps 2 4 j Pf = density of fluid jet, lb-sec /ft 2
A = missile area under pressure, throat area, ft 2 A,= cross sectional area of missile, ft 2 M = mass of missile, Ib-sec /ft
- b. Equation for kinetic energy:
2 KE = 1/2 MV These equations provide a conservative estimate of missile energy since no consideration is given for energy losses due to friction or air resistance. O 3.5-9 l l
3.5.4 Barrier Deslon Procedures Missile resistant barriers and structures wil; be designed to witnstand and absorb missile impact loads without being fully perforated in order to prevent damage to protected components. In addition, the overalI structural response will be evaluated to assure the structural Integrity due to missile impact loads. For concrete missile barriers, the possibility of generation of secondary missiles due to spaliing or scabbing will also be taken into consideration so that protectivo measures can be provioed. The design procedures are described below. 3.5.4.1 Penetration into Concrete Tarcet Structures To arrive at a method for computing the penetration into concrete walls, f ormu l as rev i ewed i n OPld.-NS I C-22 ( Ref . 1 ) were studied. Four equations were studied in ORNI.-NSIC-22. Two of these, the Army Corps of Engineers formula and the National Defense Research Committee formula, do not apply for impact velocities under 500 ft/sec. The remaining two equations are the modified 15 Petry formula and the Ballistic Research Laboratory formula. Those two formulas were compared by determining the depths of penotration for a 6-inch-diameter missile of 100 pounds and a 16-inch-diameter missile of 2,500 pounds 6l with velocities in the range of 0 to 500 f t/sec. As seen in Figuros 3.5-1 and 3.5-2, the Petry formula is the more conservativo for velocities greater than 6l 150 and 200 f t/sec. respectively. Therefore, the depth of a concrete wall or slab to which a missile can penetrate is estimated by use of the modified Petry formula: D' = KApV' [1 + e-4(a-2)] where D' = depth of penetration (f t.) l 61l K, a material constant = 2.76 x 10-3 (ft2 - sec.) Ib missile welaht J Ap , maximum cross-sect 1onal area (psf) V = Impact velocity (ft./sec.) l V' = log 10 (1 + V
) (ft./sec.)
l #'"""" T 61l T p = wall or slab thickness (ft.) l For design purpose, all Category I concrete structures will satisfy the requirement; Tp 2.2D' O 3.5-10 Amend. 61 y Sept. 1981 5 4
For evaluation of the required strain energy to stop the target (or missile - target combination), it is necessary to determine the Resistance-Displacement function which depends upon the physical fv) configuration and material properties of the target structure as well as those of the missile. The general concept of applyingcthe resistance-displacement. functions to the structural responses is illustrated by Figure 3.5-4. Under the rapid rates of strain that occur in structural elements subjected to impulsive or impactive loads, both the concrete and its rein- l forcement exhibit higher strengths than in the case of statically loaded elements. These increased allowable stresses or dynamic strengths are used to compute the element's dynamic resistance to the applied impulsive or impactive loads. In the design of such structures or missile barriers, the dynamic capacity of the structural elements is based on material dynamic strength properties which are to static strength obtained values. Thatby:sapplying f (a Dynamic
= DIF) Increase f # where Factor (DIF) f " = Dynamic Strength value,fstat=statedstrengt@talue.
The Dynamic Increase Factor depends upon the rate of strain of the element, increasing as the strain rate increases. The Dynamic Increase Factors for various materials are given as follows: DYNAMIC INCREASE FACTOR (DIF)
- 1. Reinforced Concrete A. CONCRETE DIF Compression 1.25 Diagonal Tension & Direct Shear (Punch Cut) 1.0 Bond 1.0 B. REINFORCING STEEL Tension 1.2 Compression 1.2 Diagonal Tension & Direct Shear (Stirrups) 1.0
- 2. Structural Steel Flexure & Tension 1.2 Compression 1.2 Shear 1.0 The above Dynamic Increase Factors are based on the publication " Structures to Resist the Effects of Accidental Explosions" TM5-1300, Department of the Army, Washington, D.C., June, 1969.
Similar recommendations are included in references 9,10 and 11. 25 3.5-13 Amend. 25 Aug. 1976
6 3.5.4.6 Conversion of Missile imoact Load into Eculvalent Static Deslan load The mit,ille of f ectli e structural loads are determined, based on a paper by R. A. Williamson and '.. R. Alvy, (Ref. 7). Case 1. Missile Penetratino Structural Element 6 The bas!c assumption made f or ihls case is that the force of impact "F, e and its velocity "V" both reduce Iinscrly to zero as the fragment oonotrates the structural elanent. The total kirotic energy of the missile is expended while it travels a distance "D'", which is equal to the penetration obtained by use of the Modified Petry Formula. Thus: F 1 D' = w 2 2 F , . Wy2 and 6 gDT t1 = 20' y Where: 1 6l Fr = force of impact (Ibs) V = velocity of missile just before Impact (ft./sec.) W = weigh + cf missile (Ibs) O g = acceleration due to gravity (ft./sec.2) t1 = duration of impulse or impact force (seconds) 6l The value "Fi " is generally too conservative to be used for practical design. N. M. Newmark (Ref. 8.) developed a factor for reducing the value "F g" expressed by the equation: i I=K Where F = equivalent static load (Ib.),
'T 0.5
- then K = (2u-1)o.5 T + 1 6 HL1
# 0. T t
i l O Amend. 61 3.5-13a Sept. 1981
3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED [] V RUPTURE OF PIPING CRBRP systems and components important to safety will be appropriately protected against dynamic offects, including the offects of missiles, pipe whipping, and discharging fluids that may result from equipment failures or other events. 3.6.1 . Systems In Which Ploe Breaks are Postulated 3.6.1.1 Systems inside Containment Spontaneous ruptures of heat transport system (HTS) and auxiliary sodium piping inside containment are not considered credible because of the high quality of the piping, operating temperature and pressure conditions for this piping, the inert environment provided for it, and the capability of the leak detection system to provide an early warning of any breach in the piping boundary. As a result, massive f ailures of this sodium HTS piping have not been included in the design bases for CRBRP syst es inside containment. A four inch crack, which leads to 8 gpm PHTS leak rates, has been chosen as the design basis. A description of the analyses and test results to support this position are presented in the Piping Integrity Status Report (Reference 2, Section 1.6). A similar detailed evaluation has not been performed for the sodium piping in the auxiliary IIquid metal systems. The piping parameters (e.g. t/D ratio), service conditions (temperature, pressure, duty cycle), monitoring and (N ( ,) Inservice inspection techniques are similar to those for the heat transport system. Based on this, the maximum credible crack length is not larger than the 4 Inch crack specified for the PHTS. The cell liner design is proceeding on the basis of containing the 4 Inch crack with pressures and temperatures that are characteristic of the associated sodium system. The chilled water system piping in-containment is moderate energy piping according to the definition in B7 APCSB 3-1: Pipa leaks are postulated in the piping and mitigated by the features discussed in Section 9.7.3. 3.6.1.2 . Systems outside containment For systems outside of containrcent, the intent of the guidelines in Appendix C to Branch Technical Position APCSB 3-1 (J. F. O' Leary letter 7/12/73) wilI be used as a basis for leak evaluations. Where seamless pipe is used longitudinal 31 breaks will not be posculated if all stresses are below 0.8 (!.2 Sh + SA)* Separation and Isolati(n of equipment by arrangement as shown in the figures in Section 1.2, atmosphere separation as described in Section 3A, and equipment enclosure are provided to protect safety systems and components required to shutdown the reactor safely. The high and moderate energy piping systems outside containment are listed in Tables 3.6-1 and 3.6-2, with the PSAR Section which discusses the system and the potential results of pipe leaks. Chapter 27 15.0 also contains analyses of postulated ploe leaks. f Amend. 61 3.6-1 Sept. 1981
1 3.6.1.2.1 Water / Steam Systems The definitions con' .s..ed in Appendix A to the Branch Technical Positions are considered to e applicable to the water and steam piping outside containment. Thz following is a tabulation of the nigh and moderate energy systems together with a discussion of the design features which protect the essa .1 systems necessary to shut the reactor down and to mitigate the consewences of a postulated pipe break. 3.6.1.2.1.1 Steam Generator Auxiliary Heat Removal System (SGAHRS) The elements of this system are described in Section 5.6.1. The piping from the auxiliary feedwater isolation valves to the steam drum, the piping between the steam drum and the Protected Air Cooled Condenser (PACC), and the piping from the steam drum to the Auxiliary Feed-water Pump (AFWP) turbine drive isolation valve are high pressure as defined in Appendix A to the BTP and will be evaluated for postulated ruptures. Because these pipes are located outside of the cells containing the major auxiliary feedwater components, a continued supply of auxiliary feedwater will be available after a postulated rupture. Separation of the HTS loops and their respective cells and Steam Generator Building flooding protecti . 45 l provisions (Section 7.6.5) prevent propagation of a pipe rupture event to adjacent loops and thus the essential systems to mitigate the consequences of the rupture are maintained. The piping runs from the AFWP to the AFW isolation valves and from the turbine drive steam supply isolation valve to the turbine drive are low-pressure and low-temperature lines during normal plant operation. Both lines are subjected to high-pressure during AFW operating periods and the turbine supply line is also subjected to high temperature conditions during the time the turbine is operational. However, the SGAHRS operating time is anticipated to be less than 2% of the plant operating time since the auxiliary feedwater portion of SGAHRS will not be utilized unless the Normal Heat Rejection ystem (main condenser) or Feedwater Supply System has been lost. Therefore, this piping will be evaluated as moderate energy piping for through the wall cracks during normal plant operations. The primary concern for a crack in this piping is for protection of the major auxiliary feedwater components from the accumulated water that has leaked. The major components are elevated to provide this protection and to prevent the propagation of event consequences. Other essential systems for reactor shutdown are not impacted by low temperature and low pressure leaks from this piping. No piping breaks will be postulated for the low pressure and low 27 temperature piping run from the Protected Water Storage Tank to the AFWP. 3.6-la Amend. 45 July 1978 h
3.6.5.4 Provisions for Separation of Piping and Other Redundant Features, In the CRBRP, large pipe breaks are postulated for the steam / water lines in the Steam Generator System (SGS) and the Steam Generator Auxiliary Heat Removal System (SGAHRS). For the CRBRP design, all the features of each of the three loops in these two systems are located in isolated com-partments and separated from the other two loops. The compartment walls, in c abination with provisions of pipe whip restraints where necessary, will be designed to accommodate pipe break effects and limit damage propagation. 3.6.5.5 Pipe Restraints and Locations Number and locations of all pipe whip restraints will be sunnarized and tabulated as the design is developed. Determination of pipe restraints locations will be made based upon the criteria described in Section 3.6.5.1. References
- 1. A. H. Shapiro, "Tae Dynamics and Thermodynamics of Compressible Fluid Flow," Ronald Press Co. , New York,1953.
- 2. F. J. Moody, " Prediction of Blowdown and Jet Thrust Forces," ASME Paper 69HT-31, August 6,1969.
- 3. L. D. Steinert, "PDA - P'pe Dynamic Analysis Program for Pipe Rupture Movement", NEDE-10813A (Class II), dated Fee 'ary.1976.
- 4. General Electric Co. Standard Safety Analysis Report (GESSAR) Section 3.6. 34
- 5. Nuclear Services Corporation Report No. GEN-02-02 " Final Report Pipe Rupture Analysis of Recirculation System for 1969 Standard Plant Design", May 1973.
1
- 6. ASME Publication 74-NE-1, " Plastic Deformation of Piping Due to '
Whip Loading", T. L.Gerber. l l l l 3.6-9 Amend. 34 Feb. 1977
Table 3.6-1 High Energv Ploing Systems (Outside Containment) Svstem PSAR Section Steam Generator System 5.5 Main Steam, Condensate 10 and Feedwater systems Steam Generator Auxillary 5.6.1 Heat Removal Systems (piping from aux. feed-water Isolation valves to l 61 the steam drum, piping between the steam drum and the Protected Air ! Cooled Condenser, and I the piping from the steam l 61I drum to the auxillary l feed pump turbine drive I isolation valve) l l l l l l l i i i O 3.6-9a Amend. 61 Sept. 1981 1 L
3.7.1.6 Soll structure Interaction The seismic analysis of Category I structures buried or founded on soll will be conducted using finite element techniques. The analysis will account for the strain dependent properties of the soII. The mathematical models w!il represent the struc*ure and the supporting foundation materials, soll and rock 46 down to the elevation of the fcfandation of the major seismic Category i structures (Nuclear Island). The input motions shall be applied at the surf ace level (finished grade) on an assumed rock outcrop and shall consist of the rock motions used in ti e analysis of the Nuclear Island. No credit shall be given for the soli cover or overburden in the deconvolution. No point in the final response spectra at the free-field foundation level shall f all below 60% of the design response spectra. The same limitation applies to the response spectra calculated at the elevation of the foundation in the soil-structure interaction system. The vibration motion obtained at the finished grade level should givs 6,1, resonse spectra that envelop the design response spectra. The analysis will produce response spectra at points of Interest and the forces acting on the 46 structures. This response spectra will be widened by i 15% in frequency. 33 The major Category I structures are founded on competent rock, with an average shear wave velocity of more than 4000 ft/sec. For these structures, the soll 7 (rock) structure interaction will be analyzed as follows: O l i 3.7 'x Amend. 61 Sept. 1981
3 The characteristics and properties of the rock are described l .- Section 2.5.4.4. The material underlying the foundation consists of layers of siltstone and limestone dipping at an angle of approximately 300 with the horizontal, from West to East. The Nuclear Island is located directly upon the siltstone. The geological profile shows an upper zone of weathered rock above the continuous (sound) rock. The foundation level (Elevation 715.0 ft) is about 20 to 35 ft. below the top of the continuous rock and 100 ft. below finished grade (elevation 815.0 ft). The excavation for the foundation consists predominantly of vertical or near vertical cuts, except at the West side where the excavation profile consists of a relatively small vertical cut at the lower elevations with the rock and soil above slope cut on a 2:1 slope along the bedding plane. The space between the side of the excavation and the plant will be backfilled with lean conciate from base level to the top of the vertical cut. Compacted Class A fill will extend from the top of the fill concrete to grade level. (Figures 3.7-19,20) The rock-structure interaction has been represented in the analytical models for the seismic analysis by equivalent massless foundation springs and dashpots. Due to the inclined configuration of the rock strata, half space tneory is not directly applicable to this site. To calculate the foundation springs a static finite element approach was used. This type of approach has been suggested by Whitman (Ref. 8) and has been used to calculate foundation springs for embedded structures (Ref. 9,10). Since the Reactor Containment, Confinement, Steam Generator, Diesel Generator, Control and Reactor Service buildings are supported on a conmon foundation mat only one set of rock-springs and dashpots will characterize the rock-structure interaction of these structures. The set consists of springs and dampers for translation in the North-South, East-West and Vertical directions and for rotation about axes in the same three directions. The translational spring constants were cal-culated as the total force required to impose a unit displacement on the foundation mat and vertical boundary of the embedded structures. The rotational spring constants were calculated as the total moment required to give the foundation mat and vertical boundary of the embedded structures a unit rotation. In imposing the unit displacements or rotations the foundation mat and vertical boundaries of the embedded structure were assumed to be rigid. The computer program STARDYNE was used for the finite element calculations. Three basic models were used:
- 1) Model A (North-South Direction)
This is a plane strain model based on a section through a North-South plane along the center of the Containment (Figure 3.7-21) and was used to calculate the horizontal translational and rocking springs for the North-South 40 direction an12 in, nominal diameter) SmalI Diemeter Piping Systems 1.0 2.0 (112 In. nominal diameter) Welded Steel Structures 2.0 4.0 Bolted or Riveted Steel 4.0 7.0 Structures Prestressed Concrete Structures 2.0 5.0 Reinforced Concrete Structures 4.0 7.0 9 7 (1) Reduced damping values will be used when combined stresses are considerably below 1/2 yield for the CBE and yloid for the SSE. For active components, as defined in NRC Regulatory Guide 1.48, the OBE 46 damping values should also be used for the SSE. O 3.7-24 Amend. 61 Sept. 1981
I O O O i PLAN
; _ N PL ANT I
f., j
; A CONTROL DIESEL i BUILDING (hggg SGB INTERMEDIATE BAY SGB- SGB REACTOR STEAM AUX. , SERVICE REACTOR GEN.
w BLDG. CONTAINMEPIT BAY BAY N BUILDING h Hi +
/
CONFINEMENT / i SGB MAINTENANCE
+ A BAY WF
%@Q.
- ~, Figure 3.7-19
$~
REACTOR CONT. BLOG. 1000 - SGB ,c--CONFINEMENT INTER.V .EDIATE- I
\
BAY , N CONTROL-BLDG. 900 _
\ l ,
EXISTING-, GRADE JPERATING FL. EL. 816' DESIGN PLANT FLOOD
"~
GRADE] -. , LEVEL
. EL.815' t '
# P 800- /-
# RIVER
'N s
-W ' _J
_w' w s WATER sV LEVEL l l EL 741 i w .
- -] [ - W - - -[; - W - ; p'-
m sl ~ BOT. OF WEATHERINGd 700- %. ,
' 3 ONE LTSTONE LI .ESTONE (s
600-- N \ 9' N _ i SECTION A-A i co ! $p$
"igure 3.7-20 k
# 9 e
i : i i i O . i DSL: Dynamic system loadings associated with pipe leak / rupture-loads. Since there are large failures of primary sodium , system components, these loadings are not applicable to
- these components. ' ,
l Pj: Internal Design Pressure (or Transient Pressure Loads). r P: e External Design Pressure To: Thermal effects and loads during startup, normal operating or shutdown conditions, based on the most critical trans-ients or steady-state condition. TS Thermal loads under thermal conditions generated by accidents (such as major sodium fires) and including To, - as appropriate. R: Accidents loads due to Extremely Unlikely Fa0lts (such as interfacing loads fro.n inner cell concrete structure). Transients: Dynamic loads, thennal transients and variations in pressure . . loads due to tnnsients associated with the particular ! plant condition. l To:---------Thermal effects . and loads during normal operating or. . i shutdown conditions, based on the most critical transient or steady state condition. Ro :---------Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition. E:---------Loads generated by the Operating Basis Earthquake (0B) . L El ---------Loads generated by the Safe Shutdown Earthquake (SSE) Pa ---------Pressure equivalent static load within or across a compart-ment ard/or building, generated by the postulated accident, and including an appropriate dynamic load factor to account' I for the dynamic nature of the load. Ta ---------Thermal loads under thermal conditions generated by the postulated accident and including T,. 46 Amend. 46 Aug. 1978 i l 3.7-A.lla * . _ ,.; - .. n .. _ ,_ - _ _ ., _ , _ . _ . _ -. _ _ _ ._. ~ . ; _ _ _ _ . _ _ ~.. _ _.~. - __ _. _ _ . _ _
R -----------Pipe reactions under thermal conditions generated by the a postulated accident and including Rg , Y J
-----------Jet Impingement equivalent static load on a structure generated by the postulated accident and including er appropriate dynamic load factor to account for the dynamic nature of the load.
Yr -----------Equivalent static load on the structure generated by the reaction on the broken high-energy pipe during the postulated accident, and including an appropriate dynamic load f actor to account for the dynamic nature of the load. Ym -----------Missile impact equivalent static load on a structure generated by or during the postulated break, such as pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load. In determining an appropriate equivalent static load for Y and Ym e asto plasticbehaviormaybeassumedwithappropriateductilityfalosand5slong- r as excessive deflections will nct result in loss of function of any safety-related system. S------ -For structural steel, S is the require <l section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12,1%9. The 33% increase in allowable stresses for steel due to seismic loadings will not be used. U-----------For concrete structures, U is the section =trength required to 61l resist design loads and based on methods described in ACI 318-77. 7.1.2 Structures D-----------Dead l oads or the* r rel ated interna l momeni s and f orces, including any permanent equipment / system loads and hydrostatic loads due to normal groundwater. L------- Live loads or their related internal moments and forces, including any movable equipment loads and other loads which vary with 46 Intensity and occurrence, such as soll pressure. O Amend. 61 3.7-A.11b Sept. 1981
i I 8.1.1.3 MC Class Components and Steel Containment Vessel
' DESIGN CONDITIONS Dead + Live + T' + Pi + OBE Dead + Live + T' + Pj + SSE Dead + Live + T' + Pe + OBE
[N d.+ Live + T' + Pe + SSE Jeaa Live + T' + R + OBE 46 1 Dead + Live + T' + R + SSE O 1 l *For retive pumps and valves, the SSE and'DSL loadings shall t also be included in Design Me.hanical Loads (as for the OBE) for low temperature Section III design. However, the Design Condition load combinations shall consider the SSE and DSL independent of the OBE. Amend. 47 3.7-A.16b
OPERATING CONDITIONS Normal Operating: Dead & Live + To + OBE 8.1.2 Seismic Categorv i Systems and Comoonents Not Under Jurls, ction of LSME Code The load combinations for Selule Category I systems and components which are not under the jurisdiction of the ASE Code are as given in 8.1.1.1. The allowable stress limits should be consistent with the particular design-Intended function of the component. Special consideration in allowable stress limits should be given to those components which are required to perform a mechanical motion during the course of accomptishing a system safety function. For these com pnents, functional adequacy should be assured by using more restrictive allowable limits such that the resulting deformations do not preclude operability during or af ter the seismic event, or by a combined testing and analysis program which will demonstrate component structural Integrity and capability to perform its safety functions. For those components sach as electrical and instrumentation which cannot be adequately analyzed, v!bration tests shall be used to demonstrate their Integrity under simulated seismic excitations at their supports. Analysis without testing may be acceptable only if structural Integrity alone can assure the design-Intended function. 8.1.3 Selsmic Category Ii Systems and Comoonents Sane load combinations as for the applicable Seismic Category I systems and components except the load combinations involving the SSE are inapplicable, in addition, the " Transients" loads are also inapplicable. 8.2 Load combinations for Structures The general, basic load combinations for Seismic Category I and 11 structures which involve seismic loadings are given below. The detailed load combinations for all types of loads (In addition to seismic) are given in the appropriate structural criteria documents. Design requirements for Seismic Category I and ll steel and concrete structures shall satisfy the requirements of the specifications for the Design, Fabrication and Erection of Structural Steel for Buildings by AISC; the 61l Building Code Requirements for Reinforced Concrete (ACI 318-77), and ASE B & 46 PV Code, Section ill, Division 2, as appropriate. O 3.7-A.16c Amend. 61 Sept. 1981
y Category lli structures shall satisfy the loeding combinations and design d criteria of the Standard Building Code for Zone 2. U No component of an Individual loading condition shall be included which would , render the combination nonconservative. When a particular loading condition ! does not apply, that loading condition shall be deleted from the load i combination. ! l 8.2.1 Seismic cateoorv I Structures ' 8 .2.1.1 Structures Under ASE Code, Section Ill, Division 2 (Foundation Mat Under Containment) The load components and load combinations for these structures are as given in Section CC-3130 and CC-3230 of this Code, as applicable. 8.2.1.2 Concrete Structures (Other Than Those Under ASE Section III Division 2) The Strength Design method in ACI 318-77 shall be used for design of all 61 Category I concrete structures. SERVICE LOAD ComITIONS U = 1.4 D + 1.7 L + 1.9 E If thermal stresses due to To and Ro are present, the following shall also be satisfled: U = 0.75 (1.4 D + 1.7 L + 1.9 E + 1.7 To + 1.7 Ro) Both cases of L having its full value or being completely absent shall be checked. In addition, the following shall be considered: U = 1.2 D + 1.9 E Where soll and/or hydrostatic pressures are present, in addition to all the above combinations where they have been included in L and D respectively, the 61 46 requirements of Sections 9.2.4 and 9.2.5 of ACI 318-77 shall also be satisfied. l O 3.7-A.16d Amend. 61 Sept. 1981
O FACT 0 RED LOAD CONDITIONS U = D + L + To + Ro+E' U'D+L+Ta+R + 1.25 a Pa+Yr + Yj + Ym + 1.25 E U=D+L+Ta + Ra + Pa + Yr + Yj + ym +E' The maximum values of Ta, Ra , P Yr , Yj, and Ym, including an , appropriate dynamic load factor, shaii be used unless a time history analysis is performed to justify otherwise. The above load combinations involving Yr , Yj, and Ym shall be satisfied first without considering the Yr, Yj, and Ym. When considering these loads, however, local section strength capacities may be exceeded under the effects of these concentrated loads, provided there will be no loss of function of any safety-related system. Both cases of L having its full value or being completely absent shall be checked. 8.2.1.3 Steel Structures Elastic working stress design inethods, as specified in Part I of AISC specification for the Design, Fabrication and Erection of Structural Steel for Buildings, shall be used for design of all ' Category I steel structures. SERVICE LOAD CONDITIONS O S=D+L+E If thermal stresses due to To and Ro are present, the following shall also be satisfied: 1.5 S = D + L + To + Ro + E Both cases of L having its full value or being completely absent shall be checked. FACTORED LOAD CONDITIONS 1.6 5 = D + L + To + Ro + E' l.6 S* = D + L + Ta + Ra+Pa+Yr + Yj + Ym + E 1.7 S* = D + L + Ta + Ra + Pa + Yr + Yj + Ym + E' In the above combinations, themal loads may be neglected when it can be shown that they are secondary and self-limiting in nature and where the material is ductile. The maximum values of Pa, T R 46 Yg and Ym, including an appropriate dynamic load factor, sh!ilused be, Yr. Amend. 46
- 3. 7- A.17 Aug.1978
unless a time history analysis is performed to justify doing otherwise. The O~ Ioad combinations involving Y r , Y I and Ym shalI be satisfled fIrst wIthout the
! Y,Y r and Y When considerin Yhese loads, however, maybdexcee"de.d under ^he ef fech of these concentrated loads, provided therelocal section will be no loss of fu"; tion of any safety *related system. Both cases of L having its full value or being completely absent shall be checked.
*For these two combinations, in c.wputing the required secticn strength, S, the plastic section modulus of steel shapes may be used.
8.2.2 Seismic Categorv il Structures ! There are no Category iI structures. 61 36
- 9. Testing criteria Applicable components, equipment, and assembiles of the CRBRP may be seismically qualIfled by testing in Ileu of analysis. Testing shalI be employed for complex equipment that cannot be adequately modeled for a dynamic analysis to correctly predict its response. Also, testing shall be performed for that O
O 3.7-A.18 Amend. 61 Sept. 1981
O equipment whose analysis for structural integrity alone would not necessarily assure its design-intended function. Generally, testing is perfomed for electrical and instrumentation equipment and assemblies. Several test methods may be employed such as sine beat, continuous sine, multiple frequency, etc. Theie are described in IEEE std 344-1975 "IEEE -Recommended Practices for Seismic Qualifications of Class lE Equipment for Nuclear Power Generating Stations." IEEE STD 344 shall constitute the test requirements when not specifically 46 stipulated herein. 1 9.1 Single Frequency Tests 46 l Single frequency testing, when applicable to satisfy IEEE STD 344-1975 provides conservative test motion with definite repetitive 1 effects and a real ._cic time duration to produce a desired equipment response. The equipment can be successively vibrated at a number of resonant frequencies determined from a resonance search. This constitutes severe testing under the most unfavorable conditions where a measured equipment natural frequency has been conservatively assumed to coincide with a building or supporting system natural frequency. Thus, uncertainties in the analytical determination of building natural 46 frequencies will be conservatively accounted, and the maximum peak response acceleration on the response spectrum is conservatively assumed to occur at the equipment's natural frequency. The test procedures s 46lusing single frequency, sine beat tests are given in Section 9.4.1.1 9.2 Multiple Frequency Tests When the seismic ground motion has not been strongly filtered, the floor motion retains the broadband characteristics. Specific 46l input excitation to the shake table includes time history motion, random and complex wave shapes. Multiple frequency testing provides a broadband test motion which is particularly apt for producing simultaneous response from all modes of multidegree of freedom systems. Multiple frequency testing for ground or near-ground supported systems and components provides a closer simulation to a typical seismic ground motion without introducing a higher degree of conservatism. The test 46 procedures using multiple frequency tests are given in Section 9.4.1.2 9.2.1 Multiple Frequency Input Motion For any multi-frequency waveform employed, the shake table motion must be adjusted so that the Test Response Spectrum envelops the Required Response Spectrum over the frequency r6nge for which the particular test is designed. The application of multi-frequency motion shall be simultaneous and it is not sufficient to envelop peak responses 46 independent of time.
- 3. 7- A. 19 Amend. 46 Aug. 1978 h
ATTACHENT C S0ll - STRUCTURE INTERACTION C.1 Structures founded on rock The major Category i structures in CRBRP (Nuclear Island) are founded on rock with an average shear wave velocity of 4000 ft/sec. For these structures the rock structure Interaction shalI be represented in the seismic analysis by equivalent massless foundation springs and dashpots. To take into consideration the different rock and soil materials below and around the Category I structures, the foundation springs for the Nuclear Island shall be calculated by a plane-strain finite element analysis. Analytical models for the North-South and East-West directions shall be constructed. The dynamic elastic properties of the different types of foundation materials shall be represented. The spring stif fnesses obtained from the plane-strain analyses shall be corrected to account for three-dimensional effects. The damping coefficients for the foundation dampers shall be calculated based on the equations for geometrical damping in an elastic half-space using equivalent half-space dynamic properties derived from the spring stif hesses. For this purpose the equations of C.I .1 shall be used to determine "ae rock properties (shear modulus, G). With the calculated value of G and the equations of Section C.1.2, the geometrical damping shall be calculated. C. I .1 Sorino Constants (l) (Elastic Half-Soace) 46 C.1.1.1 Circular Footinas Motion Sorina Constant Vertical kz " 4Gr o 1v Horizontal Kx = 32(- )Gro 7-8v Rocking k 8Grd y = 3 ( 1 v) Tors!on k = 43 g 3 o (1)Taken from Richart, F.E., Hall, J.R., and Woods, R.D., " Vibrations of , Solls and Foundations", Prentice-Hall, Inc., 1970. l l 3.7-A-01 Amend. 61 ' Sept. 1981 l
where: G = Shear modulus rg = Radius of footing v = Poisson's ratio Amend. 46 l AugL;t 1978 3.7-A-Cl-a
l For translation: r /4cd 0 =V , 3 ' For rocking: r o= 16cd 3n For torsion: r g= 16cd(cz + dz) 6n where: 2c = Width of the foundation (along axis of rotation for the case of rocking). 2d = Length of the foundation (in the plane of rotation for rocking). After obtaining the equivalent radius, r ,othe formulas of C.1.2.1 may be used 40 to calculate the damping values for different uotions. C.2 Structures Founded on Soll The seismic analysis of Seismic Category I structures founded on soll shall be conducted using finite elanent techniques. The analysis should account for the strain dependent properties of the soll. The damping ratios given in Tables 4f 3.7A-4 and 3.7A-C-2 for the structures and foundation materials shall be used. (~'N The mathematical model wilI represent the structures and the supporting \j foundation materials, soil and rock, down to the elevation of the foundation of the major Seismic Category I structures (Nuclear Island). The input motions shall be applied at the surface level (finished grade) on an assumed rock outcrop and shall consist of the rock motions used in the analysis of the Nuclear Island. No credit shall be given for the soll cover or overburden in the deconvolation. No point in the final response spectra at the free-field foundation level snali fall below 60% of the design response spectra. The same Iimitation applles to the response spectra calculated at the elevation of the foundation in the soil-structure interaction system. The vibrating motion obtained at the finished grade level should give response spectra that envelop 61 the design response spectra. C.3 Buried Ploes and Conduits Two ef fects are to be considered in the seismic analysis of Category I buried pipes and conduits: " Free-field" behavior, and relative displacement of pipe ends due to building motions. The " Free-field" stresses are applicable to long straight portions of buried pipes. The effects due to relative displacement of pipe ends due to building motions are critical at the ends and at bends of the Iine. C.3.1 " Free-fleid" Stresses 46 Two types of " Free-field" stresses are to be considered: Axlal and bending O 3.7-A-C5 Amend. 61 Sept. 1981
C . 3.1.1 Axial Stresses The *1aximum axial stresses shall be assumed to ie due to a wave traveling in the ground along the longitudinal 2xis of the pipe and producing a ground motion i i the same direction. The maximum axial strain, cmax , shall be calculated as:
- max
- V max /V Therefore, o max = E c max =Ev Where: vmax =MaximumgroundvelocT[y/V in the longitudinal direction.
It may be calculated by integrating the ac:eleration-time history of the ground motion at the elevation of the pipe. V = Wave propagation velocity of the foundation material in the longi-tudinal direction.
= Axial Str ess o*Noung E Modulus of pipe material C . 3.1. 2 Bending Stresses The maximum bending st ess for the " Free-field" shall be assumed to be due to a wave traveling in the ground along the longitudinal axis of the pipe and producing a motion transverse to that direction. The curvature of the pipe + max shall be deterained as:
= a h.ax max /
Where: a = The maximum ground acceleration man = The wave velocity of the foundation material. M = EI 4 max , where EI is the flexural rigidity of the pipe and M is the bending moment acting on the pipe. o = Mrg = EI 4max F o
=Er ga max 1 1 y2 where O B = bending stress and g r = the outside radius of the pipe.
The total " Free-field" stress shall be obtained by the square root of the sum of the squares of the axial and bending stresses. 46 Amend. 46
- 3. 7-A-C-E a August 1978
3.8 DESICN OF CATEGORY l STRUCTURES
, 3.8.1 Concrete containment (Not Aoolicable) 3.8.2 Steel Centainment System 3.8.2.1 Descriotion of the Containment The Containment Vessel is a low leakage, free-standing, all welded steel vessel anchored to the base mat with a steel lined concrete bottom in the form of a vertical right cylinder having an Inside diameter of 186 feet and with side walls extending approximately 169 feet from the flat bottom liner at the 4f 45 base to the spring line of the ellipsoidal-spherical dome. The cylindrical shell is embedded in concrete ep to the elevation of the operating floor. On the Inside of the Containment Vessel, there is the continuous reinforced concrete wall comprising the peripheral boundary of the Internal concrete structure. Butting against the outside face of the steel shell from elevation 45l 733 feet up to the elevation of the underside of the operating floor, there is anothe reinforced concrete wall of sufficient thickness designed to prevent y buckling of the steel shell. Neither of the two concrete walls are considered part of the containment vessel. Alumina-sIIIca insulation is attached to the 33 Inside surface of the Containment Vessel from elevation 816 feet to elevation 823 feet. For the Design Basis Accident, a minimum of 3 inches of insulation, having a value of 0.0267 Btu /hr-ft-oF, is required to limit the shell 48 61 temperature, at elevation 816 , to 1300F.
Its shell, a 1/4" bottom liner plate, one access
] 45lairlock, The vessel one emergency includes: egress airlock, vacuum relief system, one equipment v hatch, penetrations, inspection ladders, miscellaneous appurtenances and attachmeurs. The configuration of The Containment Building is shown in figures in Section 1.2. The design lifetime of the containment vessel shall 39 be 30 years.
3.8.2.2 Acolicable Codes. Standards and Soscifications 3.8.2.2.1 Codes The Containment Vessel wilI be designed, material procured, fabricated, i installed and tested in accordance with the requirements of the ASE B&PV ! 43 Code, Section ill, Division 1,1974 Edition with Addenda through Winter 1974 SC and Code cases 1713,1714,1809,1682 and 1785 and ASE-lil, Division 2,1975 Edition, Subsection CC, for the steel lIned concrete containment bottom. The design shall also meet the requirements of the Class MC Section of RDT Standard E15-2T, " Requirements for Nuclear Components". l l l O 3.8-1 Amend. 61 Sept. 1981
O The quality assurance procedures will be in accordance with RDT Standard F2-2 as well as meeting the requirements of the ASME Code,1 45 i Section III, Divisions 1 and 2. All structural steel non-pressure parts such as ladders, walkways, handrail, etc. will be designed in accordance with the American Institute of Steel Construction (AISC), " Specification for the Design, FOrication and Erection of Structural Steel Buildings (AISC, February 12,1969). 3.8.2.2.2 Design Specification Summary and Design Criteria The Containment Vessel, including all access openings and penetrations will be designed such that the leakage of radioactive materials from the Containment under conditions of temperature and pressure resulting from the extremely unlikely faults could not cause undue risk to the health and safety of the public and will not result in potential offsite exposures in excess of guideline values of 10CFR100. O 3.8-la Amend. 48 Feb. 1979
Corrosion Protection Potential corrosion of the steel containment has been considered at the portion embedded in the concrete as welI as the exposed portion above 45 operating floor elevation. The conditions which determine corroslo; are basically the electro-potential of the materials involved, the presence of oxygen and an electrolyte, temperature and any induced electro-potential from extraneous sources. These have been evaluated in the deter.nination of corrosion. The corrosion of the steel containment face in contact with the containment concrete is not a design consideration since portland cement concrete provides good protection to embedded steel. The protective value of the concrete is ascribed to its alkalinity and relatively high electrical resistivity in atmospheric exposure. ACI Committee 201 Report " Durability of Concrete in Service" identifies three basic conditions as being conducive to the corrosion of steel in concrete (Reference 2).
- 1. The presence of cracks extending from the exposed surface of the concrete to the steel.
- 2. Corrosion cells arising from electro-potential differences in the concrete itself.
- 3. Electrolysis by Induced currents in the concrete or steel.
With respect to condition (1), a minimum of 22" inches of concrete embedment 45 from elevation 816 feet down surrounds all of the steel containment. The cracking under the worst of cases is considered minimal. This quantity far surpasses minimum cover recommended by ACI 201-1 In the most corrosive marine environment. With respect to Condition 2, the potential for developing corrosion cells will be kept to a minimum by limiting the soluble salts and chlorides in the concrete. Further, the continuing corrosion of iron under these conditions requires that the hydrogen deposited at the cathode is freed or combined with 61Ioxyger. Since both these mechanisms are inhibited by the cencrete, the corrosion cells are polarized, and the reaction is brought to a standstill. With respect to Condition 3, to preclude the development of Induced electric currents and in keeping with good construction practice, all electrical equipment and structures will be grounded as determined by the resistivity of the foundation materials for the site. Foundation material resistivity surveys will be made and the result considered in the design and determination of the extent of the grounding mat. 16 l ( l O 3.8-2a Amend. 61 Sept. 1981
Protective _ Coatings Protective coatings shall be applied to all exposed surf aces of the Containment Vessel. The surf aces to be coated shall be sand blasted and prime 61l coated with an inorganic zinc silicate such as Amercoat Dimecote 6. The inside surface shall be topcoated with a fully compatible epoxy or phenolic coati ng. The outside surface shall be insulated, if required by design, or topcoated with a compat!ble opoxy or phenolic coating. O O 3.8-2b Amend. 61 Sept. 1981
Tolerances V The Containment Vessel as constructed shall not exceed the tolerance requirements of NE-4000 of ASE-lil for f abrication or erection. The dimensional control procedures shall meet the requirements of RDT STD F3-15T. The out-of-plumb tolerances shalI 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 Acolicable NRC Regulations and Regulatorv Guides 15 NRC Regulatorv Guides The applicable regulatory guides are listed below. 1.10: Mechanical (Caldwell) Splices in Reinforcing Bars of Category i Concrete Structures (Revision 1, January 2, 1973).
.11: Instrument Lines Penetrating Primary Reactor Containment 45 (March 10, 1971) 1.1 h Instrumentation for Earthquakes (Revision 1, April,1974) 61l 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 II, 1972) 1.29: Seismic Design Classification (Revision 2, August 1976) 50 1.55: Concrete Placement in Category 1, Structures (June 1973) i 45 1.57: Design Limits and Loading Combinations for Metal Primary Reactor i Containment System Components (June,1973) i 1.60: Design Respon::e Spectra for Seismic Design of Nuclear Power Plants (Revision 1, December, 1973) 45 1.61: Damping Values for Seismic Design of Nuclear Power Plants (Oct. ! 1973) 1.63: Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants (Oct. 1973) 1.69: Concrete and Radiation Shields for Nuclear Power Plants 61l (December, 1973) 1.73: Physical Independence of Electrical Systems Division 2 61 (September, 1978) 3.8-3 Amend. 61 Sept. 1981
1.85: Materiais Code Case Acceptability - ASE Section ill, Division 1, 50 1976 1.92: Combining Modal Responses and Spatial Components In Seismic 45 Response Analysis (Revision 1, Feb.,1976) 1.102: Flood Protection for Nuclear Power Plants Rev.1 (September 1976) 1.117: Tornado Design Classification (September 1976) 1.222: Development of Floor Design response Spectra for Seismic Design of Floor-Supported Equipment or Components (September 1976) 1.124: Design Limits and Loading Combinations for Class I Linear-Type 61 Component Supports Of the above, Regulatory Guide 1.63 is applicable after the foflowing changes:
- 1. Deleting " 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 50".
4 Replacing " loss of coolant accident" with " containment design basis accident".
- 5. Substituting "(Summer 1972 Addenda)" following"..... ASK Boller and Pressure Vessel Code" with ",1974 Edition".
Construction No special construction techniques are anticipated for this containment vessel. O 3.8-3a Amend. 61 Sept. 1981
3.8.2.3 Loads and Loading Combinations 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, includlag the weight of the steel containment vessel, penetration sleeves, equipment and personnel access hatches, and 45 ther attachments supported by the vessel, plus loads due to concrete shrinkage. 30 L - Live Loads, as applicable, including:
- 1. Penetration Loads (including seismic), as applicable
- 2. Floor Loads - 100 PSF
- 3. Walkways -200 lbs per Iinear foot
- 4. Equipment and Personnel Airlock Floor Load - 300 PSF or 100,000 lbs moving concentrated load
- 5. Emergency Airlock Floor Load -200 PSF or 10,000 lbs.
- 6. Polar Crane Loads (Ref. 1) 45 7. Construction Loads *
- 8. Meznnine -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 18 Support Clips - combined with a Dead Load on all clips of 200 lbs.
each. Pg - Internal Design Pressure (or Transient Pressure Loads) P - External Design Pressure e P - Testing Pressure t T - Thermal loads due to temperature gradient through walis under o
, normal operating conditions.
T' - Thermal loads due to temperature gradient through walls from accidents, such as major sodium ffres. Tt - Thermal load under testing temperature conaltions. E - Loads resulting from an Operating Basis Earthquake (OBE) E' - Loads resulting from a Safe Shutdown Earthquake (SSE) I
- A concrete placement load, resulting from using the vessel shell below operating floor elevation as the formwork for placing the reinforced concrete walls, and loads that are imposed by concrete 61 f rms when constructing the confinement shell. A snow load will be considered also during the construction period.
O , 3.8-5 Amend. 61 Sept. 1981
18 45 R - Accident loads due to Extremely Unlikely Faults (such as interf acing loads from Inner cell concrete structures). Note (1) The crane live load shall include, as appropriate, the vertical Impact load and the lateral thrust load as determined in accordance with Reference 3. Design Pressures and Temoeratures The design pressures and the associated design temperatures shall be as specified 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 and attsndant temperature loading due to extremely unlikely faults. Details of this information will be provided in the FSAR. The operating condition containment atmosphere temperature and pressure are as follows: Operating Conditicn Temperature = 700F Operating Condition Pressure = 0.0 psig Lowest Service Metal Temperature = 150F 3.8.2.3.2 Loading Combinations The loading combinations for which the vessel and its appurtenances are to be designed shall be, but not limited to, those specified in Table 3.8-1. The containment design requirements and limits siiall be in accordance with ASME-1Il, ArtIcie NE-3000. For conditions where seismic loads are involved, the desip analysis requirements and criteria as contained in Section 3.7 shall also be met. 61 30 0 Amend. 61 3.8-6 Sept. 1981
l O For conditions where compressive stresses occur, the critical buckling stress shall be taken into account. For verification of design 45l adequacy, the buckling stress criteria as given in Appendix 3.8A 4 shall be met. 3.8.2.4 Design and Analysis Procedure i The design conditions for this vessel are as follows: 1 i Internal Design Pressure 10 PSIG External Design Pressure .5 PSIG
- Design Temperature 2500F Operating Condition Temperature 700F
- ' Operating Condition Pressure 0.0 PSIG Lowest Service Metal Temperature 150F.
, Test Pressure 11.5 PSIG l
4 I 18 l The containment vessel will be designed for all of the loads specified in Table 3.8-1. A complete set of detailed fabrication and erection drawings and a stress report in accordance with the 30
- Westinghouse specification and ASME, Section III, Division 1, Sub-l section NE for Class MC Components with Winter of 1974 Addenda and Section III, Division 2 of the ASME Code dated July 1975 will 45 be provided. An "N" stamp will be applied to the vessel.
Design and Analysis t The containment vessel is a free standing, vertical cylindrical 45 [ steel pressure vessel with a spherical and ellipsoidal top head and a flat bottom steel liner plate. The vessel will be~ anchored into i the foundation with a skirt type anchorage system with concrete 45 walls on both sides of the cylindrical portion extending upward from the base a distance of 86 feet. t l- The shell will be designed using the basic membrane equations i for thin shells with the stress allowables as defined in Article
- NE-3000 of Section III, Division 1, except as noted below.
The shell will be analyzed for the external design pressure . using -the procedures outlined -in Paragraph NE-3133 of Section III of the ASME Code. The compressive stresses in the shell due to loads described in NE 3112.4 (b) will be lim 1^.ed'to the allowables of Paragraph NE-3112.4 of the ASME Code. The buckling allowables for Construction and Accident.and Environmental (including E') loads ' must also satisfy those'specified in Appendix 3.8A. 30 l, 1 le lO ! Amend. 45' July 1978 1
, 3.8-7 1
1 4
,.g..mm-,,~.-,m..-~.-
em--p----,,, , ~ - . , , , , ,,,n,- .w,.,wn,,,u.-, ,~m.,-,,,_,m,,,,-,-,,ng.r,.,,,,,,,,.mm.,,,-.wv.,_w-we.,,,- , mm, n n
In the regions of the vessel where there are substantial thermal and mechanical loads other than pressure, (eg. lHTS piping penetrations) the 30 containment vessel wilI be designed and analyzed in accordance with NE-3131 (b) of Section lil of the AS E Code. The basic stress intensity limits of j 61f NE-3221.1, NE-3221.2 and NE-3222.2 wII I be satisfled for pressure Ioads in combination with all mechanical and thermal loads, using S allowablestress,S,tabulatedinAppendixiofSection117.equaltothe The polar crane support will be designed in accordance with Subsection NE of Section lli of the ASE Code f or the portion of the shell included in the crane girder analysis. The structural parts of the crane girder which Ile within the code boundaries shall be designed in accordance with Paragraph NE-3131 (e) of Section til of the AS E Code. The design of the structural parts of the crane girder beyond the code bounuaries will be designed in accordance with the AISC Specifications. The transition region at the point of embedment (elevation 816'-0") wilI be analyzed using a Shells of Revolution Progran based on the paper, " Analysis of Shells of Revolution Subjected to Symmetrical and Non-symmetrical Loads" by A. Kalnins, which appeared in the September 1964 Issue of Journal of Aoolled Mechanics. This region will be analyzed for loads due to Internal pressure, a j thermal gradient, the containment dead load, and earthquake loads. l30 The containment wilI not be subjecied to non-axisymmetric pressure and temperature distributions above the operating fIoor ievel. They wIII occur only below the operating floor where sodium spills are postulated for specific cells. The accident conditions in each of the celis adjacent to the containment vessel will be considered in the analysis since the deformations of the reinforced concrete walls of a cell caused by the accident temperatures and pressures will Induce stresses and strains in the embedded steel shell. Since no direct connection (anchors or embedments) is provided between the steel shell and the concrete walls, the stresses in the steel shell will be calculated assuming compatible deformations at the interf ace with the concrete walls on either side and that no tensile and shear forces can be developed at the interface. The stresses cf the Individual concrete and steel components will be checked to be within the appropriate ACI and ASE Section 111, 33 Division I code lImits. 6j The confinement structure will be constructed of reinforced concrete approximately four feet thick with a 196' approximate inside diameter centered on the containment vessel. The Confinement structure will be a Seismic 18 Category 1, tornado hardened structure. O 3.8-7a Amend. 61 Sept. 1981
- 3.8.2.5.4 Leak Testing Airlocks The airlocks will be pressurized with air to 11.5 psig. All welds and seatr.
i 61l will be observed for visual signs of distress or noticeable leakage. The airlock pressure wilI then be reduced to 10 psig and a thick soap solution will be applied to all welds and seals and observed for bubbles or dry flaking as indications of leaks. All leaks and questionable areas will be repaired. During the overpressure testing the nuter door will be locked with hold-down de" Ices if required to prevent upsetting of the seals. 2 a O i 1 1 i Amend. 61 3.8-7d Sept. 1981
.,~a-- .,,,,,.,.,,,,-.-,.,-n.,..,-,,,.,.,,,,,-,-..-,-,,,,.,--,,--,,--.-.mn-.,,,-a..~nn-,+,,,,n,,, ___, _ , - -,-..-,,,-m_,,,,,,--,-,,,,,-,..v..,
The internal pressure of the airlock will be reduced to atmospheric pressure and all leaks repaired af ter which the airlock will again be pressurized to 10 psig with air and all areas suspected or known to have leaked during the previous test be retested by above soap bubble technique. This procedure wilI be repeated until no leaks are discernible by this means of testing. 3.8.2.5.5 Bottom Liner Plate Test Before placing concrete over the bottom liner plate, the leak tightness of the Iiner shalI be verIfled. AlI IIner plates shalI be vacuum box tested for ieak tightness. Upon completion of a successful leak test, the welds shall be covered with channels, and the channels leak tested by pressurization. The testing procedures for liner seem wolds shall be in conformance to the 45 requirements of Regulatory Guide 1.19 (Refer to Section 3.8.5.2). 3.8.2.6 Deslan Loadino Combination Stress Limits Details of loading conditions and the design stress limits associated with allowable stress criteria for the containment vessel will be provided as the design is developed and will demonstrate that the requirements of Section 3.8.2.5 are fully complied with. Loading combinations are provided in Table 3.8-1. The allowable stress limits shall be as defined in subsection NE-3000 of the ASE-Ill Code and shown in Table 3.8-3. 30 For buckling stress criteria, the same criteria as given in Appendix 3.8A. 130 3.8.3 Concrete and Structural Steel Internal structures of Steel Containment 3.8.3.1 Descriotion of Internal SYructures The Internal structures within the containment principally consist of the cells and other areas as listed in Table 3.8-2 and as shown on the General 61 Arrangement figures of Section 1.2. The Internal structures are enclosed by two continuous circular walls located on each f ace of the steel containment vessel between the foundation mat and operating floor levels. The circular 61l walls act as a radiation shield and pressure boundary in local cell areas, as ' 45 a support for vertical loads and carry horizontal shears to the foundation mat. The entire steel containment vessel wilI be designed for the 10 psig Internal pressure. The detailed physical description provided herein is limited to those cells which significantly contribute to the ' structural system. These celis are reinf<rced concrete structures designed to the requirements as noted in Table 3.8-2. O 3.8-8 Amend. 61 Sept. 1981
61l Cells in which there exists a potential for spillage of thermally hot sodium w11l be provided wIth steeI ce1I IIners wIth Insulation and ceiI venting O- features to protect the structural concrete. The liners will be designed to contain a spilI of high temperature sodlur:: once in its lIfetime. For Iiner 61 strain criteria under sodium spilI conditions see Appeudix 3.8-B. 24 The reactor cavity housing the Reactor Vessel and guard vessel is located near the center of the containment system. Three primary heat transport system celis form an annulus around the reactor cavity on three sides. The reactor overflow vessel and primary sodium storage vessel cell surrounds the fourth side. The main work area inside the containment vessel is the operating floor above the reactor. Since the main work area is designed for continuous occupancy, the concrete siab for the operating fIcor w11I have suf fIclent thickness to meet the structural and shielding requirements. Each cell is designed for the accident conditions and will also be designed to Ilmit accident ef fects. When interior structures interact with the containment vessel, appropriate Interactive effects will be included in the containment vessel analysis. 6_1 Thermal growth of a celi may be inhibited by neighboring elements. in such cases appropriate allowance will be made in the design to provide for the restraining of thermal loads. Under seismic loads, the structure will act as an assemblage of shear walls. The structure will be checked to insure that the shear walls are adequate to sustain the ir.teral forces from seismic and other loads. Removable slabs and plugs will be provided in areas where access is required s for operation and/or maintenance purposes. Sufficient thickness will be provided for these slabs or plugs in order to meet the structural as well as radiation shielding requirements. Ali Interior structures wIthin containment such as walIs, siabs, steei framing 61l and the El&C cubicles above the operating floor will be designed as seismic Category I structures. Therefore, no f ailure of structures within the containment wilI result from a SSE. 25 3.8.3.1.1 .[Leactor Cavity The reactor cavity is a hollow concrete cylinder closed at the bottom with a l concrete slab. At the top of the cavity, t steel ledge, partially embedded in l the cavity wall, is provided to support the reactor vessel. The support ledge I will be designed such that it will withstand loads based upon Structural l Margins Beyond the Design Base (See Reference 10a, PSAR Section 1.6). The l reactor cavity wall has a 40 foot internal diameter with a thickness of 7 feet. The Interior of the cavity is lined with carbon steel plates. See RC8 34 61 GA's in Section 1.2 for arrangement details of the Reactor Cavity. j 24 l 3.8-9 Amend. 61 Sept. 1981 l --. .- , . - . - - . . - - - - -- - . _
The Reactor Vessel Support Ledge is comprised of steel brackets with a steel ring plate at the top of the brackets and another ring plate at the middle of the brackets. The function of the upper plate is to support the reactor vessel support systun. The lower plate reacts against the bearing washers of the tie-down bolts when the reactor vessel head exerts uplift forces on the ledge. There are 69 sleeves, one for each bolt, spaced along a bolt circle of 13 '-1" radi us. The typical section of the ledge is shown in Figure 3.8-9. In the areas of the cut-outs where Primary Heat Tnnsport System pipes penetrate the cavity, the cross section of the ledge is shown in Figure 3.6-8. There are three such cut-out areas, one at each penetration. The top of the reactor ca/Ity is also the floor of the Head Access Area, which is enclosed by 6'-6" thick walls on three sides and a 4'-0" wall on the south side with its inside dimension being a 44'-0" square. 34 The temperature of the cavity structural concrete will be limited to those required by the AShE Code Section ill, Division ll under paragraph CC 3440. The ledge design will also be evaluated against a Structural Margin Beyond Design Base (Sh0DB) load of 50 million pounds upward or downward. This load is not combined with loads other than dead loads with a load f actor of one. 34 in order to mitigate the consequences of very low probability accidents beyond the design basis, such as SkEDB, the additional design features described 61 below are provided. A detailed discussion of Thermal Margins Beyed the Design Base (TMBDB) in the CRBRP is provided in Reference 10b of Section 1.6. Post accider.t intercommunication between the Reactor Cavity (RC) and the Reactor Conu !nment Buildi 3 above the operating floor is provided by two separate end Isolable ve . paths. The vents are designed to relieve at a differential pressure of 16 psid between the reactor cavity and the RC8 atmosphere. A normally open motor operated valve is provided to isolate the reactor cavity in the unlikely event of failure of the r'yture disc during normal plant operation or a minor accident. A reactor cavity liner venting system is provided for relieving pressure from behind the Iiner caused by steam and carbon dioxide (CO releas d from he concreteofthereactorcavityandPHTSpipewaycells,2khereaciorcavly (RC) liner venting system is olvided into three physically separated parts: the floor, the wall, and the ceiling. The RC cell floor is vented to non-inerted spaces above the RW operating floor (Elevation 816'). The wall subsystem is vented to the non-Inerted area in the RCB below the operating s5 floor. The ceiling liner vents and PHTS pipeway cell liner vents are vented 32 to non-critical areas below the RG operating floor. , 9 3.8-9a Amend. 61 Sept. 1981
3.8.3.1.2 Head Access Area (HAA) ! The head access area is located below the opera'%g floor level 25landabovethereactorcavity. feet long on each side and 14 The access feet area is high above theofreactor a svare shape head. The44 43 l head access area *is a reinforced concrete structure. Steel framing will i be provided in this area to support the EVTM operations. 3.8.3.1.3 Primary Heat Transport System (PHTS) Cell Each PHTS cell is a step type rectangular reinforced concrete structure. At its widest section, the cell is approximately 48 feet wide by 72 feet long. The cell is approximately 58 feet deep in the
~
6 area housing the primary sodium pump and intermediate heat exchanger. The interior surfaces of the cells are lined with carbon steel plate with the lower portion of the plate designed to contain hot sodium spills. The cells are' designed to withstand accident pressures as noted in Table 3.8-2. 3.8.3.1.4 Reactor Overflow Vessel and Primary Scdium Storage Vessel Cell This cell is a step type rectangular reinforced concrete struc-ture. The cell is approximately 26 feet wide by 69 feet long with 62 feet height at its deepest section. The interior surface of the cell is lined with carbon steel plate similar to the PHTS cells. The cell is-O designed to withstand accident pressure and temperature conditions noted in Table 3.8-2. 3.8.3.1.5 Other Cells These cells are reinforced concrete structures with various sizes. The cells required to maintain.a nitrogen at osphere during 37l Table operations 3.8-2. will beliners Cell ifned are anddescribed designedintoSection the requiredents 3A.8. noted in 3.8.3.1.6 Fill Slab A structure fill slab of suitable thickness will be provided over the bottom containment liner plate. 3.8.3.2 Applicable Codes, Standards and Specifications 3.8.3.2.1 Design Codes Applicable provisions both mandatory and recommended of the following codes will be used in the design of the internal structures: 3.8-10 Amend. 45 July 1978
- a. Building Code Requirements for Reinforced Concrete of the American 61l Concrete institute (ACl-318-77)
- b. Specification for the Design, Fabrication and Erection of Structural Steel for Buildings, American Institute of Steel Construction, February,1%9, including Supplement 1 (11/70), St.ppimient 2 (12/71) 37 and Supplement 3 (10/75).
O O 3.8-10a Amend. 61 Sept. 1981
- c. Code for Concrete Reactor Vessels and Containments of ACI ASE Committee (AS E B&PV Code, Section Ill, Division 2 1975.
45 ApplIcat'le State Codes d.
- e. Specification for the design and construction of Reinforced Concrete 61 chimneys ( ACl-307-69) . (See below)
ACl-318 wilI be extensively used for the design of the Internal structures. ACl-318 is generally based upon ultimate load design. Since loading 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 wilI isot be invoked since this is not relevant to the structures under discussion. Applicable portions of Appendix A of ACl-318 wilI be applled to the design. Since ACI-318 dcas not fully cover design requirements for thermal stresses due to temperature gradient, the recommendations cf ACI-307 61 will be used for guidance in addition to ASE Section Ill, Divls!cn 2 for specific areas. AISC specification will t,e applied to the structural steel merrbers such as steel embedments, beams, equipment and pipe supports end restraint structures. 15 l 3.8.3.2.2 Structural Soecifications See subsection 3 A.4.3 for the app, apriate structural specifications. 3.8.3.2.3 NRC Regylatorv Guides The design will meet requirements or basic intent of flie following NRC 15 Regulatory Guides: 61l a. 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category l Concrete Structurcs (Revision 1, 1/73) 44
- b. 1.15 Testing of Reinforcing Bars for Category i Concrete Structures (Revision 1, 12-28-72)
- 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 Spectru far Seismic Design of Nuclear Power 45 Plants (Revision 1, December,1973)
O 3.8-11 Amend. 61 Sept. 1981
- g. 1.61 Damping Values for Seismic Design of Nuclear Power Plants (Oct.
1973) to 1.69 Concrete Radiation Shield for Nuclear Power Plants (t?/73) I. 1.92 Combine Modal Responses and Spatial Components in Seismic 45 Response Analysis (Revision 1, Feb,1976) 3.8.3.2.4 ASTM Standards All ASTM Standards to the extent they are referenced in the codes and 61l standards noted in Section 3.8.3.2 and further specifically identified in other parts of Section 3.8.8, ?!ll be applied to the design of the facility. O l i 1 l 1 l l l l l 9 3.8-11 a Amend. 61 Sept. 1981 1 l
3.8.3.3 Loads and Loading Combinations 3.8.3.3.1 Loads. Definition of Terms and mvaanciature The following nomenclature and definition of load terms will apply to all internal s+ructures unless otherwise noted: 3.8.3.3.1.1 Normal Loads Normal loads are those ioads to be encountered during normal plant cperation and shutdown. They include the following: D - Dead loads or their related internal moments and forces, including any permanent equipment /.ystem loads and hydrostatic loads
- due to normal groundwater.
L - Live loads or their related l' ternal moments and forces, including any movable equipment loads and other loads which vary sith Intensity and occurrence, such as soll pressure.* To- Thermal effects and loads during normal operating or shutdown conditions, based on the mos+ critical transient or steady state 45 condition. Ro - Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition. 3.8.3.3.1.2 Severe Environmental Loads Severe environmental loads are those loads that could Infrequently be encountered during the plant life. Included in this category are: 45 E - Loads generated by the Operating Basis Earthquake (OBE). W '.oads generated by the design wind and specified fo the plant. (Wind load does not apply to internal structures.) l 3.8.3.3.1.3 Extreme E-vironmental Loads l Extreme environmental loads are these loads which are credible but are highly improbable. They include: E'- Loads generated by the Safe Shutdown Earthquake (SSE). I l *For treatment of hydrostatic loads and soll pressure, Sections 9.2.4 and 61 9.2.5 of the ACl 318-77 Code shall apply. b v Amend. 61 1 3.8-12 Sept. 1981 _ _- ._ ._ _ - - ., _ . ~ . . _ .
WT ---Leads generat;d by tha Design Basis Tcrnado as specificd in Section 3.3. They include loads due to the tornado wind pressure, loads due to the tornado-created differential 6 s pressures, and loads due to the tornado-generated missiles.
) (Tornado loads do not apply to internal structures.)
H ---Hydrostatic loads due to maximum flood (as defined in Section 3.4). 3.8.3.3.1.4 Abnomal Loads Abnomal loads are those loads generated by a postulated accident within a building and/or compartment thereof. Included in this category are the following: Pa ---Pressure equivalent static load within or across a compartment and/or building, generated by the postulated accident, and including an appropriate dynamic load factor to account for the dynamic nature of the load, T3 ---Thermal loads under thermal conditions generated by the postulated accident and including Tg. Ra ---Pipe reactjons under thermal conditions generated by the postulated accident and including Rg. A ---Force or pressure on structure due to third level design margin requirement Y3 ---Jet impingement equivalent static load on a structure generated by the postulated accident, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Yr ---Ec.uivalent static load on the structure generated by the reaction on the broken high-energy pipe during the postulated accident, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Y*---Missile impact equivalent static load on.a structure generated by or during the postulated break, such as pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load. In determining an appropriate egoivalent static load for Yr, Yj and Ym, elasto-plastic behavior may be assumed with appropriate ductility ratios and as long as excessive deflections will not result in loss of function of any safety-related system. 3.8.3.3.1.5 Other Definitions S ---For structural steel, S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC " Specification for the Design, ( Fabrication and Erection of Structural Steel for Buildings," February 12, 1969. Amend.'25 Aug. 1976 3.8-13 l
-. - . .- . -_. - -- - .a
The 33% increase in allowable stresses for steel due to seismic or wind loadings will not be used. U - For concrete structures, U is the section strength required to resist 61 l design loads and based on methods described in ACI 318-77. 3.8.3.3.2 Internal Structure as consvinment No portion of the internal structure provides a direct containment f unction. 45 The embedded part of the steel containment is designed such that it can withstand the design pressure without the assistance of the concrete walls. 3.8.3.3.3 Creeo. Shrinkage and Local Stresses No prostressed concrete design is considered for the design of the facliity. Therefore, creep and shrinkage loads will be only considered to the extent they are provided in the reference concre% codes or as may be warranted by prudent design approach. The loads transrerred from the support structure that generally influence local areas will be checked to insure that the local stresses are within acceptable lImits to preclude impairment of the structural function. 3.8.3.3.4 Loads Due to structural Margin Beyond Design Base (SMBDB) The steel ledge cupporting the reactor vessel wilI have a design capability to 61 sustain the loao (equal to 50,000 kips) due to SM3DB requirement. The reactor cavity wilI be aLI,, to sustain an accident pressure noted in Table 3.8-2. 3.8.3.3.5 Sodlun Fire Load See Table 3.8-2 for the accident pressures and temperature loads. 3.8.3.3.6 Hot Sodlum Solli Effect The portions of the reactor cavity and cells, where exposure to radioactive hot sodium is a design basis accident, are provided with carbon steel liners 37 designed to survive a sodium splii (see Section 3A.8). The liners will not compromise gas tightness of the cell. 3.8.3.3.7 Accident Temoerature Load See Table 3.8-2 for design temper atures. 3.8.3.3.8 Negative Pressure on the Liners Any negative pressure on the liner will be resisted by a grid of structural anchors embedded in the concrete. O 3.8-14 Amend. 61 Sept. 1981 i
3.8.3.3.9 Hot Soots s Hot spots will occur under both normal and accident conditions. The ASE-ACI-Section 111 Olvision 2 Code under 00-3440 gives the following concrete temperature iimitations: o For normal operation or any other long term period, at the f ace of structure! concrete, the maximum allowable temperature is 1500F except for local areas, such as around a penetration, where the maximum allowable temperature is 200oF. o For accident or any other short term period, at the interior face of structural concrete, the maximum allowable temperature is 350oF except local areas where the maximum allowable temperature is 650C . o Higher temperatures than given above may be allowed in the concrete if tests are provided to evaluate the reduction in strength and this reduction is applied to the design allowables. Also, evidence shall be provided which verifles that the increased temperatures do not cause deterioration of the concrete either with or without load. Under normal operating conditions, hot spots may occur at the support of the eqelpment operating at high temperatures, su(*1 as: Reactor Vessel Primary Sodium Pumps Intermediate Heat Exchanger O Reactor OverfIow Vesse! Primary Sodium Storage Vessel Primary Sodium Cold Traps NaK Cold Traps NaK Storage Vessel 46I - Primary Sodium Make-up Pumps NaK Cold Trap Pumps OverfIow Heat Exchanger 46 - Make-up Pump Drain Vessel Primary Plugging Temperature Indicators Also, hot spots may occur at the penetrations of the Primary Heat Transport System main loops, and the Auxillary Liquid Metal System piping. Under accident conditions, the hot spots may occur should sodium spilis take place. Concrete limiting temperature, as allowed by the Code, may be used for the final CRBRP structural design based on the results from proposed elevated concrete temperature tests. - 34 l O 3.8-15 Amend. 61 Sept. 1981
3.8.3.3.10 Loading Combinations The loading combinations for the internal structures will meet the following O requirements: 3.8.3.3.10.1 Loading Combinations for Concrete Structures A. Load Combinations for Service load Conditions Strength Design method will be used for design of all Category I concrete structures. For service loading conditions, the following load combinations wIlI be satisfled:
- 1) U = 1.4 D + 1.7 L
- 2) U = 1.4 D + 1.7 L + 1.9 E
- 3) U = 1.4 D + 1.7 L + 1.7 W If thermal stresses due to To and Ro are present, ti,e f ollowing combinations will also be satisfied:
Ib) U = (0.75) (1.4 D + 1.7 L + 1.7 To + 1.7 Ro) 2b) U = (0.75) (1.4 D + 1.7 L + 1.9 E + 1.7 To + 1,7 pg) 61 3b) U = (0.75) (1.4 0 + 1.7 L + 1.7 W + 1.7 To + 1.7 Ro) 9 Both cases of L having its full value or being completely absent will be 45 checked. in addition, the following combinations should be considered: , 1 34 & 2b') U = 1.2 D + 1.9 E W 3b') U = 1.2 D + 1.7 W Where soll and/or hydrostatic pressures are present, in addition to all the
, above combinations where they have been included in L and D respectively, the 61 l requirements of Sections 9.2.4 and 9.2.5 of ACl 318-77 will also be satisfied.
B. Load Combinations for Factored Load Conditions For these conditions, which represent Extreme Environmental, Abnormal, Abnormal / Severe Environmental and Abnormal / Extreme Environmental conditions, respectively, the Strength Design method will be used and the following load combinations will be satisfied:
- 4) U=D+L+T o + Ro + E'
- 5) U = D + L + To + Ro + WT 61 6) U = D + L + T,+ p,+ 1,5 p, l6 O
Amend. 61 3.8-15a Sept. 1981
Since the walls, ceilings and floors of each cell are considered as two-way
/ slabs, ti a applied loads, used in the analysis, will be proportioned to the one-food wide strips, in orthogonal directions, according to the ratio of their celative stif f nesses.
The cell design will be verified by using a three dimensional finite-element analysis with the computer program NASTRAN. The cell and adjacent structures wIlI be represented in the mathematical model which wIlI include the interaction with the containment shell and the ext <.trior concrete wall. The appropriate loads and load combinations will be used in the analysis. 33 Further detailed analysis will be performed in areas of load concentration and penetrations t.s 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 against acceptable published programs. All vertical loads will be transferred to the foundation mat by three principal structural elements, viz (a) walls of PHTS cells (b) perimeter wall around containment and (c) reactor cavity. 3.8.3.4.2 M lvsis for Selsmic Loads Equivalent static r.eismic loads as developed fron' the dynamic analysis of the structure will be transferred through the horizontal slab diaphragms and vertical sheer walls to the foundation mat. The details of seisanic analysis are described in Section 3.7. 3.8.3.4.3 Analysis for Ooenings w) Structural analysis will be performed around openings in walls and slabs ' particularly where concentrated loads from thermal ef fects are induced. Tho 31 design will account for all the stresses in those areas and proper 28 reinforcement will be provided for the relief of such stress concentration. 3.8.3.4.4 Liner Analvsis ( The liner-anchor system will be designed and analyzed in accordance with the requirements and criteria specified in paragraph 3.0 of PSAR Appendix 3.8-B. 37 Liner analysis is discussed in PSAR Section 3A.8.3.5. l 3.8-18a Amend. 61 Sept. 1981
37 3.8.3.4.5 Radiation Generated Heat Effeet N 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 AS E Code, Section Ill, Division 2 requirements. Where radioactivo piping penetrates the concrete walls, adequate Insulation or heat rcanoval measures will be provided to control the temperature of structural materials within Codo limits. Radiation generated heat will be produced as a f unction of the position of the Reactor core with respect to the structures and the temperature distributions w11I be calcuiated, in PHTS colIs, signifIcant heat w11I be generated from other sources such as piping. In all such Instances, adequate cooling capacity wilI be provided to lImit the temperature of the structural materials. 3.8.3.4.6 Reinforcement Deslan The reinforcing steel wit' be proportioned to meet the requirements of ACl-318. The bond and anchorage requirement of ACI-318 wilI be observed, since the Interior structures primarily provide a confinement function. The reinfcccoment 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 Desion The structural steel components other than the cell liner systems will be designed to t; e requirements of AISC specifications as identified in Section 3.8.2.2.1. When steel parts are stressed into 1he plastic range, an energy absorption check will be perf ormed to assure the functional Integrity. l l l l O 3.8-19 Amend. 61 Sept. 1981 l
(b) Tensile Properties Tension tests for both yield and tensile strength of reinforcing steel shall be performed in accordance with the requirements of paragraph CC-7331 of ASME BPVC 45l37) (Section III, Division 2). 61-(c) Bending Properties Bend tests of reinforcing steel shall be performed in accordance with the requirements of paragraph CC-2332 of ASME BPVC (Section III, Division 2) and the procedures in 61l 45l 37l ASTM A 370. (d) Measurement of Defonnations Measurement of bar deformations for each heat of steel will be performed in accordance with the requirements 451 of ASTM A-615-76a to establish conformity to the dimensional requirerrents. 3.8.3.6.4.2 Owner's Surveillance. All work of Contractor shall be subject to surveillance by Owner or its designated representative to assure conformance to specification 1 ("),1 (, 37 l requirements. 3.8.3.7 Testing and In-Service Surveillance Requirements Potential contact between sodium and components such as cell liners, equipment supports, anchors and anchor bolts, etc. can only occur following a sodium spill. All of the::e components will generally be located above the level of a sodium pool that results from a spill with the exception of the cell liner. Design of the inner cell system, including its functional re-quirements, is discussed in Chapter 3A. Location of anchors or supports below the sodium pool level will be minimized. Suitable insulation and seals will be provided around supports when hot sodium pools may be 'ormed. All exposed surfaces within an affected cell will be vissally 1 examined following a sodium spill prior to plant re-start. 37 n Amend. 61 3.8-22 Sept. 1981
3.8.4 Other Selsmic Cateoorv i Structures The Seismic Category I Structures other than the steel containment vessel and internal structures of the Reactor Containment Building are listed as follows: O 61l 1. Reactor Service Area of the Reactor Service Building (RSB)
- 2. Control Bullding (CB)
- 3. Steam Generator Building (SG8) 43 3:
O l i l l l [ 3.8-22a Amend. 61 Sept. 1981
- 4. Diesel Generator Building (DGB)
- 5. Emergency Cooling Tower (ECT) Structure h("' 4541 3) 6.Diesel Fuel Storage Tank Foundation
- 7. Electric Manholes 43 l 3] 8. Confinement Structure 3.8.4.1 Descriotion of the Structure; All structures, unless specifically noted, are designed to provide protection against seismic, tornado and flood ef fects for components and systems that are enclosed or supported by each structure.
3.8.4.1.1 Reactor Service Buildina The Reactor Service Building (RSB) plan and elevation drawings are shown in the general arrangement drawings in Section 1.2. The following is the description of the Reactor Service Building. The Reactor Service Building consists of a Reactor Service Arca (RSA) and a Radwaste Area (RWA). The Reactor Service Area is housed in a multi-story, reinforced concrete, Seismic Category I, tornado hardened structure. The Radweste Area is housed in a non-hardened, steel-framed structure, above grade and a reinforced concrete structure below grade (with the exception of the solid radwaste portion, above grade, which is of reinforced concrete). The foundation for the hardened RSA structure is a part of the common Nuclear Island mat for all Seismic Category i Nuclear Island Buildings whereas the RWA is founded on a separate mat. The RSA provides housing for the major portions of the Reactor Refueling and maintenance system, portions of the Direct Heat Removal System (DHRS), portions of several auxiliary systems, and portions of the Containment Cleanup System. The Reactor Service Area provides: (1) protection against soismic and tornado events as required for those components that contain new and spent fuel, (2) Intermediate transfer and storage facility for components, equipment and materials entering and leaving the Reactor Containment Building, (3) l radiation protection, primarily in the form of concrete shielding, (4) means l to facilitate material handling, (5) a safe means of entrance and egress for operating and maintenance personnel, and (6) space to house environmental 61 controls for the saiety of personnel. O V 3.8-23 Amend. 61 Sept. 1981
61 The Internal portions within the building principally consist of the cells and other features listed as follows:
- 1. Ex-vessel storage tank cell '
- 2. Fuel handling cel1
- 3. Cask shaft and corridor
- 4. Operator gallery
- 5. Compartments for inert Gas Receiving and Processing System
- 6. 125 Ton Bridge Crar,e
- 7. Ex-vessel Transfer Machine (EVTM)
- 8. Fuel Handling Control Room
- 9. Direct Heat Removal Service Components (DHRS)
- 10. Impurity Monitoring and Analysis Systems 61
- 11. ntalment Cleanup Sydem hponents Personnel access to the building is provided from the Plant Service Building.
Radiation shield walls for personnel protection of suf ficient thickness and suitable configuration will be provided throughout the RSB. Ordinary concrete will be used for this purpose with the exception of the Fuel Handling cell whose scJth and west walls will be of high density concrete. The Radweste Area mainly houses the Radioactive Waste Disposal System, the Maintenance Decontamination Facility, HVAC equipment, and a Motor Control Center. Although the RWA is categorized as a nonseismic structure, its design provides assurance that the adjacent Category I structure will not be damaged 61 in the event of an earthquake. O 3.8-23 a Amend. 61 Sept. 1981
3.8.4.1.2 Control Buildino 39l Os 6l The Control Building is a Seismic Catec,ory I, tornado-hardened, reinforced concrete Mructure extending down to the Nuclear Island common base mat at g & catior. 733'. The Control Building is a multi-story structure which has approximate inside dimensions of 122' in north-south direction and 75' In east-west direction. Exterior walls of the buliding are of reinforced concrete. Interior walls are of reinforced concrete and reinforcEJ Concrete block. The Interior floor and roof slab are reinforced concrete supported by structural steel framing and the exterior walls. The approximate overall height of the structure is 147' from the top of the common foundation mat. The Control Building will be structurally connected to the Intermediate Bay of the Steam Generator Building on the east side and to the Diesel Generator Building on the north side. On the south side it is separated from the Plant 45 Service Building. The Control Building houses the Control Room, its environmental system, and the routing areas for the cable network required to supply the Control Roan. In addition, the building houses the air conditioning units for the DGB and CB, emergency batteries, and the PHTS and IHTS motor generators for loops #1 and #2. The building is designed to accommodete the Control Room at elevation 816'-0" and to maintain separation of redundant safety-related and normal electrical cabling. Since both divisions of safety-related cables are 45 required in the Control Room, separation is maintained by the use of an upper and lower cable spreading rooms. Each spreading room contains only o? e division of safety-related electrical cable. A redundant capability to safely
) shutdown the reactor is provided at remote locations (SGB) so that loss of the d 61 Control Room cannot prevent safe shutdown. To ensure that the Control Room remains habitable in the event of a nuclear accident, Ilfe support facilities are provided and the iIving area is maintained at a positive pressure to prevent in-leakage of contamination. The batteries and their associated (quipment located at elevation 765' 0" provide plant safety related (including diverse power) and non-saf ety related uninterruptable power supply for vital electrical loads and D.C. loads. Three hour fire rated areas are used to 61 pr vide separation of redundant safety related batteries.
Detailed equipment arrangements are shown on the Control Building General Arrangements in Section 1.2. n v 3.8-24 Amend. 61 Sept. 1981
45l 3.8.4.1.3 Steam Generator BulldIna The Steam Generator Building is a reinforced concrete, Seismic Category 1, O tornado-hardened structure consisting of the Intermediate Bay, the Steam Generator Bay, and the Auxiliary Bay. The Steam Generator Maintenance Bay is a Seismic Category 1, non tornado-hardened, structure constructed of structural steel framing with concrete and metal wali siding. The Steam Generator Building houses equipment and facilities used in the production of steam. The major systems housed in the Steam Generator Building are the Intermediate Heat Transport System, the Steam Generating System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), the Auxillary Liquid Metal System, the Sodium-Water Reaction Pressure Relief System (SWRPRS), the Normal and Emergency Chilled Water System, and the HVAC System. In addition, the maintenance bay provides facilitles for cleaning, storage and repair of 61 compenents and equipment. The Intermediate Bay which houses the intermediate Sodium Heat Transfer System and the Normal and Emergency Chilled Water System is connected to the Control Building on the west side and to the Steam Generator Bay on the north side. The south side of the structure Interfaces with the Confintment Structure and 6 61 the two buildings are structurally interconnected from the foundation mat up to the roof at elevation 857 ' 6". The exterior walls and the radiation shield waiis for the three intermediate sodium loops and the primary sodium storage tanks are of reinforced concrete. Reinforced concrete floors are supported on structural steel beams and columns. O O Amend. 61 3.8-24a Sept. 1981
The Steam Ger. orator and Auxiliary Bays are designed to provide separation for O 61 the three inde,endent steam generator loops. The Steam Generator Bay consists of three cells och having three main floor levels. Each cell contains one of the three Independent steam generating loops. The south and west side of the structure are structurally connected to the intermediate Bay and Diesel Generator Building respectively. The structural system of this Bay is essentially the same as the Intermediate Bay. The SGB gantry crane runway 61 ralls at roof level are supported on the north and south walls of this Bay. 9 The gantry crane is capable of handling major equipment such as intermediate pumps, evaporators and superheaters, and transferring t!em to the Maintenance 45 Bay which is located east of the Steam Generator Bay. The portion of the Steam Generating System such as Water / Steam Circulating System and the Steam Generator Auxillary Heat Removal System (SGAHRS) is located in the Auxiliary Bay. The exterior walls and the floor system of the Auxillary Bay are structurally the same as the Steam Generator Bay. The Interior walls of the Auxiliary Bay, as well as that of the Intermediate Bay and the Steam Generator Bay, are of reinforced concrete with the exception of the wall of the elevator shaft and Cell 202B walls which are of reinforced block. The south side of tho Auxillary Bay is structurally connected to the 61 Steam Generator Bay. The Maintenance Bay is the portion of the Steam Generator Building containing a railroad siding and f acilities for maintenance, cleaning, and laydown of 1 Steam Generator Building equipment. This bay is a Seismic Category I structure but is not tornado-hardened. It is constructed of metal roof 47 decking and metal walI siding supported on structural steel beams and columns. 61lTheoverallapproximatedimensionsoftheabovefourstructuresareas follows: Inside inside Overall Length (ft.) Width (st.) Height (ft.), 48l 6 1. Intermediate Ba'/ 260 Varles from 124 17 ' to 162'
- 2. Steam Generator Bay 228 74 140
- 3. Aux!Ilary Bay 228 30 153 48l4. Maintenance Bay 84 84 108 61 The top of the foundation mat for the Steam Generator Building, excluding the Maintenance Bay, is at elevation 733' and grade is at elevation 815'. The maintenance area as well as the laydown area and railroad track of the 48 Main enan e Bay is founded on competent rock. The laydown area and rcilroad tracks are founded on Class "A" backfill.
3.8-25 Amend. 61 Sept. 1981 _ _ _ _ _ _ _ _ _ _ i
See Section 1.2 for the Steam Generator Building General Arrangements and the 48 45 general layout and configuration of the structures. 61 3.8.4.1.4 Diesel Generator Buildina The Diesel Generator Building is a Seismic Category 1, tornado-hardened reinforced concrete structure extending down to the Nuclear Island common base 45 33 mat at elevation 733'-0". The DGB houses equipment and f acilities used in the production of electrical power from the Emergency Diesel Generators, in addition, it houses the PHTS and lHTS sodium pump motor generators used to supply power to the IHTS and PHTS pumps in loop #3, and the switchgear and associated breakers for all IHTS and PHTS pumps. 45 The building is designed to allow separation of the two safety-related Emergency Diesel Generators and associated power production and distribution equipment. Another important function of the building is to provide an equipment removal path and routing area from the Nuclear Island Buildings to 6d the Maintenance Shop and Warehouse via the TGB. To fulfill this function, a corridor approximately 26 feet wide, is located along the eastern part of the bulding. An equipment removal hatch 11' 0" x 15' 0" is provided to allow for i removal of the largest piece of equipment from the SGB, G or DGB through the 61 hatch using the SGB gantry crane. In order to obtain a safe margin between the Diesel Generator Operating Frequency and the Diesel Generator Bullding Resonating Frequency, the natural frequency of the DGB is designed to be at least 30% higher than the operating frequency of the diesel generators. This is accomplished as follows:
- a. Floors at El. 816'-0" and below are supported by (3) reinforced concrete walIs.
45
- b. Floor at El . 816 '-0" is a 4'-0" thick concrete sl ab.
61 The two safety related, redundant diesel generators and auxiliary equipment are located at elevation 816'-0". The floor below, elevation 794'-0", houses the emergency electrical power distribution equipment and the diesel fuel oil pumps which transfer oil from the buried storage tanks outside. Elevation 165'-0" houses the breakers for the PHTS and lHTS sodium pumps and 13.8 kv and 4.16 kv switchgear. The base elevation, 733'-0" houses the PHTS and IHTS sodium pump motor generators for loop #3. Detailed equipment arrangements are 45 3.8-25a Amend. 61 Sept. 1981
Exterior walls, Interior walls, and floor slab at elevation 816' are of
/' ' reinf orced concrete. The other floor and roof slabs are reinforced concrete 61 supported by structural steel framing and walls. The south side of the building is structurally connected to the Control Building and the 45 Intermediate Bay. The east side of the building shares a cocmon wall with the 61 Steam Generator Bay of the SGB.
3.8.4.1.5 Emeraency Coolina Tower (ECT) Structure 45 The Emergency Cooling Tower Structure located approximately 700' north of the Reactor Containment Building, is a reinforced concrete structure which serves as a reservoir f or the Emergency Plant Service Water (the basir.) and will 45 43 house the Emergency Plant Service Water pumps and the cooling tower units. 3,0.4.1.6 Diesel Fuel Storace Tank Foundation Two diesel fuel storage tanks are located north and outside of the Diesel Generator Building. The diesel fuel tanks are buried and anchored to a reinforced concrete mat which is founded on and surrounded by compacted Class 61 A backfill material. The mat will also serve as a catchment in the event that an oil leak occurs in either tank. For detail s, see Figure 3.8-1. 3.8.4.1.7 Electrical Manholes Seismic Category I electrical manholes f or duct bank carrying safety related cables are placed at various locations within the plant site. They are f-~g relatively small reinforced concrete structures. They will be founded on and (,,) surrounded by compacted Class A backfill, and will be located partially 61 underground. An access opening in the top slab, at grade level, will be provided with a tornado missile shield cover. 43 39l32l 3.8.4.1.8 Confinement structure The Confinement Structure is a reinforced concrete cylindrical enclosure with a spherical dome. The structure is located external to and concentric with, 45 53 the containment vessel and is supported on the common Nuclear Island foundation mat. The overall dimensions are: Cylindrical portion (from El. 733' to spring line) -- 1.D.= 196 f eet, i and thickness =4 feet. Dome portion -- thickness =3 feet. 45 The structure f unctions as a tornado missile barrier, biological shield, and also as a protective enclosure against groundwater intrusion. it will be designed and constructed as a Seismic Category I structure. 33 l t l l l V ! 3.8-26 Amend. 61 l Sept. 1981 u -
-..-.4 -, ,. ,-. -, - .-- , , - . , , -,e . , -n- -. -.-.,v. -
43 l 32l 3,3,4.1. 9 Interconnection of All fluclear Island Seismic Category I l33 Structures to the Reactor Containment Building 43 l The Seismic Cctegory I building comprising the fluclear Island will have a common foundation mat. The buildings surrounding the Reactor Con-tainment will be connected to the confinement structure at all levels from the foundation to the roof, based upon the following considerations: 33
- 1) The overall structural stability against lateral loads, particularly from hydrostatic and seismic forces, is greatly increased.
- 2) The distribution of lateral loads from the operating floor level down to the foundation mat is improved.
- 3) The potential of groundwater intrusion is eliminated.
- 4) The flexible joints or connections for piping and electrical systems at the interfaces between the Confinement Structure 13 3 and other Category I buildings are eliminated.
O Amend. 45 uy 8 3.0-26a O
l i l 6 3.8.4.2 Applicable Codes, Standards and Specifications, The design and construction of all the Category I structures (except the containment vessel) are based upon applicable sections f the following codes, standards, specifications and NRC 61l Regulatory Guides. A.
^
American Concrete Institute (.' ACI - 301-72 Specification for Structural Concrete for Buildings (Revised 1975) ACI - 315-74 Manual of Standard Practice for Detailing Rein-forced Concrete Structures l ACI - 318-77 Building Code Requirements for Reinforced 45 Concrete 54 ACI - 347-68 Recomended Practice for Concrete Formwork ACI - 305-72 Recommended Practice for Hot Weather Concreting ACI - 211.1-74 Recommended Practice for Selecting Proportions for Normal and Heavy Weight Concrete (Revised 1975) ACI - 306-66 Recommended Factice for Cold Weather Concreting pd (Reaffirmed 1972) ACI - 311-75 Recomended Practice for Concrete Inspection ACI - 304-73 Recommended Practice for Measuring, Mixing, 45 Transporting, and Placing Concrete ACI - 307-69 Specification for the Design and Construction of Reinforced Concrete Chinmeys ACI/ASCE-333 Tentative Recommendaticas for Design of Composite Beams and Girders for Buildings-B. American Institute of Steel Construction (AISC) S310-1969 Specification for the Design, Fabrication and 45 Erection of Structural Steel for Buildings and Supplement No. S314. C. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Section III, (1974 Editions). U 45 p L/ Amend. 61
- 3.8-27 Sept. 1981
D. American Iron and Steel Institute AISI Specification for the Design of Cold-form:d Steel Structural Members, 1%8 Edition with 1970,1971 and 1972 suppimants E. American Society for Testing and Materials, ASTM Standards A-108 - Steel Bars, Carbon, Cold Finished Standard Quality E Surf ace and Burning Characteristics of Building Material 61 F. American Wolding Society (AWS) AWS D. I .1-79 Structural Welding Code AWS D12.1-1975 Reinforcing Steel Welding Code including Metal Inserts 45 and Connections in Reinforced Concrete Construction G. Crane Manufacturers Association of America, Inc., C.M. A. A. Speci f ication No. 70. H. American Railway Engineering Association (AREA) 32 1. Standard Building Code J. American Association of State Highway and Transportation Officials 61l (AASHTO) HB-11, " Standard Specifications for Highway Bridges",1973 Edition. K. Code of Federal Regulations, Title 29, CFR1910, accupational Safety and O Health Siandards, and Title 29 CFR1926, Saf ety and Health Regulations for 45 Construction. L. ERDA, Division of Reactor Research and Development, RDT Standards F2-2 Quality Assurance Program Requirements F2-4 Quality Verification Program Requirements M. NRC Regulatory Guides 15 See Section 3.8.2.2.3 for applicable NRC Regulatory Guides. 61 0 3.8-28 Amend. 61 Sept. 1981
_ _. _ . - - - - - _ - - _ . . - . _ - - .. ~. . ... . - . . . . - - . . - - . . . _ .. l 45 61
. N. U. S. DOE Manual '
s- . Appendix 0510 Prevention, Control and Abatement of Air and Water
- Pollution at Federal Facilities 4
Chapter 0505 Construction Safety Program Chapter 0550 Operational Safety Standards 45 ,
! 3.8.4.3 Loads and Leadino combinations 3.8.4.3.1 Loads All other Category I structures will be designed for the applicable loads I Irted in Subsection 3.8.3.3.1.
f 4 I i i J i 4 a l 3.8-28a Amend. 61-Sept. 1981 4
-v y- y ,--T-www,y,v- we e ,-cu- y- v v -- r e w . -,,-,-=w, er ,, wry, , y w w.m e y e. , wv, eyw.w w . ,vm, - ww ,., - e , e w, ., % e, w w w.,m e ,-+ wws ew,-,w,,, ,r,w w w,cyw -- e --v-,1,ww-,.--v-.w
3.8.4.3.2 Creep, Shrinkage and Local Stress Pre-stressed concrete design is not adopted for the design of the facility. Therefore, creep and shrinkage loads will be only considered to the extent they are provided in the referenced codes or as may be warranted by prudent design approach. , 3,8.4.3.3 Sodium Fire Load The cells, pipeways, and buildings where sodium fire is a postulated design basis accident, will be designed to withstand accident' pressure, and the associated temperature effects. 3.8.4.3.4 Hot Sodium Spill Effect The protective devices such as steel catch pans or steel plate liners will be provided in floor areas subject to sodium spills to prevent concrete-sodium reaction. 3.8.4.3.5 Loading Combinations All other Category I structures will be designed and analyzed for the loading combinations listed in Subsection 3.8.3.3.10. 3.8.4.4 Design and Analysis Procedures l 3.8.4.4.1 Analysis Procedures Classical theory, equations and numerical methods will be used as necessary in the analysis of the structures. Classical methods used in the analysis will be in accordance with standard textbooks, handbooks, and papers as used in engineering practice. The following computer programs . will be used in the static analysis:
- 1. NASTRAN
- 2. MARC CDC 44 3. MRI - STARDYNE 12 4. Other in-house computer programs 4
Loads and loading combinations as delineated in Section 3.8.4.3 will be considered. For dead loads, live loads, wind loads, tornado loads and accident loads, all of the methods listed above will be used. Wind loads, tornado loads and accident loads are converted to equivalent static loads and will be applied to the structure as uniform or concentrated loads. Wind and tornado loadings, flood loadings and missile loading applied on structures are discussed in Sections 3.3, 3.4 and 3.5. Amend. 44 3.8-29 April 1978
, -- ,.w. . . . . _ ,.m._. ..-..-_._.y,-. _,_,._.r + --r- -->e-*-v ~ e---- -&* - - - --------+=r
f Seismic analysis of the structures is covered in Section 3.7. The mathematical models will be as shown in the aforementioned section. Equivalent static seismic loads, as defined by the dynamic analysis, will be transferred through the horizontal slab diaphragms and vertical shear walls to the foundation mat. ' Walls, floors and columns are the basic structural components of the buildings which will be analyzed to carry and transfer the gravity and vertical loads to the common mat which is founded on rock. The lateral loads will be carried through horizontal diaphragm (floor) action to the vertical resisting elements (walls) depending upon their relative rigidity or stif fness. Torsional effect will be considered in the distribution of the lateral loads. All Seismic Category I buildings of the nuclear Island including the Confinement Structu"e, RW, RSB (with the exception of the Radweste Area), CB, DGB and SGB (with the exception of Maintenance Bay), will be on a common mat 45 which is founded on rock. The design and analysis of the common mat will be performed as described in ' Se i n 3.8.5.4. 6 3.8.4.4.2 Deslan Procedures Design procedures will be in accordance with the applicable portions of the codes, standards and specifications listed in Section 3.8.4.2. The results 61l derived in Section 3.8.4.4.1 wilI be used in design of structural steel and reinforced concrete. Reinforced concrete structural elenents will be designed by the strength 54 method in accordance with ACI 318-77.
"tructural steel frames or components of the buildings will be designed by the elastic analysis method in accordance with the provisions of the AISC Specification for the Design, Fabrication, and Erection of Strut tural Steel for Buildings.
l Classical methods used in the design are stanc:rd textbooks, handbooks and l l publications as used in engineering practice. The basic design approach for the structures is given in the following subsections. O 3.8-30 Amend. 61 Sept. 1981
l
, 3.8.4.4.7 Reactor Service Building sd 3.8.4.4.3.1 Reactor Service Area This portion of the Reactor Service Building is designed as a Seismic Category 61 I structure and will be analyzed as a multistory reinforced concrete structure made up of slabs, columns and walls. Interior walls that are required for shre lding will be used as structural walls. Lateral loads will be resisted by shear walls with the floor and roof slabs acting as diaphragms.
The overhead crar,a runway rails will be supported on reinforced concrete brackets off the structural concrete walls. The crane will be designed as Seismic Category I equipment having such features as redundant receiving holsts and brakes. See Section 3.8.4.4.1 for the foundation design of the 61 building. 3.8.4.4.3.2 Radwaste Area The Radwaste Area of the RSB is an independent, seismic Category lli designated structure. it consists of a steel framed structure above grade and of reinforced concrete slabs and walls at and below grade level and partially above grade (Solid Radwaste Area - on the east side). The Radwaste Area is supported by a reinforced concrete mat which is founded on sound siltstone with adequate bearing capacity. The foundation for the west end of the Radwaste Area is at grade elevation and is founded on compacted structural backfill.
,O The Radweste Area structure is designed to meet the requirements of the Standard Building Code. In addition the structure below grade as well as the Solid Radwaste Area above grade are designed as a reinforced concrete structure.
4 The upper part of the Radweste Area, the steel framed structure is designed to ensure that the adjacent seismic Category I structure of Reactor Service Area is not camaged nor its safety functions compromised during an SSE. 61 3.8.4.4.4 Control Building i As described in Section 3.8.4.1.2, the Control Building is structurally connected to the Diesel Generator Building and the Intermediate Bay of the Steam Gcnerator Building. The building is a box-type structure with exterior walls of reinforced concrete and intermediate floors and roof of composite construction made up of reinforced concrete slabs on structural steel framing. < i The Interior structural steel columns are designed to carry vertical loads while the exterior walls, roof and intermediate floors are designed to resist vertical as well as lateral loads. See Section 3.8.4.4.1 for the f oundation design of .ne building. i O v l 3.8-31 Amend. 61 Sept. 1981 < l l t-_.--____-_-__-_-_---__-_____-_---.____-_.-_
3.8.4.4.5 Steam Generator Buildina As described in previous Section 3.8.4.1.3, the Steam Generator Building Is O structurally connected to the Confinement Structure, Diesei Generator Bullding and the Control Building. The Intermediate firors and the roof of the Intermediate, Steam Generator and Auxillary Befs will consist of composite structural design (structural steel framing anc reinforced concrete slabs) on structural steel columns designed to carry the .ertical loads. The Maintenance Bay wilI be designed as a space f rame with beams and columns resisting moments and shear loads. The Intermediate floors and the roof, In 61 conjunction with reinforced concrete exterior and interior walls, will resist the vertical and lateral loads. The gantry crane, which is Seismic Category 1, is supported on the roof of the 45 Steam Gen rator and Maintenance Bays and the Diese! Generator Building. As described in Section 3.8.4.4.1, the reinforced concrete mat for the buildings will be founded on the rock. O Anend. 61 O 3.8-31a Sept. 1981
3.8.4.4.6 Diesel Generator Building O The Diesel Generator Building is a box-type multistory building which has reinforced concrete interior and exterior walls. The roof and intermediate floors are reinforced concrete slabs supported on structural steel beams with exception of the floor slab at El. 816' which is completely of reinforced concrete. The vertical loads will be resisted by the floor and walls; and the 61 lateral loads will be resisted by the floors designed to act as diaphragms and 4 j the exterior walls to act as shear walls. The concrete walls and slab will be 44 designed to support the diesel generators with consideration for vibration Induced effects from the equipment to structure. The building is founded on rock as described in Section 3.8.4.4.1. 3.8.4.4.7 Emergency Cooling Tower Structure The Emergency Cooling Tower Structure is founded on rock and will be analyzed 43l g and designed as a circular basin with a cooling tower superstructure above. 3.8.4.4.8 Diesel Fuel Storage Tank Foundation The diesel fuel storage tank will be anchored to a reinforced concrete mat. The anchorages and the mat wilI be designed for earth loads, seismic loads, J tank loads and hydrostatic uplift loads. 3.8.4.4.9 Confinement Structure The confinement structure will be analyzed as a shear wall below the roof level of adjacent buildings. Above this level the structure will be analyzed as a concrete shell. Loading induced from the shell will be carried through the structures below as shear walls to the foundation mat. The exposed portion of the Confinement Structure will be designed for tornado missile impact and the analysis will include all loadings for Seismic Category I structures. . 33 , 3.8.4.5 Structural Acceotance Criteria The design criteria relating to stress, strain, gross deformations and f actor of safety are identical to those described It Section 3.8.3.5. 3.8.4.6 Materials. Quality control. and Soecial Construction l Amend. 61 3.8-32 Sept. 1981 I
3.8.4.6.1 Materials Concrete All structural concrete work will conform to ACI 30?-72 (Revised 1975) and the concrete will have a minimum compressive 45 strength of 4,000 psi at 28 or 90 days. The high density concrete, which will be used for shielding purposes, consists of Type II Portland Cement ceniorming to 45 ASTM C150-77 and heavy aggregates conforming to ASTM C 637-73 (Specification for Aggregates for Radiation Shielding Concrete). The density of the proposed heavy aggregate concrete will be a minimum of 210 pounds per cubic feet. The compressive strength of the high density concrete will be 4,000 osi, which is identical to the normal concrete used for all of the Seismic Category I structures. 33 Reinforcino Steel All reinforcing steel will conform to ASTM A-615-76a, Grade-60. When reinforcing steel is required for arc welding, the material 45 shall confon?. to ASTM or ASME standards.. Structural Steel 45l Structural steel shapes and plates will conform to the require-ments of ASTM A-36-75. When special type of steel is used to h meet a specific requirement, the material will conform to the ASTM or ASME standards. Amerid. 45 July 1978 3.8-32a
3.f A.2 Quality Control A formal quality assurance organization and reporting system will be employed to assure that Seismic Category I Structures will be built in accordance with applicable codes, standards and specifi-cations. The responsibility, coordination and monitoring of quality control functions of.the cognizant organizations are outlined and defined in Chapter 17.0, QUALITY ASSURANCE. The quality control standards and inspection requirements for concrete, reinforcing steel and structural steel, as described herein. are also applicable to "All Seismic Category I Structures." i The materials to be used fcr the structures, together with the quality control standards and inspection requirements during con-struction will be described in Burns & Roe's Material and Construction Specification. The following is a general description of quality j assurance requirements. 3.8.4.6.2.1 Concrete All concrete materials will be purchased in accordance with Burns & Roe specifications t.nd tested by the testing laboratecy 4 45 for conformance and acceptance. The quality of all concrete materials will be periodically checked by the testing laboratory during the progress of construction 45l I to assare continued compliance with the specifications. i The procurement of production concrete will be from a supplier who will be responsible for the quality control of the concrete includ-ing the required sampling, inspections and records to confirm that the concrete does in fact meet specification requirements. The supplier 45 will be required to provide c:stified quality control personnel to conduct quality control and inspection program. In depth surveil-lance of the concrcte supplier's Quality Control Program will be performed by the constructor to verify acceptability of aggregate, 45 cement and concrete samples. Constructor's Quality Assurance Personnel will perform surveillance of the quality control functions of the staff assigned , 45 to the testing laboratory. . j Batch plant inspection will be provided by constructor with 45 sampling and testing as follows: -
- a. All ingredients wilI be :.;ampled on a planned basis by constructor and submitted for testing to testing laboratory.
- b. Ccnstructor will certify the mix proportions of each batch and will prcvide a delivery ticket for each batch docu.enting 45 the date; time loaded; mix proportions of water, cement, fly ash,
~
aggregates and admixtures; concrete design strength and
- identification of transporter.
Amend. 45 July 1978 3.8-33
Constructor will Inspect each load at the point of discharge for slump, air 45 content and temperature. Weather conditions records will be maintained during placements. Sanples of fr'esh concrete will be taken and delivered to the testing l aboratory for casti.'q, curing and testing of strength test cylinders. Test cylinders will be made for each 100 cubic yards or fraction thereof placed in any day. As a minimum, three test cylinders will be cast for 28 day strength concrete; one cylinder is tested at 7 cays and two cylinders are tested at 28 days. Three test cylinders will be cast for 90 day strength 61 concrete; one cylinder will be tested at 28 days and two cylinders at 90 days. Unit weight of concrete representative of each set of cylinders cast will be 4Sdetermined at time of casting cylinders to measure and verify the radiation shielding csnsity requirements. Concrete cylinders for compression testing will be made and stripped within 24 4yhoursaftercastingandmarkedandstoredinacuringroom. These cylinders will be made in accordance with ASTM C 31-69 (Reapproved,1975), Method of Making and Curing Concrete Compression and Flexure Test Specimens in the Field. 4g Compressive strength tests wilI be made in accordance with ASTM C39, Test Method for Compressive Strength of Molded Concrete Cylinders. Slump, air content and temperature will be taken when samples are submitted for strength test and for first batch placed each day and every 50 cubic yards 45 pl aced.
, Slump tests will be perf ormed in accordance with ASTM C143-74, Standard Test 45 Method, for Slump or Portland Cement Concrete.
Air tests will be performed in accordance with ASTM C231-75, Standard Test Method, for Air Content of Freshly Mixed Concrete by the Volumetric of 45 Pressure Methods. Weight /yleid test will be performed daily during production in accordance with 61 ASTM C 138-77, Test Method for Unit Weight, Yield and Air Content 45 (Gravimetric) of concrete. Inspectors for the contractor placing the concrete will Inspect reinforcing, forms and concrete placement. in the event that concrete is placed during l f reezing weather, or that a freeze is expected during the curing period, additional strength test cylinders will be made for field curing in accordance 45 with ASTM C-31-69 (Reapproved 1975). The evaluation of the strength test results will be in accordance with Chapter 45 17 of ACI 301-72. l 3.8.4.6.2.2 ReInforeIno_ SteeI 47lThequalItycontrolstandardsandInspectionrequirementsofreinforcingsteel are described in Section 3.8.3. O 3.8-34 Amend. 61 Sept. 1981
3.8.4.6.2.3 .S.tr_nctural and Miscr!Ianeous Steel For ali material used as structural steel, milI test reports giving chemical composition and physical properties will be obtained for approval. Fabrication and erection will be ir accordance with AlSC specifications and Section III of ASE Code. Inspection will be contcted at the f abrication plant as well as in the field. Welding !nspection will be as outlined in Burns and Roe's Structural, Material, and Construction Specifications. 3.8.4.6.3 .Sp., , Construction Technlaues
- The structuras will be constructed using normal construction methods and techniques.
3.8. 4 .7 Testino and in-Service Surveillance Reat'Irements Ther e are no testing and in-service surveillance requirements for the structures. 3.8.5 Foundation and Concrete Sucoorts 3,8.5.1 Desc.-Iotton of the Foundation and Sunoorts 3.8.5.1.1 General Descrlotton (3 Q The foundation for the following Seismic Category I structures consists of a combined reinfcrced concrete mat laid out to envelope these structures. See Figure 3.8-2 for the mat layout. 32 I (a) Reactor Containment Building (RC8) (b) Confinement Structure 45 (c) Reactor Service Building (RSB), excluding Radwaste Area (d) Cor. trol Building (CC) (e) Diesel Generator Building (DGB) 45 (f) Steam Generator Building (SGB), excluding the Maintenance Bay [3 v The thickness of the combined mat based upon the stress and stability 61l considerations is 18' except under the RC8 where ! - Is 15'. The mat slab rests up;n a firm rock strata and its bottom is l'cated at El. 715 which is 100 feet below the finished grade. A fill slab .:f suitable design is placed 6 ver the RCB portion of the mat. 4 33 The Emergency Cooling Tower Structure will be supported by a reinforced 3 '. concrete mat founded on competent rock. The SGB Maintenance Bay foundation will be on competent rock. The diese; fuel oil storage tank foundation, the Emergency Plant Service Water System supply and return headers in the yard and the underground Class IE electrical ducting and Category I pipe will be founded on and surrounded by the compacted structural backfill. O 3.8-35 Amend. 61 Sept. 1981
3.8.5.1.2 Design Features The combined mat concept for the Seismic Category I structures noted in subsection 3.8.5.1.1 rests upon two basic considerations, namely; (a) to reduce seismic responses at component supports, and (b) to minimize buoyancy effect due to maximum flood on relatively lighter portions of the foundations by combining them with heavier portions. Several provisiont will be included to prevent the intrusion of groundwater or 33 61 flood water into the steel containment. They are:
- a. A continuous membrane waterproofing system along all external building faces to grade.
- b. A thick reinforced concrete structure designed for crack control with continuous waterstops provided at alI external construction joints below plant grade.
For Seismic Category I buildings, appropriate drainage systems, where required, wilI be providad to dispose of any potential intrusion of groundwater. Flood water protection provisions are discussed in Section 3.4.1 of this PSAR. 3.8.5.1.3 Load Transfer The loads from the superstructure will be transferred to the foundation mat via structural elements such as columns and walls. The load transfer structures will have necessary configurations, and strength to meet the shielding as well as structural requirements. Since the mat is founded upon competent rock, no relative local subsidence is expected. Therefore, no adverse effect due to this factor will be considered in the foundation design. The mat analysis will provide for all intarface loads arising fue to the interaction between the mat and connected structural elanents such as walls and columns. 3.8.5.1.4 Large Eculoment Suooorts The Reactor Vessel is supported on a steel ledge, partially embedded in the RV cavity walIs. AlI vertical and Iateral forces on the RV wIII be transferred 61 through the ledge to the cavity wall and the foundation mat. O Amend, 61 3.8-36 Sept. 1981
A- 61lTheprimarysodiumpumpsa..Jintermediateheatexchangersaresupportedfrom the operating floor by cannister type supports. The operating floor in the neighborhood of these equipments is supported on structural framing, which in turn is supported by the cell walls and the interior perimeter wall. The equipment loads are thus transferred to the foundation mat via cell walls and the perimeter walI of the RC8. 45 The superheaters and evaporators wilI be supported by structural steel floor framing which will be supported on reinforced concrete walls and columns. Lower lateral seismic supports will be provided for the superheaters and evaporators. The intermediate sodium pumps will be supported on reinforced 61 cor. crete brackets attached to the walls. 3.8.5.1.5 Rainforcing Pattern The structure will essentially act as an assemblage of shear walls under lateral seismic forces. The shear reinforcement will be used where shear stress in the wall due to lateral forces (seismic and soll pressure and hydrostatic pressure) exceed its shear capacity. Any flexural load at the wall / mat interf ace will be resisted through the vertical rebars. The reinforcing pattern at the above Interf ace wilI be conventional type dowels 61 projecting from the mat and lapping with the vertical reinforcing bars in the wall except as noted below. Horizontal reinforcements in the wall will be provided as required by temperature, flexure and other considerations. The vertical rebars in the RG interior walls will be developed through the fill slab by use of hooked bars. Sufficient amount of reinforcing steel from Containment peripheral wall is anchored into the foundation mat to overcome d the seismic overturning moment of the RW Internal structure. Mechanical 61 splices are used for stress transfer through the Containinent bottom liner. 3.8.5.2 AnolIcable Codes. Standards and SoectfIcations The applicable Codes, Standards and Specifications are identified in subsections 3.8.3.2 and 3.8.4.2. In addition, the following reguletory guide is applicable to that portion of the mat considered as part of the 4E mtainment. No.1.19 Nor Destructive Examination of Primary Containment Liner Welds. l The portion of the combined mat within the limit of RCB provides the containment f unction in conjunction with the bottom Iiner. Therefore, this portion of the mat and the liner will conform to the requirements of the ASE Boller and Pressure Vessel Code, Section lil, Division 2. The remaining l portion of the mat, starting from the outer f ace of the confinement structure outwards wil l conform to ACl-318, "Bulldins, Oode Requirements f or Reinforced 61 Concrete". l f l Amend. 61 3.8-37 Sept. 1981 1
O 3.8.5.3 Loads and Loading Combinations 3.8.5.3.1 Loads and Loading Combinations for the Portion of the Mat Below RCB The portion of the mat within the confines of Division 2 of the ASME Boiler and Pressure Vessel Code, Section III (see Section 3.8.5.2) will be checked for the service and factored loads as defined in paragraphs CC 3220 and 3222 of that code. The loading combination will be checked in accordance with the requirements 45l of Table CC-3230-1 of Division 2. G l l l l l Amend. 45 July 1978 l i 3.8-37a
29
.O 61 3.8.5.4.2 Reactor Vessel Sucoort Structure O The Reactor Vessel (RV) designer wilI establish the forces on the supporting steel ledge resulting from a seismic event by performing the necessary dynamic analysis. The seismic loads transmitted to the ledge support will be 61 transferred to the combined mat through the RV cavity shield walls. See subsection 3.8.3.4.1 for discussion on the analysis of the cavity. Details of the design analysis will be provided later.
3.8.5.5 Structural Accentance Criteria 3.8.5.5.1 Stress See discussion under subsection 3.8.3.5.1 3.8.5.5.2 Strain See discussion under subsection 3.8.3.".2. 3.8.5.5.3 Gross Deformation See discussion under subsection 3.8.3.5.3. 3.8.5.5.4 Differential Settlement See discussion under subsection 3.8.5.3.1 3.8.5.6 Materials. Quality Control and Soecial Construction See discussion under subsections 3.8.3.6, 3.8.4.6 and Chapter 17. Portion of the containment structure subject to the requirements of ASE code, Section 61 111, Division 2 will meet all testing requirements of that code. 3.8.5.7 Testina and in-service Surveillance Reautrements There will be no in-service surveillance requirements. l O 3.8-39 Amend. 61 Sept. 1981 l
References to Section 3.8
- 1. Appendix 3.7-A, Seismic Design Criteria for the CRBRP.
- 2. Chapter 3, SEQUOYAH Nuclear Plant FSAR Docket No. 50327, Appendix 3.88.
- 3. Specification for Electric Overhead Traveling Crane, Specification No. 70, Crane Manufacturers Association of America, Inc. 28 O
l l l l l 3.8-40 Amend. 28 l oct. 1976
O O O TABLE 3.8-3 4 STRESS LIMITS FOR STEEL CONTAINMENT VESSEL Load PRIMARY STRESSES General Local t$encing eius Primary And Combination Region Membrane Membrane Local Membrane Secondary Peak Number Of Vessel (Pm) (PL ) Stresses + (Pb+P) L Stresses (1) N/A Sm 1.55m 1.5Sm 3Sm N/A (2) N/A .85S y 1.25Sy 1.25Sy N/A N/A
." (3) N/A Sm 1.5Sm 1.5Sm 3S m Consider for 9 Fatigue Analysis "l (4) (6) (8) N/A Sm 1.SSm 1.5S m N/A N/A
- (5)(7)(9) Not Integral Sm 1.5S m 1.5Sm N/A N/A
& Continuous . Integral & The greater of The greater The greater of
! Continuous 1.2Sm or Sy. of 1.8S m or 1.8Sm or 1.5Sy N/A N/A 3 1.5Sy l , NOTE: Buckling allowables, factors of safety and definitions for local areas are provided in Table 3.8A-2, , 47 " Buckling Analysis for Steel Containment Vessel" page 3.8A-12a. EE F !B m ." N 4 4
1 O
-,. _.I I ,
,A
!V ' - -
""\.a
\. ...
.. ~
s ., i
; i
.. , ; . =..g,.,, y .i. n -f.. _ 4 j
,; -+
,- i g ..-e.. : ,
t 1-- - - - - - . - _ _ _ . . . _ _ _ _ . . .
' 'y,,- 7,, ;
/
I f [.' }
\ [._ ___ .
's
/ ~
y/ I
, p!. . . ... / . .
3 :: i y !, y .. p tq 7__. , I , e l' I/ I I f f f f f
= ____ _. . _ _ .
T k f
, , - . _- - \_ ._. _ . . _ . . _ _ . - . -
s,. -. .
- d: . .
!l
.. ll
! ' trd
, , !d
'L
- " .JaL_. n s .' =L su , _ _ u!
=, . 2_3,}l DIE SE L FUEL Oil TANK FOUNDATION AL AN
_**a_ e.:, mamew-g ;, ;s,w g- -- 4,-' ..> _
= oO4-., < e=ame ."*
- 2-is. #.o .
. ?- - + .
- 1 ynwwns.
,/,/
I Q U'* ! 1 EL 1 *- 0
- mg. 7 y.,.
1,.
.* i sc., .
j / /9.# a3 4b GRAPHIC SCALE h -qn,* f a'u , . . q;,,,, j '
, , , , *@'__.-;,,,,t_.-,*
- .a =P -.~) ,
, " ' ,* , i EJ stau 'a' N >
.r-... .
re ... . .,,,,,,,,,,,,.,,,,,
- m. .
SECTlON A-A ** * = *s we Ov.J Figure 3.8-1. Diesel Fuel Oil Storage Tank Foundation Amend. 61 3.8-48 Sept. 1981
o 4
@r- @ @ @ -@ @D 9-
- e. r 3 11 l
__ s.y i l rw rr-e- nr
, n:(
r. N PLANT @ t
- [ ;
- g. _ _ ._ .,
hg I Co=reet stos een 3 ~ su ;,ge.. 5.a o<
@---+- .
l (_ m 4 , l <te m y l g
=
_.-g
@@ @ @ @ @c; . Ir1,,& L_._ u c aam ua -
.SS'tgij*Lsw 58 . .-
-- g
," - b--
_ , , , , , , f,'
). l, ie-r g ,_ . _
- r. [= '
l
/ l, i h, 6 p-
} / I , .
/ .
de ', a V
. :c . . . . . . .
sam
- mt c
3;
& c, j
y *: u,---.'- n= sr k . Q j g e4 - - I h_ _ m _._ :z , ,i- m {- -
, 41 O
. 7) i e /~m s _.
a :: 0
- I : i J. 7r m: - \
.' m,7 , 3d.
p: @ z.- 77- tjy ...
.,,,e7,3 p o s - - i Va r3 a, / -.
~
\
'l j.'.t' g
,' ',s kmJ . h, . '.
g
........um.
.awan arm--!- -a --
.'s . .
r G.-. <
'a' "n.'. 4y
-18 0* b l
i - n GE if
. g, s( Nq' 1 ..
.y:.
*=" h
,f z;
@r k ij -
s je
- 0. . . ,,. ! ... ..
y
.. \ V - -
tI -Q c A GE l l y8 ,,,,, y-roe c= ree st d" '". cr. . s e,== rie-v 7 h-
- e. /
p q ,. u
^
i 3 .. . ..
./ @. ., : :.
idl n@-~ -e- Fr 4 '4 x,,j, -- 3 Ljc @,,f 7l %,r8l
-,e p i
-.-] ,
c , +g 4 1 1
.,D . n
, , wl mw =Tl' _ast ne. . . . . .- r** , ,,
10 e- *{ y; O"D) _gg u,
.- * ' 'i i t' 5 '
,g, s m m o B,ys=
g , _ ,o. , g-l I'. I j /
/
Jg !g- )
,. j i
_a - n . n. -,- : i i
- a. . .
- n g -g rr r b er ,.
f r..___rer - m f I str 2s a err J re! .r r ow rr r d, T wr I I iser i sr.c re
, I lA _er:r - _ ns.e h h h (".6) 'h 90*
1, DEAC798 C0ssTAsuwEssf SLD4 b k
@ @bVS @
PL A N AT EL 733' O' 0 0S (sis DCB i SG B crus) scale .eo e root otrastroseoEwaroe sea ovits.e owes messee=ceo oee PLA= has no CowTROL SLDC 4 SC B att Snowse By SvuEA TO *aDICATE
/ wie.Ces AR*AS udSLDG LOWER C4 4AREA UP. # C.TEM OVERS Cowuose WALLS SETWEtes SLDGS C8t
[] DdDi8"aftS w&LLS CwSRED t'. f CDedTeck DLDG gg MEDICATES WALLS Cove #ED ey ass etLLs 4 AUE BAY NS520-5
M e I 1 ' s
; +..g-
@ hi ,j -
,s.e ----
s :4s tsu _ <<<r .sce std
,!.ib__te: s** @M \
l
. ef o!. . *c "
d*
- I b
!J--
'TLp 3-
-Q l wi a_ '.5+ m w 75 %
n jes.l
.p
'A j ; _ si d mid b6N7i-"I . s' E U.* ue l l3 I
! J P
t-Fg p,.:g2,%,;pJ e ,coso
-.a o,. s .a i -
b ::4 , g m , a 7 j f
--44 ,[~_ 4-g gv. @
.,y@4 q 3g_:
'ei
--4-- g A ne, I I h*
r i m
,s .
w tv. [R - ' e
%" gKt ,g.a A. ,; ,y, l' ;, 'i'. 7 ...~...!
_._g ,h g g v
r g --
. c _ n. .- g s l@ .
ri t 6 U ~& f@' f ' 1
*T' _ti w b .,
F',4T T !
~-
1____ A ,, ~*s,i i sts j
$='" ;.. LI -
L -
'g,'ss a J'. Y ! 'O-k- LT- .$1 ! Md )*4
[ s 3 *g m.e __ si 4 ,e 4 i ,o d i cs a t-s; I es wom., v, -. a. ,o a_ -,
-t-=1
, v -
Si;
- a. ovs . .so .,w an o., v os ca se newso ei. >=n, r.c o 4.s s .o. o a
, fj.,g 7 ? b e o.s *%cu o-m se eo Mi 3 Jo4%es wo* uan<ao av t.T*r a os wa aeova cAu na moves s Tuse s,oe os
, ; q [ g T.a nmT av a o s Aw , .ot to aacs.r.o r.s*
', 4- SC Two AD. ACE st .,,,a3 S4NL3 SEpowmED S Mu/A%EOuSLY j i e' 1
-----=-_---- - - - - . -u 7--- ,C s .m m.ccEs
,~ei cen .a.nta
. er o-. Aws = ces m =ots "r u m y 3 e,ss moica+ss wou. oowa owe ho
,,, ; , g ,
7 _f. p r COUR PLAN-UPPER LIFT / sot ELW4 Top et. tsf-0) vos
.- .sD x Ama . . a- s
- e. . _ _ _T s...b' RAI
* *'558
,,.t.
6; ggse _ ss e-gaca ss.,e we 3.g py.
-f 7 l muoar- -
*L.f
- o. h e'p..M, 4p<
..,-,,,,s _.
=7
( "' ;' - 5 M' I ci r;
~'-~~
,o d _ ,o s La'.-- -
= 7
" Ji 'r
Or M' b* [ERATOR l3
+
N t
-h ;m y\ ': a f' si en .
sG %, 1 . T zi\
) ,s ,
$'a a .-
g ( 3 as- ;/ ) w p 4. g w o.r ),l$'a'*"
+.>
*c H- ) L, p' [iB ls ~ ~ ^
h. M e; a I .. 'Ti OsH v T; 8 Th( .cc.
"- l @ <is i & t..
s _t M'* ~
# J _._ +
y
.) >
ce,_ )M.7' f -0. f
< v% e os [ 31 s g
s
,a
_ b SM,- f . f i- 0' ,a, e, ' D g j _L __ l .d .4_ _ . _ _ _ Q , t,
.., ; (
- w. ,
3 +.
~ '
9- . I [.L2.d isse _.EG.Tp/,1.'
~+ - "
sK w.o- se s- ,o s 1
,c d . ss.e s ns I. . ss.,-
g y ra. u ,,;- s i i -
,, 4 y 6, n d
w, l4Tm .1 , i nc__ser.: rrhe are .
) Il 3 2r _e_
gIE POUR PL AN - LOWER LIFTreet<>s'oTerrt nte) em e so s sorri \l L-ettv pri lI
%(._ _
) 3 h.__as.r__h' ,'
GRAPHC SCALE
. Figure 3.8-2 P Nuclear Island Structures -
Common Mat Plan 3.8-49 Amend. 61 l Sept. 1981 g 5
l . O 6'- 4"
- EL 6 i6'-O" 9F if ea..
o s ;> e . . . s j ._.s. .s .,s . . a g 3 . . V V A A 5'-3"
- EL 734' O"
- 1r v
. *^.1 1
.,- 7 .-
a. y * . y B O a 5 -6,,
- EL 752'- ll" y v n
.t
.1
^
.I f- '.:
*A
-et 733'-o"
% ;.s .- v o ., , .'
.s.s- a,, s , ,s-s , g ,s , ,
.,,.s. .
. p ,
- FIGl':<E 3.8-5 TYPICAL VERTICAL CROSS-SECTION OF PHTS CELL O Amend. 61 3.8-52 Sept. 1981
*<----,,,r., , ,--..,-,.-.,-,,...e--.,-.,-_-.....-y.nnw,.-~,,-.,-,m. -,.a,.--, .,.--m, e,-- - - - - r. .,.-n~., -, ,
I 37' 6"To 4. RCB a. n 4.7'-4'[ . v O A y// , /////i'// V y V
" s' ,%' k :a.
. o>
a >l, 4 $'.4" 6 t O ' . ' .'g.0"
~
0 m g.c30' 0 8TO(RGB l t i
-v
- 6 RCS---
l
,4'-3" N
IS v *A 5 '- 7"
.# if
) '.~,..,y'
, 3 '.
cn %wm u ,y
'///7773977 1
, 43'- 9 E' .
w l FIGURE 3.8-6 HORIZONTAL SECTION A-A 0F PHTS CELL O 3.8-53 Amend. 33 Jan. 1977
t i a
~
32-6* # v
- i. .
6'-O" u a >:> , , s ;s: ,
.. . g. s .- .
a ,- . Q 3 n , b 5/ l , l O si , O - cn ; 5'- !" ) dp RCB u
- a 4 -10 i
} N 8
'J) v .
. s. 5'-7" ,.'.
.a ,,
. if a- ,"
.s.,.-
s / y ,,
> *.s. r.. .a .a-,.. .'.
s.
^
3 _ /\
'VH/'
t l l FIGURE 3.8-7 HORIZONTAL SECTION B-B 0F PHTS CELL 3.8-54 Amend. 33 Jan. 1977
1 l REACTOR CAVITY SE AL REACTOR VESSEL SUPPQRT
-- p. .- .: :_. p
~' -, RE ACTOR VESSEL E LEV. 802*-0" . _ ,
;;'- 3 -
.-).'. ,.,{ ..l- ,
.s s_ s s s s s s s a
,: . t - 3-i' I d SUPPORT
- SHIELD
- 1. s:: , RING s ._. :
r - -+m m' y
- c. s N
x s . EL EV. 791'-5" .\ i
<-- . (
b e TOP OF CONC 7 , l ,' f / l.', EL. 786'-3" 7 - i i; it ' /. i ,- l g,". , . ' i + GU ARO VESSEL PIPE I-
/l O, - ,
l l
, / j(
,\ .
-),
I ..,::. s 1 ,
., s ti x w FIGURE 3.8-8 Section Through Ledge Support At Cut-Cut' 3.8-55 Amend. 61 Sept. 1981 g
E L EV. 816*-0"
- 3.
fM.' . '
-ji,
,' REACTOR VESSEL REACTOR ,(('3
CAVITY 5 .s SEAL t i}i
$':I REACTO R
,,,,, O VESSEL SUPPORT '
7 :] REACTOR VESSEL [/ E L EV. 802'-0" s-
. % ff ~ !k;$,I',.' - ,
w/esw/a/swww/www - - I ; : ,9 [ '[ S c l (. l,- .
- s . '.l. E h[ l 6' '$ >
SUPPORT f : , ,$.-l; ,
.. ':ti s SHIELD 6 '
-4> ,
{ RING / L Lcr l r'e '
/
[ 'y sit.9' s .' j- ; - '
/ ' :
- . .t.
1.. ; e U: F. ' s d.g c. HOLDDOWN - STU D. l l -
; /
; , s c . : 4 ! ?
///< //
mum .
,.:,e 'j.
; c:;.
I
.I'A -
FIGURE 3.8-9 Typical Section Through Reactor Vessel Ledge Support O Amend. 61 3.8-56 Sept. 1981 c-, e-- - > e -
~r *+
v 'e-+ e- -e um+1 7 *w---+-e -~e v - e -a** * - > w- --=e- 6-w*e9--v"+ = *
- e'- --*
LlST OF FlGURES Figure Page Number Title Number 3.8A-1 Buckiing-Stress CoeffIclent, Cc, for Unstiffened 3.8A-13 Unpressurized Circular Cylinders Subjected to Axial Compression 3.8A-2 increase in Axlal-Compressive Buck!Ing-Stress 3.8A-14 Coefficient of Cylinders Due to Internal Pressure 3.8A-3 BuckIIng Coefficients for Circular CyIinders Subjected 3.8A-15 . to External Pressure 61l 3.8A-4 b for BuckIing-Stress Unpressurized Circular CoefCylinders f telents, C ,Unstiffened 3.8A-16 Subjected to Torsion 3.SA-5 increase in Torsional Buckling-Stress CoeffIclent of 3.8A-17 Cylinders Due to internal Pressure 3.8 A-6 BuckIIng-Stress Unpressurized Circular CoefCylinders b for fIclent, C ,Unstiffened 3.8A-18 Subjected to Bending 3.8A-7 increase in Bending Buckl ing-Stress Coef ficient of 3.8A-19 O)
% Cylinders Due to Internal Pressure 3.8A-8 Buckl ing-Stress Coef ficient, K for Unpressurized 3.8A-20 Curved Panels Subjected to Axial Compression 3.8A-9 Buckl ing-Stress Coef ficient, K s 3.8A-21 Curved Panels Subjected to Shear, for Unpressurized 61 l 3.8A-10 Buck! Ing-Stress Coef fIclent, K s 3.8A-22 Curved Panels Subjected to Shear, for Unpressurized 3.8A-11 Reduction of Design Stresses Required to Allow for 3.8A-23 47 Blaxlal Stresses of Opposite Sign U
Amend. 61 3.8A-lii Sept. 1981 _ - _ . _ _ _ _ . _ . _ . , , _ . . . . _ . . . ~ , . . _ _ , _ . _ . _ ... . . . _ _
i For Design Load Combinations with SSE, the criteria is the same with the O exception that all of the above allowables are to be increased by f actor of 1.2. NOTE: The stresses to be used in the above Interaction equation are alI to be taken for the same point of the shell under evaluation. Points , evaluated wilI extend to a distance of not less than 3(Rt)l/2 from the 4 61 operating floor. 47 i i } o i i i O h 4 i LO
< 3.8A-11 Amend. 61 Sept. 1981
*-r*e eevewe-wn *= w e -ime* & we +&g area et M w *srw + p*M-39 www w g-eee upeWW t T49"- F9'4W MNM' E N 'W-+- Pp-tW_, p edm N WP9'
TABLE 3.8A-1 Buckling Factors of Safety to be Used in Conjunction with the Critical Buckling Stress Data Provided in this Appendix
' Loading Condition
- Buckling Factor of Safety Construction 1.25 (Local Buckling)
Testing Not Applicable Normal In Accordance with ASME-III Code Design (Including SSE) 1.67 (General Membrane) 1.10 (Local Buckling) Design (others) In Accordance with ASME-III Code
- See Table 3.8-1 " Loading Com.binations" Page 3.8-41.
47 l l l I 3.8A-12 ) Amend. 47 l Nov.1978 l l
p 3.10 SEISMIC DESIGN OF CATEGORY l INSTRUMENTATION AND ELFCTRICAL EOUIPMENT b 3.10.1 Seismic Design critula The Category I instrumentation and electrical equipment will be designed against failure to perform their intended functions during and after an earthquake of the Intensity of the Safe Shutdown Earthquake (SSE) during normal and accident conditions. The seismic qualification progran for Class 61 IE equipment is described in Reference 13 of PSAR Section 1.6, "CRBRP Requirements for the Qualification of Class 1E Equipment." The structural requirements for foundations and supports wilI be in accordance with Section 3.8. In tfdition, the Plant Protective System (PPS) actuation system and its cont. Iled devices will be designed to have the capability to initiate a protective action during she SSE. In addition to the general cr*teria as stated above, the standby power supply system and Category I instrumentation and electrical equipment necessary for safe shutdown will be designed to withstand seismic disturbances of the intensity of the SSE during post accident operation. Category I instrumentation and electrical equipment and components will be designed to ensure the functional Integrity of the equipment under the specified operating conditions. This equipment will require a detailed 53l Investigation to demonstratr its ability to withstand seismic forces while perf orming its intended f unc !on without leading to fuel damage or unacceptable release of radt tion. IEEE Standard 344-1975, "lEEE Recommended Practices for Seismic Qualification 53 f Class IE Equipment for Nuclear Power Generating Stations", supplemented by I Regulatory Guide 1.100, Rev. I will be used as the basis for all seismic 46 qualification of Class IE equipment. The qualification tests will be based on either single frequency sine beat testing at resonance or multiple frequenc.y testing. Single frequency sine beat testing at resonance constitutes severe testing under the most unf avorable conditions where a measured equipment natural frequency has been conservatively assumed to coincide with a building 46l rdetermination supporting system natural frequency. of building natural Thus, uncertainties in the analytical frequencies wilI be conservatively accounted and the maximura peak response acceleration on the appropriate response ' spectrum is conservatively assumed to occur at the equipment's natural frequency. However, when single frequency sine beat testing is used for qualification, addit'onal testing with multiple frequency motion is performed I to comply with the general requirements of IEEE Std. 344-1975, unless they are 46 f ully satisfied by the sine beat testing alone. (See Appendix 3.7-A). The factor of 1.5 in Section 6.6.2.1 of IEEE-344-1975 wilI not be used in either the sine beat or the multi-frequency testing. The peak sine beat acceleration and the maximum acceleration of the multi-frequency motion input to the shake table will be at least equal to the ZPA (Zero Period 25 Acceleration) on the appropriate response spectrum. J 3.10-1 Amend. 61 Sep;. 1981
The static coefficient analysis in IEEE-344-1975 Section 5.3 would be used for simple systems for which it has been demonstrated that this simplified analysis provides adequate conservatism. This h type of analysis, which uses the maximum peak response acceleration on the response spectrum regardless of frequency, plus an additional factor of 1.5, is generally very conservative. However, IEEE 344-1975 is used as the criteria for testing Lnd not'for analysis. Sine sweep testing is not used for equipment seismic qualifi-cation. This type of testing would be used to locate the equipment's natura? frequencies for reasonant sine beat qualification testing n discussed above. 25 The means by which a manufacturer can qualify the equipment include analysis or testing the equipment under simulated seismic conditions or qualification by combined test and analysis. The sei provided to the manufacturer with appropriate responsesmic specification
-spectrum curves at will be the floor elevations and instructions on their use.
The general approach employed in the dynamic analysis of Category I equipment and component design will be based on the response spectrum tech-nique where applicable. At each level of the structure where Class IE equipment are located, horizontal spectra for each of the two major areas of the structure and a vertical response spectrum will be developed. For certain Categorv I complex equipment for which analytic model-ing is not feasible, the equipment supplier will be required to perform dynamic testing using methods described in IEEE Standard 344-1975. The above procedures are applicable to the analysis of seismic design adequacy of Class IE equipment, including supports such as cable tray supports, batteries and rac" , instruments and racks, control con-soles, medium voltage switchgears, unit sub-stations, motor control cen-ters and motors. Suppliers of such equipment will be required to submit ttst data and/or calculations to substantiate that their components and systems will not suffer lors of function before, during, or after seismic loading due to the SSE. All cable tray supports are designed by the response spectrum method. Analysis and seismic restraint measures for tray supports are based on combined limiting values for static load, space length and com-puted siesmic response. 1 O 3.10-2 Amend. 25 l Aug.1976 l 1 I
The following basis will be used in the seismic analysis of typical m) cable tray supports:
- a. All Class IE cable tray supports will be designed to meet the requirement by dynamic analysis using the appropriate seismic response spectra.
- b. The support system will be designed to exclude all natural frequancies in a band cevering the peak or peaks of response-spectrum curve,
- c. Maximum stress will be limited to 90 percent of minimum yield to compensate for effects of higher modes and minor inaccuracies in method of analysis.
The design of typical instrument racks and supports for the instrument tubing is based on the attainment of a fundamental natural frequency of more than 20 Hz so that the floor seismic input will be transmitted through the support without amplificatien. The equipment will be analyzed as an assembly that simulates the intended service mounting, thereby accounting for p.'.a;ble amplification of the seismic input by the equipment support. Seismic documentation submitted by the vendor will be reviewed by the design engineering group to ensure that the support system has been considered. Where it is necessary to test individual devices (e.g., relays or instruments) separate from the panel on which they are mounted, the acceleration of the panel at the device locations will be checked to ensure a level less than that at which the devices are qualified. The structures that will be seismically qualified as Seismic Category I are listed in Section 3.2, Table 3.2-1. The Class IE equipment is listed 'j 59 in Table 3.2-3. 3.10.2 Analysis, Testing Procedures and Restraint Measures The seismic qualification of safetv r' ted instrumentation will be 59 performed and documenW as specWed in Reiem:e 13 of PSM Sec&n IJ. h Category I electric equipment such as battery racks, instrument racks, and control consoles located in Category I structures will be supported and restrained to resist uplift or overturning resulting from seismic forces. 1 Amend. 59
\ 3.10-3 Dec. 1980
O t TM LE 3.10-1 HAS BEEN DELETED 61, t i O l 1 l O 3.10-4 Amend. 61 Sept. 1981
3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EOUIPMENT 3.11.1 Eautoment 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 radiation 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 Eautoment Environmental OualIfIcatIon orocram aualIfles orototvole eautoment to hiah concentrations of sodium oxide aerosols. Equipment will be qualified for environments of low conentrations of sodlum oxides as reautred throuah a generic test oroaram. Where possible, the equipment comprising the safety related 180 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 k 3.11.2 OualIfIcatIon Tests and Analvses The program of environmental qualification tests and analysis for Class 1E Equipment is described in Reference (13) of PSAR Section 1.6, "CRBRP Requirements for Environmental Qualification of Class 1E Equipment." This document establishes the qualification program which will be conducted to qualify Class 1E 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 61 Reference 13 of PSAR Secilon 1.6 and summarized, as appropriate, in the FSAR. l l O V 3.11-1 Amend. 61 Sept. 1981
3.11.4 Loss of Ventilation All plant locations containing safety related control and electrical equipment, that need a controlled environment to maintain the l required operability, are to be provided with redundant air conditioning and/or ventilation facilities for the needed environmental control. 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 Enclosures containing safety related equipment will be designed to withstand the effects 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 compartments or rooms or by enclosing the safety related equipment l in nonabsorptive, noncombustable, explosion proof casings. Ap7licable l code requirements including those of the National Electric Code will be satisfied, as appropriate. t O l l l l l 1 3.11-2 0
3A.2.1.3 Deslan Evaluation i V' The provision of a head access area to which operating personnel wilI have ready access enhances both the safety and availability of the plant by permitting surveillance and inspection of the many systems and components 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 from any of the seals in the head. Detectors will be located in the head access area. The equivalent of the volume of the HAA will be recirculated in about two minutes. The rapid circulation and ventilation rate 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 buffer gas between the seals, wilI tend to minimize leakage and to dilute the cover gas prior to any leakage into the head-access area. 3A.2.1.4 Testina 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 The design basis for the removal of heat from the Head Access Area is to maintain the temperature below a level which would be deleterious to tq Instrumentation, equipment, seals and personnel working in the area. Heat will v be removed from the Head Access Area by providing a Unit Cooler. Refer to 61 Section 9.6.2.2.1 fcr detailed description of HAA area cooling system. Consequences of loss of the heat removal system have been examined. The loss would result in a natural convective air flow that would transfer heat to the reactor containment building atmosphere. In the worst case (loss of off-site 4 power and power to the head access area cooling system) the R W atmosphere is expected to rise to a maximum of 1200F from its normal. temperature of 80 to 850F. This would result in a temperature rise of the components within the head access area which will not exceed 400F.
- The inflatable seals have been designed for an operating temperature below
- 1250F under normal conditions. The controlling factor is the diffusion through
! the seals. If required, the increased diffusion accompanying the maximum temperature rise of 400F wilI be accommodated by purging the region between the inflatable seals and the sodium dip seal. This will be ascertained by the on-going seal design studies. 1 l Other equipment in the head access area would remain functional at an RC8 atmospheric temperature of 1200F. 2.5 l l b) 3A.2-3 Amend. 61 Sept. 1981
n 3A.4 REACTOR SERVICE BUILDING (RSB) ( ) V 3A.4.1 'DesIan Bases The Reactor Service Building (RSB) design is bas =,d on:
- 1) Providing housing and structural support for portions of the Auxiliary, Liquid Metal Reactor Refueling, and Maintenance Systems. Tabl e 3A.4-1 61 lists the functional systems located in the RSB. Table 3A.4-2 lists i the supporting systems located in the RSB.
391
- 2) Providing protection again.st seismic and tornado events, and some degree of confinement for several systems that contain radioactive materials and spent reactor fuel.
29 3) Serving as an Intermediate transfer and storage facility for new and spent fuel, other components, equipment, and materials entering and leaving the Reactor Containment Building (RCB).
- 4) Providing radiation protection, primarily in the form of concrete shielding, to meet the radiation shielding-zoning criteria presented in Tabl e 12.1-1.
- 5) Allowing installation, removal, repair, maintenance, and re-installation of equiptrent and components housed therein.
29
- 6) Providing safe entrance and egress for operating and maintenance v personnel who perform work in the RSB.
- 7) Providing environmental control (heating and ventilating, Inerted cell, etc.) for various functional systems.
- 8) Providing sealing protection against ground water and flood water conditions.
- 9) Providing security to safeguard fuel per 10CFR73.
, 3A.4.2 Design Descriotion l l 3 A. 4 .2.1 Reactor Service Building l l Following is a brief description of the RSB and certain systems located therein. Tabl e 3A.4-1 lists all the systems the RSB and the appropriate PSAR references for detailed information of these systems. Additional Information l Is provided in Section 3.8.4.1.1. General Arrangement Drawings are provided in Section 1.2. l L l O Amend. 61 3A.4-1 Sept. 1981
3c The RSB is a reinforced concrete, and with the exception of the radwaste area 61 is designed as a seismic Category 1, tornado hardened structure above and below grado level at El. 815 ft. The RSB structure, in addition, houses an airlock, cranes, a freight elevator, and a railroad track that allows railroad cars to 39 be brought directly into the building. The airlock,13 '4" In Internal diameter, is provided between the RSB and the 39, 61 RW. The elevation of the airlock is adjusted to provide a 8-f t. wide by 8-f t. high passage between the cperating floor of the RSB (El. 816 ft.) and the RC8. The minimum clear space between the doors of the airlock is 20 ft. A 125 ton capacity bridge crane is provided in the RSB. The hook on the crane 61 will have a learance of 42 ft. above the operating floor. l 39 A hardened railroad door 18'0" x 22'0" is provided between the RSB and the Radwaste Building (RWB). This door is designed to withstand tornado generated l 61 missiles. Personnel access and egress is provided in the RSB structure from all levels, via four staircases and one elevator. The elevator and one of the staircases 61 are located in the southwest corner, the other three staircases are located in ) the northwest, southwest and northeast corners, respectively, as shown in the general arrangement drawings in Section 1.2. Corridors are extensively 61 provided throughout the building for rapid egress. l Leakage of radioactive gas from the various systems within the RSB wilI be i restricted by utilizing commercially avaliable seals To limit, under normal , cperating conditions, the dose rate within the RSB due to radioactive gas j leakage below 10% of the zoning cr!teria, as established in Table 12.1-1. Additionally the RSB Internal pressure is maintained at negative 1/4" W.G. to restrict the release of radioactive contcminarts to the atmosphere. l The foundation for the west end of the Radwaste Area is at grade elevation and is founded on compacted structural backfill. The Redwaste Area structure is
' designed to meet the requirements of the Standard Building Code. In addition I the structure below grade as welI as the Solid Radwaste Area above grade are 61 designed as reinforced concrete structure.
The upper part of Radwaste Area, the steel framed structure is designed to 61 ensure that the cdjacent seismic Category I structure of Reactor Service Area 29l is not damaged nor its safety functions compromised during an SSE. l The RSB has been designed as Seismic Category I co,sistent with its safety l function. The sections in this report dealing with the various systems located i 39 in the RSB (see Table 3A.4-1) present their individual seismic category l requirements. 29l I 3d O 3 A.4-2 Amend. 61 Sept. 1981
Auxillarv Coolant Systems p 3 A.4 .2.2 The decay heat generated in the EVST sodium is removed by a Na-Nak heat exchanger. The Nak, in turn, releases its heat through an air blest heat exchanger. These heat exchangers, Na-Nak are located at elevation 765'0" of 61 the RSB. A complete, redundant system of heat exchangers is present, as a 39 standby in the event of a failure. The redundant systems and their accompanying lines are routed Independentally of each other by physical barrier separation. The FHC is cooled by the Recirculating Gas Cooling System, which, in turn, gives up its heat in gas-Dowtherm J heat exchangers. The Dowtherm J gives up this heat to a Chilled Water System Dowtherm to water heat exchanger. 61 Section 9.7 provides more details on this system. 3A.4.2.3 Deleted 39l 29 3A.4.2.4 Heating and Ventilation 61l ventilating The Heating theand Ventilation plant SystemSection atmosphere. provides9.6for air-conditioning and 12.2 provides and more details on this system._ 3A.4.2.5 Sodlum Fire Protection The Sodium Fire Protection System provides the means of detecting, alarming, 38 containing, and controlling sodium and/or NaK fires. Details of the SFPS are described in Section 9.13. 3A.4.2.6 Recirculatina Gas Cooling The Recirculating Gas Cooling System provides cooling of the Inerted cells and is described in detall in Section 9.16. 47 3A.4-3 Amend. 61 Sept. 1981
47 3A.4.2.7 Reactor Refueling The Reactor Refuel ?ng System performs all handling operations on coro 44l assemblies destined f or the reactor, frun the now fuel shipping and recolving to the installatloa of the spent f uel shipping cask into the rLilroad flat car. Its functions are performed in the RSB and RCB. Section 9.1 providos details of this system. The three major areas occupied by this system within the RSB Reactor Servico 61 area are as follows:
- 1) Ex-Vessel Storage Tank - The EVST, located below the operating floor, will recolve, hold and cool all coro essemblies discharged from the
' tor vessel prior to shipment offsite.
- 2) Fuel Handling Cell - The FHC is a subfloor hot cell which can hold up to three coro assembiles for inspections, measurements and transfer to spent the fuel shipping casks for offsito shipment.
- 3) Two New Fuel Unloading Stations - Each station wilI contain one new 20 fuel assembly in a new fuel shipping container.
3 A.4.2.8 Nuclear Island General Puroose Maintenance Eculoment The Nuclear Island General Purpose Maintenance Equipment System provides the capability for main <enance of the Nuclear Steam Supply System (NSSS). These maintenance operations ./o accomplished within the Reactor Containment
; Building, The Reactor Servico Building, and the Steam Generator Building. The i system provides general purposo equipment and procedures for removal and replacement of radioactive and/or sodium serv!co components of the Fuel Handling System, Heat Transport System, Auxillary Systems, and Reactor Systems.
Capability is also provided for sodium removal and decontamination. The system also providos the general purposo equipment used for the removal, repair, maintenance, and reinstallation of equipment and components housed within the 59 nuclear Island. O 3 A.4-4 Amend. 61 Sept. 1981
3A.4.2.9 Auxilfarv Llauld Metal The Auxiliary Liquid Metal System provides the facilities for purification and cooling of the sodium in the ex-vessel storage tank (EYST). The EVST sodium storage is provided by the Primary Sodium Storage and Processing System of the Auxillary Liquid Metal System. This is discussed in greater detail in Sec+ Ion 46l 9.1, 9.3. The maximure activity in the EVST sodlum (i.e., af ter 30 years of plant operation and with no EVST cold trapping) is given in Table 12.1-23. 3 A. 4 .2.10 Inert Gas Receivino and Processing Sections 9.5 and 11.3 present details of this system. The radioactive inventory, by isotope, present in the various cells within the RSB are given in Table 12.1-12 through 12.1-18. 3A.4.2.11 Imourftv Monitoring and Analysis The impurity Monitoring and Analysis System provides for the sampling monitoring, and analysis of sodium and cover gas impurities in the CRBRP systems. The system provides the following areas of Impurity monitoring and 46l sampi ing:
- 1) EVS cover gas sampiIng
- 2) Primary cover gas sampiIng 46l 3) EVS sodium sampiIng Section 9.8 presents more details of this system.
3A.4.2.12 Fuel Failure Monitoring Section 7.5.4 presents detalis of this system. The isotopl. gas activity In the sampling trap cell (gas tag analysis) and the cover gas monitor cell are presented in Table 12.1-20. 3A.4.3 Design Evaluation The RSB is designed to house the various systems listed in Table 3A.4-1. Each of the systems containing radioactive fluids or components will be housed in 61 separate cells, that are provided with walls of adequate thickness. These walls, in addition to providing radiation protection to operating personnel, will act as a confinement barrier. Accidents considered by the individual systems housed in the RSB are presented in Chapter 15. O 3A.4-5 Amend. 61 Sept. 1981
3A.4.4 Tests and Inspection A CRBRP Quality Assurance Program is established ti assure that critical structures are built in accordance with specifications. This pro-gram is described in Chapter 17. Principal Materials Used in the RSB - Concrete, reinforcing steel, steel liner plates, and structural steel - are manufactured in accordance with nationally recognized standards. User installation tests and inspections are detailed in construction specifications. Converi.ional methods will be used to inspect the cell liners. These methods may include:
- 1) Visual inspection of welds
- 2) Dye penetrant
- 3) Vacuum box Tests and inspection will be performed during construction of the RSB structure, to verify conformance with construction specifications and appli-cable parts of building codes.
The tests and inspection of systems within the RSB are discussed in detail in those sections of this report pertaining to the individual systems housed by the RSB (see Table 3A.4-1). 3A.4.5 Instrumentation Requirements The RSB will be sufficiently instrumented to pi tide for the safety of both operating personnel and the general public. This instrumentation will include such items as neutron counters for EVST and the FHC area, radia-tion detectors in all accessible areas, exhaust monitors for the H&V System, etc. The specific instrumentation requirements for the various systems in the RSB will be the joint responsibility of the functional :l/ stem and its corre-sponding instrumentation system. These pairs of systems, together with a brief discussion of their instrumentation requirements, are given in other sections of this PSAR (see Table 3A.4-1). The responsibility of providing general radiation monitoring (i.e. , not within the jurisdiction of any func-tional system) will be the Radiation Monitoring System. Section 12.2.4 of this PSAR presents the requirement for radiation monitoring in the RSB. O 3A.4-6
! O O O l
! TABLE 3A.4-1 1 MAJOR SYSTEMS LOCATED IN RSB l 29
- - System Name PSAR Reference i
1 Chilled Water (Normal & Emergency) 9.7 j 15
.39 3
4 Heating and Ventilating 12.2 {
- Sodium Fire Protection 9.13.2 j 47l Recirculating Gas Cooling 9.16 j Reactor Refueling 9.1, 7.6 i
! 5 Maintenance (NSSS) 9.2 l b ! l 0 Auxiliary Liquid Metal 9.3 i
- Inert Gas Receiving and Processing 9.5, 11.3
- i Impurity Monitoring and Analysis 9.8 ;
t Fuel Failure Monitoring 7.5.4
. Radiation Monitoring 12.2.4 ;
^ Non Sodium Fire Protection 9.13.1 g g' 29 f i
.a e-l '
t se 4
TABLE 3A.4-2 SUPPORTING SYSTEMS LOCATED IN THE RSB System Name Service and Function Building Electrical Power Electrical power for outlets, equipment, and emergency power Communication Communication within the RSB and to outsido areas Lighting System Provido space and special lighting Reactor Containment Provido equipment air lock and maintenance cask door that opens into the RSB Plani Annunciator System Provide annunciators within appro-priate areas of the RSB Piping and Equipment Provide trace heating on selected Electric Heating and Control piping in the RSB Radiatloa Monitoring Provido radiation monitoring throughout RSB Plant Protection Provido plant protection systems and devices to prevent unauthorized entry into the RSB and fuel storage vault Containment Cleanup System Provide cleanup operations for the 61 RC8 Annulus Atmosphero during a TEDB event. O Amend. 61 3A.4-8 Sept. 1981
3A.5 STEAM GENERATOR BUILDING 3 A.5.1 Design Basis The Steam Generator Building (SGB) design is based on:
- 1) Providing housing and structural support for portions of the Intermediate Heat Transport System, Steam Generation and Steam 61 Generator Auxiliary Heat Removal System and Maintenance Systems. Table 3A.5-1 lists the functional systems located in the SGB. Tabl e 3A.5-2 lists the supporting systems located in the SGB.
- 2) Providing protection against seismic and tornado events.
- 3) Aliowing removal, repair, maintenance and re-installation of equipment and components housed therein.
- 4) Providing environmental control (heating and ventilating) for the various functional systems.
- 5) Providing sealing protection against ground water and flood water conditions.
- 6) Providing safe entrance and egress for operating and maintenance personnel who perform work in the SGB.
p 3A.5.2 Design Descriotion Following is a brief description of the Steam Generator Building. Table 3A.5-1 lists all the systems inside the SGB and the appropriate PSAR references for detailed information of these systems. General Arrangement Drawings are provided in Section 1.2. 45
* '"" "' "9 "'* 9*Y " * *"
61 reinforced coner ete structure with an attached Seismic Category I unhardened structure for equipment maintenance and repair. The building is functionally subdivided into four bays which house the major equir ent noted below: Intermediate Bay i Ex-Containment, 61 Primary Sodium Storage Vessel Intermediate Sodium Piping Cable Distribution 15 Chilled Water Systems V]
/
i l i 3A.5-1 Amend. 61 Sept. 1981 l
, _ . _ _ __ ___ -_- . _ , _ , _ . _ . - -- ~___ ,
i l Steam bunarator Bay Intermediate Sodium Dump Tanks l Evaporators Superheaters Intermediate Sodium Pump Intermediate Cold Traps intermediate Sodium Expansion Tanks 61 Reactor Pr oducts Separation Tanks Auxillary Bay Instrumentation and Controi Panels Auxillary Heat Removal System Steam Generator Recirculation Pump Steam Drum Mainte..ance Bay 61 Intermediate Sodium Rrsnoval System and Cleaning Cell Maintenance Platfc. ins for Steam Generator, Superheater and lHTS Puy Railroad Siding Tiie intermodlate Bay, Steam Generator Bay and the Auxiliary Bay are tornado-hardened Seismic Category I structures and will be reinforced concrete
- enclosures with reinforced concrete slabs on structural steel framing. The 61 Maintenance Bay housing noncritical components and systems will not be a
, tornado-hardened structure but will have a Seismic Category I steel framework. The metal wall siding and metal roof docking will be designed in accordance l with standard practice for industrial buildings. Interior structures will l consist of structural framing, floor grating or concrete decks. Structural i steel wilI recolvo a firo protection cover in accordance with the requirements l 61 45 of the Standard Building Code. l Postulated accidents as discussed in Chapter 15 do not impose a leak rate capability on the SGB (except the primary sodium storage cell in the Intermediate Bay) or require more than a 3 psi differential pressure copabilIty as stipulated by the tornado considerations. A brief description of the structural design features of the building is given in Section 3.8.4. A detailed description of the Heating, Ventilation, Cooling 61 nd Air-Conditioning System serving the building is given in Section 9.6. Design characteristics of the SGB to mitigate the offects of seismic and 61 tornado events and steam line ruptures, resulting loads and the building structural design criteria are described in Chapter 3. General Arrangement Drawings are provided in Section 1.2. 1 3A.5-2 Amend. 61 Sept. 1981
3A.5.2.1 21 eat Transoort and Steam Generator Svstems. The SGB houses the three intermediate heat transport loops. Each loop i transports heat from the respective IHX to the steam generator system and returns the cooled sodium back to the IHX. All piping and components are
- Seismically supported and housed in Seismic Category I, tornado hardened 61 structures. The steam generator system consists of 2 evaporato.s and 1 ,
j ' superheater per loop with their associated steam drum and recircu!ation system. t Section 5.5 provides more details of this system. ' 3A.5.2.2 Steam Generator Auxiliarv Heat Removal System SGAHRS The SGAHRS is a safety related system which provides safe shutdown of the reactor for both short and long periods of time should the normal steam dump to the condenser be Inoperative. This system is housed entirely within the 61 Auxillary Bay. Section 5.6 provides more detailed description of this system. 3A.5.2.3 Heatina and Ventilatino i The Heating, Ventilating and Air Conditioning System provides for air-conditioning of the atmc.:phere within the SGB. Sections 9.6 and 12.2 presents 61 m re detalis of this system. 3A.5.2.4 Fire Protection The Sodium Fire Protection System provides the means of detecting, containing, alarming and controlling sodium fires. Details of the SFPS are described in A 35 Section 9.13. 3A.S.2.5 Maintenance A maintenance bay is provided adjacent to the steam generator cell. This structure provides housing for maintenance stands for both an Intermediate ! 61 sodium pump and an evaporator /superheater module. Additional description is provided in Section 9.2. 3A.5.3 Deslan Evaluation Sodium fires within the Steam Generator fullding represent an additional loading in the building as discussed in Section 3.8.4. The sodium fire 61 accident sequences and resulting loadings on the Steam Generator Building are provided in Section 15.6.
- O 3A.5-3 Amend. 61 Sept. 1981
Each of three Intermediate sodium loops and its associated sieam generator system are physically separated in independent cells to prevent an accident in one Ioop from causing falIures in any of the other ioops. The extent of the 61 Isolation is described below. o Separate primary sodium loops contained within separate cells in the Rm, o Separate intermediate sodium loops contained within separate cells in the Intermediate bay. o Separate steam generator systems and their associated portions of SGAHRS contained within separate cells in the steam generator building. Reinforced concrete walls provide loop separation in all buildings and steel catch pans will be provided to retain sodium at locations of potential spillage. Each cell containing a portion of a loop is independent *y controlled to permit loop-to-loop Isolation. Piping systems do not intercontect in the intermediate system loops, except for small vent and d ain lines which can be closed of f by valves to permit loop-to-loop Isolation. Drains from coils contelning portions of a loop will be Isolated by suitable valving. Electrical, instrument and control cabling will be routed separately to each loop so that cabling for one loop does not pass throup,h a cell containing another loop. The same applies to piping and ventila*lon ducting. Where unavoidable, piping and duct penetrations between celis containing loops, wilI be provided with protective missile sleeves or simila protective measures. Separation of the loops in the auxillary bay is provided to prevent propagation of an accident into an adjacent cell. Reinforced corcrete walls and barrier doors provide thIs proteetIon. The arrangement and crientation of components (e.g., piping, valves, rotating machinery) is such that the protection of vital equipment in adjacent cells from equipment generated missiles is provided. In cases where piping systems do interconnect the steam generating system loops, (i.e. main steam lines, normal and auxiliary feed water lines) automatic isolation is provided to prevent propagation of a foult f rom one loop system to another. The auxillary feed water system is designed so that no single active failure following the Initiating event can prevent auxiliary feed water from reaching the operating steam generator loops. This separation is necessary to ensure the operability of the SGAHRS System after an accident occurs in one loop. O Amend. 61 Sept. 1981
l TA.6 DIESEL E NERATOR BLllLDING 3A.6, Desfgn Bases The Diesel Generator Building is immediately adjacent to the Control Building. It contains safety related emergency electrical power su'sply equipment to 15 { supply emergency shutdown power in the event of loss of all of fsite AC power. The Diesel Generator Building design is based upon: 61 l 1) Providing housing and support for portions of the functional systems listed in Table 3A.6-1. Table 3A.6-2 lists the supporting systems in the DGB.
- 2) Providing protection apsinst seismic and tornado events for those systems in the DGB.
- 3) Allowing removal, repair, maintenance and re-Installation of equipment and compor.ents housed therein.
- 4) Providing safe entrance and egress for operating and maintenance personnel, who perf orm work in the DGB.
- 5) Providing environmental control (heating and ventilating) for various functional systems.
fi V
- 6) Providing sealing protection against ground water and flood water conditions.
3A.6.2 System Design Dcscriotion The Diesel Generator Building is a tornado hardened, seismic Category ?
. structure. The two diesel generators are supported at grade elevaMon, and are separated from each other by a wall designed to withstand the impact of any 61 missile generated by a diesel-generator casuaity. tsui Ming materials and adequate walI thicknesses provide the required fire barriers; doorways are provided with the required fire rated doors. Openings in exterior walls and roof are missile protected with barrier walls or a Penthouse. The building is also protected against floods. (See Section 3.4)
The Diesel Generator Building HVAC system provides the required ventilation to 61 maintain adequate temperatures under normal and emergency conditions for the various areas in the building. The description of the Diesel Generator Bullding HVAC System is presented in Section 9.6. 1 An equipment removal hatch is provided in the building for the purposes of installation, maintenance, and removal of equipment located on the elevations Delow grade. i l 3A.6-1 Amend. 61 Sept. 1981 l
Complete separation is maintained between the redundant diet,el generators with their respective auxiliary support systems to ensure that a malfunction or f ailure of an active or passive component will not Impair the capability of at least one diesel-generator to supply power for a safe plant shutdown. The diesel-generator auxiliary support systems, which include fuel oil, lube oil, starting eir, and cooling water systems, are classified moderate energy systems and, as such, are subject to through-the-wa8I pipe leakage cracks with possible subsequent flooding. Each diesel generator cell is designed to prevent 61 propagation of flooding to the other diesel generator cell. A set of M. V. switchgear buses, unit substations, and motor control centers are located in the Diesel Generator Building. These electrical components are 53 separated by a wall to prevent a casualty to one cell from propagating to the 61 other redundant coll. 15l The Diesel Generator Building is protected by the seismically supported Non-61 Sodium Fire Protection System. The appropriate fire extinguishing systems wilI be provided throughout the building to minimize the adverse ef fects of fires to safety related systems. 3A.6.3 DesIon Evaluation The Diesel Generator Building has the following fcatures that collectively provide the capability needed to satisfy CRBRP General Design Criteria 2, 3, and 4 as described in Section 3.1.
- 1) The building is designed to withstand natural phenomena such as earthquakes, tornadoes, tornado-generated missiles, and floods as described in Sections 3.8, 3.3 and 3.4, respectively.
- 2) The building and the systems contained are protected against fire using detection equipment and the appropriate fire extinguishing devices as described in Section 9.13.1.
- 3) All systems inside the Diesel Generator Building, important to safety, are reiundant and are separated and protected so that the failure of one system will not cause the f ailure of the redundant system.
O 3A.6-2 Amend. 61 Sept. 1981
e 3A.6.4 Testing and insoection Pre-operational and periodic teus and inspections will be performed for the HVAC systems and Fire Protection Systems in the Diesel Generator Building to assure the adequate and reliable performance of these systems as described in Sections 9.6.5 and 9.13.1. Testing and Inspection of the systems contained in the building will be - performed as described in Sections 8.3, 9.7.4, 9.9.1.4, and 9.9.2.4. A CR8RP Quality Assurance Program is established to assure that critical structures are built in accordance with specifications. This program is described in Chapter 17. Principal materials used in the DGB - concrete, reinforcing steel and 61 structural steel - are manufactured in accordance with nationally recognized 3 standards. User installation tests and inspections are detailed in construction specifications. Tests and inspect!on wilI be performed during construction of the DGB structure, to verify conformance with construction specifications and I applIcabie parts of tuiiding codes. 61l The tests and Inspection of systems within the DGB are discussed in detail in those sections of this report pertaining to the individual systems housed by the DGB (See Table 3A.6-1). O l i O
/ mend. 61 3 A.6-3 Sept. 1981- ~_ _- _ - , ,. ~ - - . _ . . _ . - - . _ - . _ _ . _ - - _ . - - . _ _ _ - _ , _ _ - - _ - _ _ _ . _ - . . , _ . _ , _ . _ -
i f i
! TABLE 3A.6-1
{ SYSTEMS LOCATED IN THE DGB System PSAR Reference
! Building Electrical Power Systems 8.0 .
Plant Protection System 7.0 . 15
- j7l Service Water System' 9.0 1'
61 l Heating and VentiIating 9.6 i l i 1 i } } i L I , i I l l 3A.6-4 Amend. 61 Sept. 1981
TABLE 3A.6-2 SUPPORTIf1G SYSTEMS LOCATED If1 THE DGB Communications Lighting Plant Fire Protection Plant Protection Radiation Monitoring 9 l l l 9 3A.6-5 l
,m Flow patterns in the region immediately above the core have been investigated I l in water table tests. These tests have shown that a torroidal flow pattern V
exists in the mixing chamber located directly above the core. A large portion of the stream to stream temperature differences are reduced in this chamber before the flow exits. Tmperatures in these flow streams dif fer substantially, hence the mixing adjacent to the Inner surface of the mixing chamber results in thermal striping. The material selected for the exposed 41 surf aces in the mixing chamber must therefore have an endurance stress limit 59 in excess of the maximum anticipated stress amplitude produced by fluid 58l mixing. This requirement led to the selection of Alloy 718 for tne exposed 41 surfaces of mixing chamber components. 4.2.2.2.1.8 Core Restraint System Design of the CRBRP core restraint system is based upon the Iimited free bow concept. Essential features of this concept are illustrated in Figure 4.2-47. Fixed peripheral formers provide lateral support to the core assemblies at two locations above the active core. A third support at the core support plate elevation completes the lateral support configuration. Relief of restraint loads for refueling in the limited free bow core restraint concept is achieved by allowing the core assemblies limited freedom for unrestrained bowing during the core startup and shutdown transients. The amount of free bow permitted is controlled by sizing the gaps between core assembly load pads, and between the peripheral load pads and the adjacent core O formers. The upper bound of the allowable core and former gaps is defined by V a conservative analysis of the of feet on critical core components of a step compaction of the core through the range of free motion. permitted by the gap configuration. The resulting core step reactivity insertion is not allowed to produce transient heating rates in the fuel which would result in the fuel pin upset condition damage Iimits being exceeded. it is evident that the core restraint system in its entirety includes all the reactor assembiles plus elments of the core support structure and the upper internals structure. Only the core formers, their associated retention and positioning hardware and the removable radial shield essemblies are categorized as core restraint hardware. I 41 n
, (a) 58 49
.2 W 4 Amend. 61 l
Sept. 1981
4.2.2.2.1.9 Removable Radial Shield The radial shield assemblies are made up of stainless steel rods held within thin walled hexagonal ducts. These assemblies are designed to be as flexible as possible in order not to contribute to the off-power restraint loads. A close-fitting support block is inserted inside the duct at the ACLp to provide axial restraint for the shield rods and to absorb seismic loads that are transmitted through the ACLP to the core former. 4.2.2.2.1.10 Core Fomer Structure The core former structure is composed of three substructures, the upper core former ring, a spacing cylinder, and the lower core former ring. The core former rings are comprised of profile milled segments assembled into continuous rings, as illustrated in Figure 4.2-46. The above core load plane former ring, called the lower core former ring, is mounted on a ledge machined in the inner diameter of the core barrel. The spacing cylinder, called the support ring, provides holddown for the lower core former ring and support for the top load plane fomer ring called the upper core fomer ring. The upper core former ring has six lugs that fit slots in the top of the core barrel to transmit seismic and other loads to the core barrel. A series of L-shaped keys are circumferen-tially slipped into the gro."ve on the inside of the core barrel, between each of the six lugs, and trapped by means of a radially oriented dowell pin on either side of each slot. These L-shaped keys prevent vertical displacement of the core fomer rings. The upper core former ring is centered in the core barrel cavity by means of the six radial lugs. The lower core former ring is centered in the core barrel cavity by means of radial shims. 4.2.2.2.1.11 Maintainability All the reactor internals except for the reactor assemblies, are designed for a 30 year life with no scheduled maintenance. However, provision has been made to permit removal of the lower inlet modules to assure full plant life and malfunction recovery capability. Contributing factors which may require malfunction recovery capability include:
- 1. Potential damage to the reactor assembly receptacles, as a result of in.sertion and removal of reactor assemblies.
- 2. Potential wear or partial plugging of strainers and orifices, as a result of coolant induced changes.
58 4.2.2.2.1.12 Surveillance Material surveillance coupons are contained within special 51 assemblies located in removable radial shield positions and a fuel transfer and storage assembly. In addition to these special assemblies, Amend. 58 O 4.2-175 Nov. 1980 L -
V 59l 4.2.2.4.2.4 Mechanical Loads Design Loads The design condition loads are ghe deaa weight and pressure. The design temperature of the UIS is 1220 F. Nonnal Loads During normal operation, the UIS carries no mechanical load except its own weight and loads due to actuation of the control rod system. The upper shroud tubes carry the dashpot loads resulting from the primary control rod scram arrest accelerations, and the lower shroud tube is designed to react a 19,000 pound upward load incurred in exercising the control rod breakaway joint. The UIS is designed to preclude the occurrence of adverse structural and dynamic effects due to flow induced vibration. Where possible, the entire structure and its components are designed such that their natural frequencies do not coincide with any vortex shedding frequencies. Component mechanical stresses caused by flow induced vibration are required to meet the limits of the ASME Boiler 59l Code Section III and Code Case 1592 (N-47) for normal conditions to ascertain structural integrity with regard to fatigue. During refueling operations, the UIS is raised and lowered k so that the rotating plugs may be positioned to provide access to various reactor locations. Misalignment of the UIS keys with respect to the keyways in the CFS will cause loads between the keys and keyways. Both the normal and frictional force resulting from this misalignment re considered in the analysis. Upset Loads t The upset mechanical loads on the UIS are the seismic input for the operating basis earthquake (0BE). The UIS is designed for OBE in accord with the criteria described in Section 3.7. Emergency Loads l The UIS is designet to accommodate loads due to loss of primary holddown (hydraulic oalance). Loss of hydraulic balance is classified as an emerger.cy event and is assumed to occur five times, but for conservati',m it is analyzed as an upset event. l 59 51 l l Amend. 59 l [vs) 4.2-214 Dec. 1980
-ee+'
=p7 w % -4 s w- - +-- g. &----->e Mt ey r,,,,rq _ ,-,, m ,.9 qy,%ga y _,,y , ,ps-- w- -
p x- ay-- *9W-- ..-,v
The UlS is designed to withstand the effects of a safe shutdown earthquake (SSE). The SSE is a faulted condition, however, to be conservative, the UlS is designed to satisfy the ASK Code criterie for emergency conditions when in the operating configuration and subjocted to the SSE loacs. SS 4.?.?.4.2.5 Thermal Environment Operating conditions for the UlS are specified f r e 30 year histogran using the ASME Codo categories of normal, upset, emergency, and f aulted conditions for the mechanical Ioads and steady stat,e and tranclent temperatures. Plent capacity is 75% giving a full power life of 22.5 years. Normal Loads j The UlS is designed to accommodate thermal striping during normal operation. SS The Uls surf aces directly exposed to the more severe thermal striping are the Instrumentation posts, control rod shroud tubes, keys, tho internal surfaces of the chimneys, and the UlS mixing chamber. Sodium exiting from the chimneys 5d will subject the support columns to thermal striping. The reactor operating temperature, for long term steady state of fects in a J cumulative damage anasysis, is based on reactor coolant outlet temperature of 61l51 10000F with a 2o, uncertainty. Durir.g refueling operations, which are normal operating conditions, the UlS wili be at the refuoling tempe:sture of 4000F. Normal operating temperature trans11nts such as startup and shutdown are less severs than the upset and emergency events and are enveloped by them. Uoset and Emergency Loads Steady state temperatures with 2 uncertainties for a reactor outlet nozzle 51 temperature of 10150F are used to begin transient analysis of UlS components. O Amend. 61 4 .2 -21 5 Sept. 1981
[V arising from radiation induced creep and swelling in the core assembly ducts. Uncertainties in dimensions, environment, material i properties and the model are all considered to arrive at enveloping contact loads. Contact loads are predicted for both normal and non-seismic off-normal events for both on-power and off-power (refueling) conditions. Non-seismic loads obtained from this model are combined with seismic and other loads as indicated in Section 4.2.1.3.2.3.3. These combined loads are used in the structural evaluation of reactor system components which interface with the core restraint system. Loads calculated at refueling conditions are utilized to demonstrate that the refueling limit for assembly insertion and withdrawal is satisfied. System model results predict core assembly bowing and dilation effects as influenced by the environmental conditions (temperature and flux) within the core. Core assembly distortions are predicted to be sufficiently small so that the interassembly contact limits and the assembly handling envelope limits are satisfied. c) Top End Misalignments The evaluation of miselignment of the core assembly handling socket is based on stackup of top load plane gaps from a given assembly to the farthest locatior on the upper core former ring. Thermal and dimenstional uncertainties and permanent component misalignment x effects are also included. This very conservative procedure is j used to set the top load plane gaps to insure that assembly top end misalignment limits are satisfied. d) Duct-To-Duct Cuntact Interassembly contact at non-load plane locations due to the combined effects of bowing and duct dilation is also computed. Present evaluations indicate that local non-load plane contact between assemblies will initiate between fuel assemblies during the second cycle of irradiation. The results show, however, that no general duct-to-duct non-load plane contact pattern is' established. 4.2.2.4.3.2 Material Properties The analyses of this section utilize material properties found in Reference 181 for irradiation swelling and creep. Analyses were performed using nominal fonns and with uncertainties conservatively applied. 4.2.2.4.3.3 Analysis of Assembly Bowing and Duct Dilation Core restraint analyses were performed using the NUB 0W-3D core restgaintsystemanalysiscomputercode(seeAppendixA). The code analyzes i a 30 sector.of the core restraint system including irradiation and mechanical 59 influences as shown in Figure 4.2-85. , p V Amend. 59 4.2-222 Dec. 1980
Assembly duct temperaturo data and neutron flux data used in the core 53l restraint analysis are shown in Section 4.4.3.3.5, " Core Assemblles Duct Temperatures", and Section 4.3.2.9, " Vessel irradiation". Those data were applled over a time period corresponding to two cycles of reactor operation to determino the thermal, Irradiation swolling, and creep bowing response of the modoled assemblies. Examples of assembly bowing profiles and Interassembly loads are presented in Figures 4.2-88 and 4.2-89 for the following conditions in the operating cycle:
- 1. At power, start of cycle one.
- 2. At power, end of second cycle (328 days).
Assembiv Bowino and Interassembly Load Patterns
~
51l Influencing the bowing profiles in the timo domain were the effects of Irradiation swolling and creep. The effect of Irradiation creep is to relax loads caused by on-power thermal bowing, while swelling act to bow the assemblies in the direction of increasing lateral thermal and flux gradient. Figure 4.2-90 11lustrates the Interassembly load pattern at ACLP due to on-power thermal bowing loads. Irradiation creep offects occur almost immediately, however, a delay or incubation period is required before swelling effects becomo pronounced. For the row 9 assembly, swelling became significant during the second cycle of operation (compare Figures 4.2-88 and 4.2-89 ) . Sudden Core Radial Motion The presence of Interassembly gaps at the above core load plano gives rise to the potential for Inward radlal motion of the coro assemblles. A positive reactivity insertion due to assembly motion requires a general radial inward movement of the core. A conservative design proceduro employed in CRBRP is to set the ACLP load plano gaps so that the maximum sudden inward motion of the coro assemblles within the confinos of the ACLP radial gap would result in a step reactivity insertion no greater than 60d when the reactor Is at f ull power. The assembly motion necessary to cause a step Insertion of this magnitudo is improbable. During reactor startup, the establishment of , temperature gradlents across the assemblies will cause them to bow generally 61l inward tending to close ACLP gaps in the core region in a predictable and controllable manner. Once the ACLP gaps are closed, further Inward motion of the coro assemblies is not possible. Consequently, tha only possible way for a sudden Inward core motion to take placo is for the coro not to compact radially inward at the ACLP as it is brought to power. The only non-compaction effects which have been identified and experienced in core array mechanical Interaction testing are the combination of high friction (p>0.6) and assembly azimuthal rotation. Consequently, in the analysis of the step insertion event it is assumed that during the insertion of assemblles into the core, the assemblles are rotated 59 uniformly so that the Interassembiy load plano gap is eliminated. O Amend. 61 4.2-223 Sept. 1981
It. is further assumed that the friction coefficient is sufficiently high that thermal bowing forces generated as the reactor is brou9ht to power will not be sufficient to overcome load pad frictional forces and realign the 1 assemblies. . Secondly, it is assumed that when full power is achieved, a seismic event occurs which produces forces sufficient to overcome the load i pad frictional forces and realign the core assemblies into their nominal orientation. The rotational alignment of the core is assteed to occur
; 'addenly, displacing the core radially inward. ; The conservatism of this procedure can be illustrated by examining j the assumptions which are employed in the step insertion procedure.
1 1) Core assembly load pads and the core formers are assumed to be at their nominal dimensions. Fabrication experience indicates that the as-built assembly load pad dimensions in the aggregate will be larger than noainal and that the core former ring smaller than nominal thus the as-built load plane gap is likely to be smaller
- than the nominal load plane gap. Furthennore, tolerance effects between adjacent assembly faces would result in the smaller of the distribution of interassembly gaps controlling the compaction process. The net effect of load pad tolerances would be to reduce the load plane gap.
1 1' 2) Load plane surfaces are coated with a low friction hard surface coating of chromium carbide. The mean friction coefficient of
- F- this surfa'ce coating in the reactor operating range is 0.2 to 0.4. -
Analytical and experimental evidence on core array mechanical simulations indicates, that for this range of friction coefficient, ! , non-compaction effects are not significant.
- 3) The rotational alignment of assemblies within the core is expected to be distributed statistically about the nominal orientation as i
determined by the core assembly and core former load plane as-built j surface dimensions. No operational bias has been identified which could preferentially orient the core assemblies so as to close the load plane gap but still permit installation and removal of assemblies into and from the core. l R,eactor e Assembly Bowing Reactivity i During a change in reactor power-to-flow ratio, temperattere gradients { ,
; change or develop across assembly ducts, causing the assemblies to bow.
Lateral motions of the core regions of these assemblies result ir. a reactivity i change. This reactivity change differs from that discussed in the previous paragraph in that it is assumed to occur in a predictable and centro 11able manner in response to duct temperature changes, i Figure 4.2-92A depicts the row average assembly bowing patterns and corresponding reactivity effects that develop during a power to flow ratio 59 4.2-2n Amend. 59 Dec. 1930
transition. At near-zero power-t>-flow ratios, the assembiles tend to bow freely within the constraints of the interassembly and peripheral load plane gaps. The presence of a peripheral gap at the TLP core former ring permits a not outward motion of the core region of the outer core fuel and radial blanket assemblies producing a negative reactivity ef fect es shown for the 0.2 power-to-flow ratio pattern. When the closure of TLP gaps from the outer blanket rows to TLP core former ring prevent f urther outward motion, the core regions of the outer core fuel and radial blanket assemblles bow Inward as shown for the 0.4 power-to-flow ratio pattern. The not inward motion of the fuel and radial blanket assemblies continues until the ACLP gaps close from the center of the core to the radial blanket rows. During this phase the bowing reactivity contribution is positive. Subsequent increases in power-to-flow ratio result in more complex S-shape bowing patterns and a net outward motion of the active core regions of the high worth fuel and radial ble-hot assemblies. During this phase, the bowing reactivity contribution is siegative as depicted for the power-to-flow ratio equal to 1.0 pattern in Figure 4 .2-92 A. 43 Figure 4.2-928 shows the bowing reactivity characteristics which are predicted f or nominal thermal, nuclear and load pad dimensional data at various times in the fuel assembly lifetime. Differences between the bowing reactivity patterns are attributable to the Irradiation ef fects of creep and swelling in the reactor assembly ducts. Early in fuel assembly life (125 and 250 days in Figure 4.2-928), the high worth assemblies at the fuel-radial blanket interf ece are bowed outward at the ACLP at refueling conditions as a result of Irradiation creep. On a subsequent reactor startup (power-to-flow transition), these high we-th assembiles traverse inwardly to a greatcr extent than initially straight assemblies. Later in the f uel assembly lifetime (500 days in Figure 4.2-928), the high worth assemblles at the fuel-radial blanket interface bow inward at the ACLP at refueling conditions as a result of 61 l swelling. The startup bowing reactivity late in the fuel assembly life resembles the characteristic behavior of Initially straight assemblies. Figure 4.2-92C shows the bowing reactivity characteristics which are predicted fc verious assumptions of nuclear and thermal data, load pad mechanical interaction and dimensional uncertainties. The curves in Figures 4.2-928 and 4.2-920, when combined with other significant reactivity ef fects such as the Doppler ef fect, are used in the reactivity feedback evaluations which are provided in Sections 4.3.2.8, Reactor Stabil Ity, 7.7.1.2, Reactor Control System and 15.1.4.5, Reactor Assembly Bowing Reactivity Considerations. Withdrawal Loads at Refuelino The frictional components of assembly withdrawal loads are obtained from the 59 NUBOW-3D analysis. The offects of uncertainties are combined O Amend. 61 4.2-225 Sept. 1981 k-
V tation is that the surrounding structure impacts the driveline and absorber at the guide points, thereby generating impulsive lateral Interaction loads which characteristically exist for less than a millisecond. Specific loading ef fects at the point of impact related to the presence of the coolant and the local dynamic deformation of the material cannot be separated uniquely. However, both are embodied in an " Effective Coefficient of Friction", applicable to impact load condition, which is expected to be substantially less than the design value of a 1.0. A test, described below, was performed 61 to determine this "Ef fective Coef ficient of Friction". Coefficient of sliding friction test data from sources external to CRBRP Indicate that friction coef ficients for material couples occurring in the control rod systems are substantially below 1.5 (Ref. 40). These tests were performed using a pin slider on a flat plate under steady loading. These data, contained f.. References 79 and 80, are summarized in Table 4.2-36A. This table contains data taken at temperatures ranging from 4000F to 11600F and represents the maximum observed dynw ic friction for each couple. The ETEC data for inconel 718/718 did not represent the maximum observed value during siIding, but rather the combination of the initial and final values. These data appear to he inconsistent with the other data for inconel 718/718. To resolve this inconsistency and to create a large body of applicable data for a verloty of material couples appropriate to the control rod systems, a series of tests on the sliding coefficient of friction using a pin slider on a m 61 l flat plate have been performed. These data, presented in Table 4.2-36B, represent the average value of the sliding coefficient of friction for each material couple for a variety of temperatures, pin pressures and over two stroke lengths. For certain couples such as inconel 718/718 and inconel 718/ 316SS the complete tests were repeated and two lines of data appear in Table 4.2-368. Average values of the sliding coefficient of friction are presented because these data are more appropriate to the conditions that exist during scram than the momentary peak values or the Initial and final values presented in Table 4.2-36A. Also, data in Table 4.2-36B has been presented as a function of temperatures and pin pressure to allow review of these functional variables. Since the data in Table 4.2-368 show no significant dependence on temperature, the upper three sigma data have been averaged over temperature for each of the material couples in Table 4.2-368. The results are shown in Table 4.2-360 for comparison with the prior, less extensive Table 4.2-36A. These data demonstrate a dependence on pin pressure with the lower pin pressures having the higher coefficients of sliding friction. This pressure dependence occurs frequently in pin on plate measurements and may not be a material couple characteristic. There also appears to be a slight correlation with stroke length, in this case, the longer stroke length of .750 inches 59 would be more appropriate for the control rod system during scram. O O Amend. 61 4 .2-23 8 Sept. 1981
In summary, the data presented in Table 4.2-36B and 4.2-36C repre-O sent the most recent and most appropriate values of sliding friction data for the material couples used in the control rod system. The upper three sigma values of Table 4.2-36B, as selected for the most appropriate analysis conditions, are the recommended values for design analyses. Tests were performed to determine the effective coefficients of friction under dynamic impacting conditions such as would occur between the PCRS fixed boundaries and the translating elements during a seismic event. In these tests, the effective dynamic coefficient of friction was determined by solution of the equations of motion based on measured impact loads and drop times of test rods subjected to lateral dynamic excitation (Ref. 182). These tests were simplified simulations of the circular PCRS driveline and hexagonal control rod oscillating within their respective constraints. To reduce complexity but still retain the principal dynamic effects on the coefficient of friction, the test articles were a straight circular rod traveling through three bushings, and a hexagonal rod traveling in a hexagonal duct. Bushing clearances were chosen to provide both lateral and rotational impact under lateral dynamic input similar to PCRS driveline response to seismic excitation. The hexagonal rod to duct clearances were typical of PCA design clearances. Material couples were prototypic of the PCRS (i.e. , Inconel 718 on 51 59 Inconel 718, circular, and Inconel 718 and 316 SS, hexagonal). O l l l Amend. 59 O 4.2-239 Dec. 1980 m
l l
\
O l I TABLES 4.2-24 through -29 61 HAVE BEEN DELETED O l l 1 i I l l l 4.2-368 (next page is 4.2-374) Amend. 61 l Sept. 1981
i g
> CHIMNEY lJ\ N-- >
V d v
)
M's -
\
k O; k, e INSTRUMENTATION CONDulT
/
Cb "
'%s s J \
/ '
, h
/
[ \ Ag,
,k\
sAsic STaucTURE ~ v\ LOWEn SHROUD TUBE INSTRUMENTAT10N POST b ) O "'"" 4.2 45 0 instrumentation post Installation Amend. 51 sept 1979 4.2-534
L E R R A B e E N R 0 1 O T C 8ED 1 7 Y O
, N L
L A A
/ e
[/ , E R R O E C M e RR E r u T N - WF O L O S t t c u r R , _ r E e S -
- n n
N I o Y F A W 1 $ J L e r Y , L G AI G C o E K D N N I R T R DN A l0 R L DE 6. 4-2 O A O E IH P - X 4 Y I S A P F e U r W S u g Y E K pe y i F n . f s E R O R ' { E C M RR EO P P F U T N AI YS R E T K S E R $Gi 3"{b O
".$ Ep !
!~
/
Q b d. The gamma background at the detector location. _ The gamma background will be a primary factor it determining instru-ment sensitivity at a given location.
- e. The ability of the SRFM to accurately 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 for sub-criticality detennination.
Extensive analyses at ARD and both analyses and' experiments at' Oak Ri.dge 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 signi-ficant 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 shut-down 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 O' L 54 4 counts per second at each BF3 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 (45% below 0.1 ev) at the SRFM location, eight inches behind the front face of the graphite block. This enhancement of the thermal flux inside-the graphite block has been con-firmed 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 Keff. For these reactor configurations, the flux level at the SRFM varied 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 M% analyses have shown that the flux monitoring requirements for the s H can be satisfied with the background 51- associated with a maximum d%Me iucl assembly withdrawn to a point l- 61 64 inches where its fully i nee position for any core location (,,/ 4.3-9 Amend. 61-Sept. 1981'
The count rate due to background is minimized by shielding in the form of boron carbine slabs which surrounds all sides 6f the graphite block except the %nt face. The shielding is used to reduce the count rate 59 from neutrons rhich are scattered into the graphite block from the reactor cavity walls. Normal refueling procedures will require that no assembly of any type be in the FT&SA while 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 59 lofthecore. For this case, the requirement of (c) above are imposed. 54 Ihe gamma dose at the SRFM location immediately after shutdown has been analyzed in detail to assure that the sensitivity of the BF neutron counters is not adversely affected. The type of BF3 neutron 3 54l 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 sensivitity falls off rapidly. Calculations have shown that the local gamma dose rate at the SRFM Tocation is less than 100 R/hr with appropriate shielding in front of the graphite olock 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 instrumentati , near critical. The recomended 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 l kinetics rod drop (IKRD) technt,ue. Second, the calibration constant, which
- relates the subcritical reactivity to the count rate, must be determined.
l This is accomplished by inserting the previously mecsured reactivity worth (the same control rods described above) and noting the corresponding count rate. This same calibration constant is then used to imply subcritical reactivity when all the control rods are inserted and the reactor is fully
.autdown.
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 performed numerous rod-drop experiments (Reference
- 2) in the Tower Shield Facility 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 effect of neutron streaming in the reactor cavity. Future analyses will investigate these reactor cavity effects 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 fuel reloadings. The relationship between the steady-state SRFM detector count rate and the subcritical reactivity 51 is derived from the point kinetics equations: Ament. 59 Dec. 1980 4.3-10
_ ._. -_. _ . . _ ~ .- _ . _ _ _ _ _ _ . - _. _ i i. 1 lO TABLE 4.3-43 j j ZPPR-7 CONTROL R0D WORTH CALCULATION-T0-EXPERIMENT RATIOS
- Beginning-of-Life End-of-Life Phase B Phase C RG. 4 0.916 (0.963). 0.906 (0.973)
Row 7 - Flat 0.898 (0.987) 0.887 (0.952) , Row 7 - Corner 0.992(1.074) - 0.905 (0.986) i 'O - i 4
- Calculated with standard two-dimensional (hexago-11 ' planar geometry) coarse-nesh direct eigenvalue difference diffusion theory methods using 9-group ENDF/B-III data. . Values in ( )
51 , from four-mesh per ZPPR drawer diffusion calculations. i i i i l l Amend. 51 i Sept. 1979 t
- () 4.3-144
, ,r,w,% ,--oe.v-y-,,,-eye 4 ,gy,y-y,%y,v,-,-w,.--,,%e,ge -, 3 y r. y , -, - ,, -, ,y w .w y y y vm ww.r w e y ww-y,,,,,,-gir + w w -w g--=,,erve w ww-w w..gw-.**-ttww er w - w - erwv v +w, v ~= i m
TABLE 4.3-44 COMPARIS0N OF NUCLEAR PARAMETERS FOR CRBRP AND FFTF CRBRP FFTF LAYOUT Number of Fuel Assemblies 156 73 aer Enrichment Zone 28 voter Enrichment Zone 45 Number of Test Loop Locations - 9 Number of In-Core Control Rods 15 9 Number of Inner Blanket Assemblies 82 - 61l Number of Radial Blanket Assemblies 132 - Number of Radial Reflector Assemblies - 108(I) Number of Removable Radial Shields 306 - DIMENSIONS Assembly Pitch (meters) 0.1209 0.1198 Core (2) Equivalent Diameter (meters) 2.019 1.200 Core (2) Cross-Sectional Area (meters) 3.203 1.1 31 Active Fuel Height (meters) 0.9144 0.9144 Height-to-Diameter Ratio 0.453 0.762 56 l Axial Blkt. Height, Upper / Lower (meters) 0.3556/0.3556 - Inner and Radial Blanket Height (meters) 1.6256 - INITIAL CORE ENRICHMENTS AND FUEL MASSES Enrichments (Pu/U+Pu) Inner Enrichment Zone 0.328 0.224 Outer Enrichment Zone 0.274 Enrichment Ratio (Outer / Inner) N/A 1.22 Isotopic Composition of Feed Plutonium Pu-238 0.0006 - Pu-239 0.8604 0.864 Pu-240 0.1170 0.117 Pu-241 0.0200 J.017 Pu-242 0.0020 (I) Includes positions for as many as fifteen peripheral shim rods. 0.002 h 51 (2)
" Core" includes fuel, inner blankets and in-core control rods.
Amend. 61 4.3-145 Sept. 1981
F t l d
- 37 ft4 in. : : 37 fte in. :
+ + 16.75 in.
. :p;
.3.' ;,[, . a., f_'g; '. . it,
+- 6.0 in. g,
-* k WATER 188n- DUMP PUMP TANK J
- SUCTION f _3 S in. EVAP -
j E
' WATER TO d ,ATER 10in.
10 in. R ECIRC PUMP DUMPTANK DISCH TO EVAP 12 in. STEAM 10 in. RECIRC PUMP DISCH DUMP f " " " "' y
'. "A" s ( TO SH f TO EVAP "B" 'l fi.* .' " . .
j h \ ( .* . s q%.* j.: g . me , fl _ _,. .._ -DI ^
. . PS
_ Y~ . 16in i
- i 16in. t VENT 5
' I j eJWRPRS
~
4
'd 18 in. SH TO SWRPM VENT lI ,
9 VENT ,? EVAP In )-TN-- .+--- ' f- l
. .}-
g I I -.y
-.I 24i '
N - w -_-+ 6 M ERHEATER, gp .) , ,, g; [' 24 w _,
; ,e i - 6
% - ~ '
s TO -t,DQ- - *-
~lh TO 24in. g SWRPRS EVAP"A" EVAP "B" a
- i N -
' . Eh Vi d lE
,V, '
11 ft-6 in. SWRPRSVENT d k . g - m- , i l 24in.SWRPRSf ff VENTFROM
- -- h e 244n.
SWRPRS VENT
^
_1
._ I l'j _'
i EVAP . FROM EVAP " I'h ~ l l 4 fte in. -.+> d 4 ft4 5 l 18in, s
.I.
g 18 in; *-+
---g D
O-SOOlUM DUMP ' ' ' EVAP .1l .
- . . J 1] d
~
18 ft4 in. TANK TO [_;-- MIXING s
--+-- . -
' ~ * ' ' t --
*/
.. TEE f *--- ' i p +- 18 fte in.-M e f 1: . e .e. k: . a.c. h / ,, ; f . 3 4d.'
24 in. MtXfM6 TEE
/
}24in.N l 24 h. Na COLD t
C. TO Na PUMP LEG FROM LEG TO ! STEAM Na PUMP GENERATOR CELL FLOOR EL 7334 ft. FLOOR EL 7664 ft. l L
s e i 1 4 R
- . g:, :: A u 30 gr0 7. ,;, ;, y, l.0 in _. _ __ - _ _ , __
y RUM} k. m ' a.-2
, 4 DRUM
- _ .. } l 22 in. TO SH W4. .' <. -
ll 1: 16 6n. h
'I; l
i
~ T;; , ;.g . , . rj:, ,,
, j. -- j
^
^
4;= ,
.y 'e F PER- i 163n.~
C h.'SWRPRS VENT TO - FLARE IN -r I-STACK ( ) 5 fte in] c-EVAP"A" . EVAP "B"
< -- > . < -- > j I _[ .
I t
. . < 3
-tg ii 2 n
n ii , 2 n. T MP E TANK I , n= m~.w'
- k.
- l u . ,,_ ..
- 27,t4 6n. : '? STEAM GENERATOR BUILDING T MAINTENANCE BAY NOTE: DIMENSIONS ARE FOR REFERENCE USE ONLY FLOOR EL 8060 ft. - ROOF PLAN - (ABOVE EL 8734 ft.
81 288-01 Figure 5.1-Ic. Intermediate Heat Transport System General Arrangement (Sheet 2 of 3) Ar.iend. 61 5.1-17d Sept. 1981
'l
' s l I
.\
'.i
[ l PACC UNIT 'Q2 N4 in.
-EL 901 ft4 n.
DISCHARGE TO ATg 4 SPHERE
- ... m - EL 8m N4,n.
u *
%%i 1 l
EL B73 tte en - pg , l , l h l "!
~ EL 862 ft4 h ,
-+"
c; RELaEF W ALVE a
' " c-l I
I m _ Fa^h .; ,
? '
#[EAM DRUM f IHTS 16 in. STE AME I-
~' SODIUM SODIUM
, TO TURBINE --
PUMP > * - 3 CONDENSATION E XPANSION f , 5 4 _$-/ RETURN FROM PACC TANK P, "- i m , 24 < h, SUPE RHE ATE R f l8 m STIAh 10 m. F E EDWATE R ( l t, ) TO DRUM -
% { J
).%J.f{} C+ [ TOSTEAM
. 7 DRUM D- Z
~
ROM 4
'* AP "'8" lo n
'E l ,{ . j
- 2 1
[i 'C
- C. I
- 24 in. Ne [,e _ g , . i
{l i GR ADE EL 815 h4 in. g PUMP ;j ;., . 1 7 DISCH 26 m .
- E L 806 h 0 h_ _ ,- ,
TOP OF STEE L
- q ,
L, 6 ~,_ _ _ - ,19 in,. UM , l
- l ij P EL 794 h4 m.- *'~
+h
~d g . I +
SUCTION f i'
' 24 in. No F ROM MISING -
8A g TEE TO PUMP ~ 2 'l TO F~** r: v P . i 24 en Na J
"= ^
IO in. Na COLD LEG COLD LEG [ l,- b 4 ; _ a $ g 3 .EVA R ORp TO MIXING TEE hj /qi~F#T I
@ 12 in. ] 'y TANK ECC "
- 1. ; C .7
{* = - O fo C T:[_"((__ E .{
! 1 j
! 24 m. Na Igs ,
['. HOT LIG *
,,., l _1 =
- Et nsNe w c:'
_ _ . . J. __ _ ,y ,, i 34 in * ,* } s ; -
, g3* TO SODIUM DUMP SWRPRS i; TANK TANK
~'
. e I L 4 i
- AXfLIARY BAY
- F'-P - ~ i :
r e EL m No
. .. v. .
. .... ..- ... - - 5;. c, - -
TYPICAL . TYPICAL CROS$ SECTION FROM STE AM GENERA 10R CELLS ONE. TWO AND THREE f g l
\
1 l l l t 1 8
' l 16 in. SWRPRS VENT TO ATMOSPHERE l4
, l
-- EL 9IE fte in EL 9Ct ft4 m.J T I PACC UNil TOPOF STACK STE AM GENE RATOR CELL E L 888 ft4 in. J DISCHARG : e VENTILATION EXHAUST TO TOPOF VENTS GE"I STACK TO f ATMOSPHE RE RELIEF b ALVE DISCH) AT p EsE lp EL 886 N4 :n.
- E L 888 N4 4n.
-~
. . . . . Y _,_- .
R ~Pocc-
, _- 7 l'.4CC UNIT
- 1 R-4 PACC VENT F** SYSTEM 52 I !
. o
'[ f }
- E L 862 ft. 0 m.
f , II ( ; - E L 862 ft4 m. f '.,' 2 FLOOR 8 g i j . -- , 4
-7 J
[ 3
< STE AM TO PACC 241n. SWRPRS * -- -
p 'p",E ifENT TO . ; ACK k
"A}N"e- .;- to in STE AM TO TURBINE f
I'll l' f
~
~
) DRUM y. $5TE AM DRUM'-
.! . - . - .. - - . . . r%- '
EVAP %" --- - ----
- 4f t g E f E g
* -'~ C'1 m. "
I * + SJPE HEA ER [ TO E GRADE EL 815 ft4in. g' gg , P
;p f ,
l VENTURI l I f FLOWMETER I wv *' Y ILOOP 2 ONLY) Il ' ( k EIl
- l
- EL 806 ft4 4n. E L 806 N4 m.
.. TOP OF STEEL f [ , , _. _ :. TOP OF STEE L
! i y~
t~ *]1l7 ?' SUCTION
~
d 18 en. PUMP C '
. i s in. PUMP SUCTION j: . 6 in (
, = -
18 in. Ne
'7 s1 us _ =3 g
i FROMSH s M, - E L 782 ft4 in. 12 in. 12 *.- ~- - ;l j , EVAPORATOR TOP OF HE AD g <1 .p lo in. PUMP , (OUTSIDE) FLOW DISCH WATER DUMP s) p-- , NO2ZLE I PUMP \ . TANK I LOOP 2g , RECIRC ' Q ,. _ _ t -l ONLY ] PUMP
- EL 765 ft4 in.
&h[ - EL 765 fte m-FLOOR
... . g L TWO l
L. 4] , PROTECTED WATER j f l!l l STORAGE TANK g SWRPRS TONK 7\ , . f
- E L 746 '.4 m.
1 FLOOR
! 7 \
i . . ,I i.
/'
.; - ' ' 722 a
- 6 -
g.
. .; . , Et,7g4 in.
g MOTOR-DRIVEN TURBINE-ORIVEN AFW PUMPS apwpuup (LOOP TWO ONLY) (LOOP ONE ONt Y) l lROSS SECTION FROM STE AM GENF RATOR CELLS SECTION THROUGH AUXILIARY 8AY ONE,7WO ANO THREE 81-288-02 Figure 5.1-Ic. Intermediate Heat Transport System General Arrangement (Sheet 3 of 3) Amend. 61 5.1-17e Sept. 1981 e d
/s
Riser Elastomer Seals ( The balance of the seals on the riser assembly operate at temperatures below 41 1250F. Uoner Internals Structure Jackina Mechanism The UlS Jacking mechanism utilizes metal buf fered seals in the 4000F areas. These seals are part of the mechanical assemblies. The seals will be removed with components at the appropriate maintenance period. Elastomer seals are 41 located in the cooler regions, have a service life of five years, and will be 57 replaced using hands-on maintenance. Llauld Level Monitor Ports Pluas Four of these components, operating at 4000F, are located en the reactor vessel head and provide receptacles for holding the liquid level monitors. Three small port plugs are attached to the top surfaces of 1he closure head rotating plugs by partial penetration welds, two on the Intermediate and one on the large rotating plug. One large port plug is bolted to the top surf ac) of the large rotating plug and is sealed to the plug by double metal "0" riags. The 25 61 l seals remain attached to the port plug during installation and removal. Because the port plug remains stationary relative to the head assembly, the metal "0" rings beneath the plug flange are not expected to require 57 maintenance. O 5.2.1.4 Guard Vessel V The guard vessel provides for the retention of the primary sodium coolant in the event of a leak in the portion of the primary coolant boundary which it surrounds. The guard vessel geometry assures reactor vessel outlet nozzle submergence after such a leak which will maintain continuity in operasing primary coolant loops to provide core cooling. The guard vessel also provides a uniform annulus for in-service Inspection of the reactor vessel, with clearances that preclude contact with the reactor vessel and piping under accidant conditions. Insulation for the reactor vessel and a heating system for the reactor vessel to be used prior to sodium fili and during prolonged shutdown are also mounted upon the guard vessel. The maximum and minimum widths of the radial gap between the guard vessel and l the reactor vessel have been conservatively calculated, taking into account all relevant factors such as tolerances on the diameters of the two vessels, permissible out-of-roundness of the two vessels, possible daviations from straightness due to manufacture and subsequent operation, thermal expansion, initial deviations in the alignment of the two vessels, etc. The transporter for the television camera will be designed to accommodate itself to this maximum possible range of gaps as it moves in the space between the two 25 vessels. 5.2-4a Amend. 61 Sept. 1981
\
b.2.1.5 R* actor Vessel Preh:at The Reactor Vessel Preheat System will control the dry hect-up and cool down of the Guard Vessel, Reactor Vessel an. internals between ambient (70 F) and 400 F and if required will provide make-up heat for that lost to the Reactor Cavity during prolonged shutdowns. The heat will be provided by tubular electrical heaters mounted between the Guard Vessel and insulation. These heaters will be arranged circumferential1y around the Guard Vessel and will be grouped and controlled in zones of uniform heat output. Temperature sensing devices will monitor the Guard Vessel temperature in each of these zones and provide the necessary feedback for power level adjustments in the heaters. The heaters will be mounted to the same framework which supports the Guard Vessel insulation. Ceramic offsets will be used to offset the framework and heaters from the Guard Vessel surface. The heaters and framework will therefore be electrically isolated from the Guard Vessel . Convective barriers, reflective sheaths and the Guard Vessel insulation will be used to optimize heat input to the Guard Vessel and minimize losses to the Reactor Cavity. Preliminary preheat, startup, and shutdown analyses have been performed on the Reactor Vessel and Guard Vessel to determine the temperature differences which will result in opening and/or closure of the annular gap between the two vessels. By necessity the preheat analysis is very preliminary since no firm preheat procedure has yet been developed. Figures 5.2-4 through 5.2-6 show the temperature differences between the Reactor Vessel and Guard Vessel in the inlet and outlet plenum regions for the three transients in question. As shown the largest positive temperature difference between the Reactor Vessel and the Guard Vessel occurs in the outlet plenum region during startup (335'F) while the largest negative temperature difference occurs in the outlet plenum region during shutdown (-214 F). The nominal radial gap between the reactor vessel and guard vessel is 8 inches at assembly and at the end of preheat. This gap decreases to approximately 7.6 inches minimum during start-up and increases to approximately 8.3 inches l maximum during shutdown. During preheat the gap also increases but to a 1.sser value than during shutdown due to the smaller maximum temperature dif ference. Variations in the axial gap between the bottom of the reactor vessel and the inner surface of the guard vessel are noted between the states shown in the table. Thus the largest axial gap is 11.0 inches at the dry cold condition and the smallest gap is 6.2 inches at the end of the neating phase of preheat. g 5.2.2 Design Parameters Overall schematic views of the reactor vessel, closure head assembly, inlet and o. .let piping, and guard vessel are shown in 56 Figures 5.2-1, l A and 18. The top view is given in Fiaure 5.2-2. O 5.2-4b Amend. 56 Aug. 1930
O 41l 17l V 58 36 5.2 . 2 Closure Head The closure head consists of three rotating plugs which are constructed of SA 508 Class 2 steel. Each plug contains a major penetration eccentric to its outside diameter. These rotating plugs are interconnected 17 l by means of a series of plug risers. Sealing between the plugs is accomplished by sodium dip seals and double inflatable seals of elastomer material. At 36 its top, the large rotating plug has an outer diameter of 257 in., and an inner diameter of 176 in. The large rot.-ting plug provides access to the - vessel liquid level interior for the ex-vessel transfer machine and the core coolant monitors. The intermediate rotating, plug (175 in. 0.D. and 68 in. I.D.) provides access to the vessel interior for the control rod drivelines, upper intervals support columns, and the liquid level monitors. 41 l The small rotating plug (67 in. 0.D.) provides access to the vessel interior for the In-Vessel Transfer Machine. The thickness of each rotating plug is 41 l 22 in. Rotation of the plugs will be accomplished by a gearing and bearing system attached to the plug risers. The nozzles for each penetration 41 l are constructed of an austenitic stainless steel. 58 Each rotating plug is provided with a mechanical lock O and electrical interlocks which prevent plug rotation during reactor operation and refueling when plug rotation is not desired. The mechanical locks include the following:
- a. Each plug includes _a separate positive lock to assure that the plug cannot be moved, and will not drift from its normal operating position during reactor operation.
This lock will be installed to prevent relative rotation 58 l between each bull gear and the bearing outer riser whenever the control rod drivelines are connected. The locks shall be manually installed at the end of each refueling cycle, and will be removed only during the refueling period when plug rotation is necessary.
- b. The plug drives are designed to be self locking to react to any seismic torque occurring during refueling, which could rotate the plugs and thus damagc a fuel or blanket assembly during removal from the core.
4 The electrical interlocks include the following:
- a. During reactor operation, the r!ug drive and control system keyswitch is in the OFF position, the control system is deenergized, and there is no power to'the plug drive motors.
~ Amend. 58 5.2-6 Nov. 1980
- b. Electrical Interlocks are provided to prevent the plugs from being inadvertently rotated by their drive system unless the upper Internals are raised and locked, and the IVTM and EVTM are in a safe condition.
- c. An electrical interlock is also provided to prevent vertical operation of the IVTM or operation of the EVTM over the HAA during operation of 24 the plug drive system.
56 Each rotating plug has attendant thermal and radiological shielding extended to a depth of 74.65 in, beneath the top of each plug forging. The shielding is composed of a series of plates f abricated from carbon steel, and stainless 17l steel . The cover gas between each set of plates attenuates thermal conduction and thereby acts to decrease the heat flux Imparted to the rotating plug. A heating and cooling system is provided to maintain the closure head at 400o (nominal) as well as providing heating and cooling for other small head 17 mounted subassembiles. A gas entrainment suppressor plate assembly is positioned beneath the head 45l17thermal and radiological shielding at a depth of 122.65 in. beneath the top of each rotating plug. It protects the head shielding from being contacted by the core coolant and minimizes the amount of cover gas entrained in the core 17l coo; ant. The assembly is designed to accommodate all normal, upset, emergency, and faulted conditions. sd l2s 42 in plan view, the subassembly consists of 33 plates at the same elevation with horizontal gaps between them. (Fig. 5.2-3) These plates have penetrations in line with the head penetrations to allow thc passage of the head mounted components into the cutIet plenum. Each plate is supported by means of a y central support column af fixed to the lower shield plate. These central columns, when possible, consist of tubes which surround closure head penetrations. Support columns which do not surround penetrating equipment J w'lI be capped to minimize the amoun+ of cover gas entrained in the sodium 611 pe< The support columns will be Inserted through eversized penetrations in the lower shleid plate, accurately positioned and then attached to the top surface of the Iower plate by rneans of bolting. The support columns wIII be attached to the suppressor plate by means of welding. This attachment weld is located above the region of the suppressor plate where high thermal gradients occur by using a plate with an extruded weld neck. The top end of the support column, which protrudes through the lower shield plate is composed of 21/4 57 42 Cr-1Mo. material to minimize the differential expansion with the carbon steel shield plate. The lower, in sodium, portion is austenitic stainless steel. l The use of a single support provides adequate support while lessening the thermal stresses by permitting the plates to flex freely under the expected 2 thermal gradient. l l l l 9 5.2-c,a Amend. 61 Sept. 1981 l
O O O Table 5.2-1
SUMMARY
OF CODE, CODE CASES AND RDT
< STANDARDS APFLICABLE TO DESIGN AND MANUFACTURE OF REACTOR VESSEL, CLOSURE HEAD AND GUARD VESSEL Closure head' Pressure Internals Guard 57 Component /Cetteria Reactor Vessel Boundary (as appropriate) Vessel Section Ill. Addenda thru Winter Addenda thru Winter Addenda thru Winter Addenda 1'hru ASE Code, '74- '74 '74 Summer '75 1974 Edition ! Class 1 Class 1 .ss 1 Class 2 "
ASE Code Cases 1521-1,1592-2.1593- 1682,1690 1521-1 1592,1593,1594 0,1594-1,1596 1, 1592-4,1593-1 if electa- by sup-
- 1596-1,1682,1690 pller 1521-1 &
1682 RDT Standards E8-18T, # 3 E15-2!E-T, 11/14 EIS-NB-T, 11/74 E15-ta-T, 11/74 1 57l Mandatory E15-2ts-T, 11/74 Amend thru 6/75 Amend thru 6/75 Amend ih.u 6/76 Amend thru 1/75 7 F2-2, 8/73 F2-2, 8/73 F2-2, 8/73 F2-2, 8/73 i m Amend thru 3/74 Amend thru 7/75 Amend thru 7/75 Amend thru 7/75 61l F3-6T, 12/74 F3-6T, 12/74*** F9-4,9/74 F3-6T, 10/75 F6-5T, 8/74 F6-5T, 8/74 F6-5T, 8/74 Amend thru 2/75 Amend thru 2/75 Amend thru 11/75 F7-3T, 11/74 F7-3T, 6/73 F7-3T, 6/75 m F9-4T, 9/74 M1-1T, 3/75 F9-4, 9/74 , I 5.~ M1-2, 3/75 F9-4, 9/74
~ Amend thru 7/75
*For those reactor vessel and closure head components Internal to the pressure bounr.ary special purpose high cycle f atigue 57 curves and creep damage rules have been developed as discussed in Appendix 5.2A.
Table 5.2-1 (Continued) Closure Head Pressure internals Guard Component /Crlterla Reactor Vessel Boundary (as appropriate) Vessel Sj RDT Standards M1-1T, 3/75 M1-2T, 4/75 M1-4T, 3/75 M1-4T, 3/75 M1-6T, 4/75 Amend 1-7/75 M1-6T, 4/75 M1-10T, 3/75 M1-10T, 3/75 M1-11T, 3/75 Amend 1-7/75 M1-11T, 3/75 M1-17T, 3/75 M1-17T, 3/75 M2-2T, 12/74 y M2-2T, 12/14 M2-7T, 3/75d H2-5T, 1/75 M3-10T, 7/. Amend 1-2/75 M2-7T, 2/75 M7-4T, 3/75 m M2-18T, 4/76 5] M2-21T, 12/77
$ M3-6T, 3/75 M3-7T, 4/75 MS-1T, 11/ 74 MS-2T, 5/73 MS-3T, 12/74 Sj MS-47, 1/75 w to M6-3T, 2/75 P 5. M6-4T, 2/75 5m co ~
M7-3T, 11/74 Non-Mandatory F9-5T, 9/74 F9-5T, 9/74 F9-5T, 9/74
**Functionsf ly designated Class 2, and constructed to rules f or C1sss 1, but not hydrostatically tested or code stamped.
g mExcept for the three rotating plugs, for wh!ch the applicable Issues are F3-6T, 3/69 for LRP & 61 SRPs F3-6T, 5/74 for IRP. M2-7T, 2/69 for LRP & SRPs M2-7T, 2/14 for IRP. O O O
The IHTS piping will be supported from the building structure with constant I ad support hangers and rigid rods and wilI be restrained with seismic [v) 61 snubbers. Attachments to the piping for supports will be of the clamp type on the outside of load bearing insulation. If any attachment requires direct support to the pipe full penetration welds will be used. Piping penetrations through the Reactor Containment will be a flued head, rigid type seal. P; ping penetrations through the Steam Generator Building will not provide leal. tight seals. The piping within the lHTS consists of large sodium containing piping which must be Installed per detailed drenings and rigic quality assurance requirements. There is no piping chich can be field run. 5.4.2.3.4 Intermediate Heat Exchanger The CRBRP Intermediate Heat Exchanger (IHX) serves to transfer reas or thermal energy from the radioactive primary sodium to the non-radioactive Intermediate sod'um. The lilX is a counterflow shell and tube type unit with a vertical or! Station in the plant. The design arrangement provides for downflow of the coolsd (primary) fluid and upflow of the heated (Intermediate) fluid to enhance natural circulation for reactor decay heat removal. A detailed description of the IHX design is given in Section 5.3.2.3.2. 5.4.2.4 Overoressurization Protection n The IHTS has no isolation valves within the normal circulation path so that isolation of individual system pipe sections or components is not possible. for some reason the system necomes blocked, the intermediate pump wilI not If overpressurize tre system as the IHTS structural design is sufficient to withstand the purcp shutof f head. In the event of an argon supply valve f ailure, the system would not be overpressurized as the combination of argon supply pressure of 115 psi and pump shutof f head would not exceed the IHTS design pressure. This is true even if the pump shutof f head associ etM with the PHTS pump were reached instead of the IHTS pump shutoff head. The system may be suujected to overpressure in the event of a water or steam 40 le k in the Steam Generation System. For large or intermediate sodlurrr-water reactions, the resulting pressure increase due to the formation of reaction products in the faulted evaporator or superheater module is relieved through I rupture disks. (See Section 5.5.2.4). 5.4.2.5 Leak Detection System l 5.4.2.5.1 Leak Detection Method _s i The methods used to detect Liquid Metal to gas leaks from pipes and components l of the IHTS are aerosol detectors, cable detectors, contact detectors and visual inspection with back up from smoke detectors. See Section 7.5.5.1. 28 l l (D V 5.4-10 Amend. 61 Sept. 1981
A sodium level monitoring system is provided to monitor any leakage 56 between reactor and intermediate coolant occurring in the IHX. The method is described in Section 7.5.5.2 in detail. 5.4.2.5.2 Indication in Control Room Audible alarms will be sounded in the control room as described in Sections 7.5.5.1 and 7.5.5.2. 5.4. 2. 5. 3 IHTS Coolant Volume Monitoring The IHTS coolant volume is monitored by the level indicators in the ' IHTS pump tank and in the expansion tank. Details are discussed in Section 7.5.5.2. These monitors coupled with the sodium temperature measure-56 ments allow monitoring of the total sodium in the IHTS loops. Small leakages of sodium from the IHTS can be replaced by use of the sodium fill system. 5.4.2.5.4 Critica' Leaks Critical leaks are discussed in Section 5.3.2.5.4. Detection 29 capability is discussed in Section 7.5.5. 5.4.2.5.5 Sensitivity and Operability Tests 56l Periodic maintenance will provide for checking the operational readiness of leak detectors. During installation and checkout, the correct electrical functioning of each leak detector and level detector will be tested. ( 5.4.2.5.6 Confinement of Leaked Coolant If there is any leakage from the IHTS in the RCB it will be confined as described in 5.3.2.5.6. Any leakage from the IHTS in the SGB will be l29 contained in the catch pans and will be detected by the leak detection system. Fires as a result of sodium spills are evaluated in Section 15.6. Leaks h8 in the IHX are still contained in the passive coolant boundary, and no leakage 56l into the RCB will result. 5.4.2.5.7 Intermediate / Primary Coolant Leakage Primary to Intermediate coolant leakage is very unlikely due to the higher operating pressure of the intermediate system. The IHTS pressure shall be maintained at a minimum of 10 psi higher than the PHTS pressure at 41 all points in the IHX during all normal modes of operation. Intermediate to primary coolant leakage detection is described in Section 7.5.5.2. 5.4.2.6 Coolant Pur1rication (IHIS) The IHTS coolaat purification is accomp:1shed by six cold traps, two in each of the three loops. All six traps are normally in operation, however, operation 58 l of a single trap per loop will still maintain required Na purity. These cold O Amend. 58 5.4-11 Nov. 1980
O O O Table 5.5-7 (Continued) I i SGS Ple'ING AND THEIR DESIGN CHARACTERISTICS NO. NO. ASE CODE
- COMPONENT PER PER SEC. III DESIGN PIPING AND HEADERS SIZE LOOP PLANT CLASS REQUIREENTS 54 l Separator Tank Equillzer 24", sch XS 1 3 3 125 psig, 800 F 4
- 3. Sodium Dump Subsystem l Sodium Dump Tank Inlet Piping j' f/ Dump Valve 4", sch. 43 5 15 2 50 psig, 965 F
, Sodlum Dump Tr.nk Vent Line to Stack 8", sch. 40 1 3 3
- i Sodlum Dump Tank Equilizer j Gas Line to isol. Valve 6", sch. 40 1 3 2 50 psig, 700 F 1 4 Water Dump Subsystem & Relief Lines Steam Drum Relief Valve inlet Piping 6", sch. 120 2 6 3 900 psig, 535 F 8", sch. 40 2 6 3 300 psig, 420 F
- , Steam Drum Rollef Valve Discharge j
Piping 6", sch. 80 3 9 3 900 psig, 840 F i ? Evaporator Relief Valve Discharga g Piping 10", sch. 40 2 6 3 300 psig, 420 F 8", sch. 40 2 6 3 300 psig, 420 F ! 6", sch. 80 4 12 3 900 psig, 535 F Evaporator Water Dump Valve inlet l 6I Piping 4", sch. XXS 4 12 3 2400 psig, 650 F Evaporator Water Dump Valve Discharge 6", sch. 80 2 6 3 900 psig, 535 F i Piping 10", sch. 80 2 6 3 900 psig, 535 F
- Water Dump T,enk Discharge Piping 6", sch. 80 1 3 3 300 psig, 420 F Water Dump Tank Inlet Piping 10", sch. 80 1 3 3 900 psig, 535 F 59
- Design requkemeMs will be gom amer N waluaMon of #ansients.
4j ! n 3g : < = CL k w
TABLE 5.5-7 (Continued) SGS PIPING AND THEIR DESIGN CHARACTERISTICS NO. to. ASE CODE COMFONENT ER PER SEC. til DESIGN PIPING AND HEADERS SIZE LOOP PLANT CLASS REQUIREMENTS
- 5. Drum Blowdown 59 Drum to SGB WalI 6", sch. 160 1 3 3 2200 psig, 650 F
,m 6. Na-H2O Leek Detectors m IHTS to Leak Detection Sodium Isolation Valves 3/4" 4 12 2* 325 psig, 985 F i e 59 l 47 isolation Valves to Leak Detection Modules 1/2" 4 12 3 325 psig, 985 F 47l 41
- Designated ASE lil, Class 2, optionally upgraded to Class 1.
EN Pe O O O
. . _ _ . . - _ _ _ _ _ _ _ _ _ - - _ _ _ _ . _ - _ . . - -- - - - . . - - . _ . - - . . _ - . ,=_. .-
4 i i 4 i Table 5.5-10 i SWR DESIGN BASIS ) i f ASME CODE CATEGORY l i OTHER STEAM GENERATORS AND IHTS EQUIPMENT l - LEAK DESCRIPTION (1) FAILED STEAM G:NERATOR AFFECTED RPST IN THE AFFECTED LOOP 44 Smali Leak In One Tube Upset Normal Upset One EDEG# Followed by Two Faulted Faulted Ettergency Additional Single EDEG j Failures at One Second i 61 Intervals (Total 3 EDEG's) (1) See Section 5.5.3.6 for detailed descriptions and basis 61l
- Equivalent double ended guillotine e
i T lC =
?N
- N8 i F O
i d i ) i 4
O' TABLE 5.5-11 59 HAS BEEN DELETED O 5.5-53 Amend. 59 Dec. 1980
- . ._ _ _ - - . . - . . . ... . - _ _ - - _ ~
TABLE 6.1-1 LIST OF ENGINEERING SAFETY FEATURES IN CRBRP I Engineering Safety Features PSAR Section 18 l Reactor Confinement / Containment 6.2 Reactor Containment-Isolation System 6.2.4 & 7.3.1-25l Annulus Filtration System 6.2.5 g Reactor Service Building Filtration System 6.2.6
- H.abitabilIty Systems 6.3 i
Reactor Guard Vessel 5.2 l Guard Vessels of PHTS Major Components 5 .3 Residual Heat Removal System 5.6, 7.4.1 & 7.6.3 1 25 39 j. I i 1 l l i i N l 6.1-2 Amend. 61 Sept. 1981 sw -re-,,- - - - , r,..-n,-w,-,---w-,r-,,_n-,--r-,..,n,-,,,,--,,---- ,,~n,,-,,,---,..,.,-,-. ,, ---,,,,,.,m,.n_w,- e-, =,- ,-n--. ,-,--m
V Argon and nitrogen supply lines shall have two automatically initiated containment isolation valves. One valve is located inside the containment, 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 boundry outside of containment results in isolation of the line penetrating containment. Remote manual actuation is also provided. Closure of the valves for loss of supply side pressures assures that failures 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 remote manually actuated isolation valves. One valve shall be located inside while the other is outside of containment as close as practical to the containment. The valves are closed if system leakage in the equipment outside of containment is detected by cell . monitoring equipment. Manual actuation is appropriate since the ex-containment equipment associated with these lines is protected and designed to withstand the effects of postulated excontainment events without exceeding acceptable guideline exposure limits in unrestricted areas. 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 IHTS shall have at.least one isolation valve (either automatic, locked closed, or capable of remote manual operation). 29 i (V^) The isolation valve shall be located outside of containment as close to the containment as practical. Provisions for 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 included in this category are delineated in Tabie 6.2-S and identified in Table 6.2-SA. The IHTS boundary is maintained intact as a containment boundary by pro-viding in-depth protection for the piping and ccmponents. The seismic category 1 Steam Generator Guilding (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 design also incorporates-internal structures to provide missile protection for the..IHTS (from internally generated missiles), separation between heat removal paths, piping restraints and impingement shields. These design features provide assur-ance 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 Safe Shutdown Earthquake (SSE).
The integrity of the IHTS boundary will be monitored through a combination of inservice inspections, sodium to air leak detection,;and maintenance of a 10 psi (minimum) IHTS to PHTS ap in the IHX and leak detection for IHX leaks (IHTS level probes 'nd IHTS sodium radiation monitoring). These mechanisms provide assurance thtt leakage between the primary and intermediate systems, as well as leakage from the IHTS to the air will be detected early and correc-tive action initiated before significant amounts of IHTS sodium are leaked. 29 O V Amend. 31 Nov. 1976 6.2-10a
31 l The in-depth protection and Integrity monitoring provisions assure that the containment function of the IHTS will be maintained through all expected environmental conditions and external accidents. However, even if a loss of Integrity were to occur It will not result in unacceptable off-site radiolo01 cal consequences. Section 15.3.3.3.2 presents the results of an evaluation of the potential radiological consequences of a sodium-water reaction occurring with an undetected leak in the lHX. This evaluation shows that the site boundary doses are considerably less than the dose Iimits given in 10CFR20. Based on the foregoing, 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 safety of the CRBRP by reducing the decay heat removal reliability. Inclusion of Isolation valves presents additional failure modes which could cause loss of heat removal in the affected loop. 29 15l The chil led water lines penetrating containment shall have one remote manual isolation valve located outside the containment as close as practical to the containment. Since this closed system is not connected to the containment atmosphere nor does it contain radioactive mc4 rials, failures in the system outside of containment cannot result in release. The penetrations associated with the decontamination cell within containment shall be Isolated to meet the requirements af General Design Criteria 54 and 56 even though these lines are not normally connected to the containment atmosphere. Pending detailed evaluation of the accidents, this is the prudent design course. The isolation valves associated with these penetrations are designed to automatically close or be remote manually operated. 61 O Amend. 61 6.2-11 Sept. 1981
Two 100% redundant filter-fan units consisting of a heating coll, demister,
.' prefIlter benk, HEPA fIIter bank, pressure malntenance and exhaust fan, annulus recirculation fan, with associated ductwork and accessories, are provided fcr the annulus exhaust, recirculation and filtering. This insures that no single active failure will prevent 100% operation of the annulus 36 f iI tratton system.
6.2.9.4 Tests and insoections The annulut, filtration system shall be tested per the requirements of , Regulatory Guide 1.52. Containment penetrations shall be tested per Appendix J to 10CFR50 in order to verify bypass ieakage assumptions used for 25 radiological accident analyses. 6.2.6 Reactor Service Building (RSB) Filtration Svstem 61 6.2.6.1 Desian Basis The RSB filtration system is designed as an Engineered Safety Feature (ESF) to 61l filter tro RSB exhaust air in order to mitigate the consequences of the Site Sultabil Ity Source Term (SSST) event. 61l The system is designed to function continuous;y. 6.2.6.2 Svstem Design 61l The RSB is maintained at a minimum 1/4" negative water gauge pressure as described in Section 9.6.3.1.1. The RSB Filtration System is used and designed to maintain the RSB at a 61 minimum of 1/4" negative water gauge pressure and filter the RSB exhaust under all conditions except when the railroad door is open. A network of ducting is utilized in supplying and exhausting air to various floor elevations and/or celis in the RSB. This mode of operation exhausts 18,000 CFM of air through l 61 the missile protected exhaust on the Reactor Service Building (RSB). 36 l l l O Amend. 61 6.2-14b Sept. 1981
During accident conditions the RSB Filtration System will automatically shif t to (i an air recirculation mode of operation exhausting that amount of air 1700 CFM) required to maintain a minimum of 1/4" negative water gauge pressure. The filter system will be designed as a Safety Class 3 system and will meet the requirements of Regulatory Guide 1.52. The filter system will be designed to achieve a minimum of 995 particulate and 95% adsorbent ef ficiencies. 6.2.6.3 D_qsion Evaluation 61 The RSB filter system is designed to filter 18,000 CFM of air of which 1700 CFM of air is exhausted while 16,300 CFM of air is recirculated during accident conditions. The exhausted air is designed to of fset building in leakage air which maintaining 1/4" negative water gauge pressure. ; Two (2) 100% redundant filter f an units consisting of a heating cost, demister, pre-filter bank, HEPA filter bank, cleanup filter f an, with associated ductwork and accessories, are provided for the RSB exhaust, recircul ation, and f iltering. This insures that no single active f ailure will 61 prevent 100% operation of the RSB filtration system. The system ducting is designed to exhaust air from all potentially radioactive areas. Capability exists to isolate the supply and exhaust air flow to the areas where an accident has occurred and to maintain these areas at a greater negative pressure than other areas. This capability is designed to prevent the spread of airborne radioactivity ' from contaminated to clean areas within the building. 6.2.6.4 Test and insoection The RSB filtration system will be tested per the requirements of Regulatory Guide 1.52. Visual inspect!on w11I be conducted on instalIation.
,36 l
l l l l 9 6.2-14c Amend. 61 Sept. 1981
O O O C Table 6.2-5 Lines Penetrating Containment .g _ 8 %E %? 2 e a l o 82.2 be e Z B Z BE% -8 SJ
+ 2 e e Eb a se ; .> b e ye 8 e .
e i vi . 3 +88 8 % .2 w 3 '8 e*& >,3 .2 v8 83 3 bk *h .
%% 'n D W .2 % .h v 8 ., b% 2% b* o' Et i! E as 8 38, 08 '; I'G i Ra s 2%8 33 83 88 28 Penetration EN EE DE D NG .* B E ENO DEE e w" E EN SE GG EN Decontaminatior.
Waste Water Auto- Au to- Remote Return 9.2 2 Gate
- 3" CIS Closed Closed matic 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 IHTS Piping m Loop No. 2
'm 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 m
IHTS Piping Loop No. 3 . 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 ] IHTS Piping L l Loop No. 3 i 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 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 I d.g. IHTS Piping Loop No. 1 g- 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
@ 47
Table 6.2-5 LINES PENETRATING CONTAlfetENT (Cont'd.) 8
. -m IS
$ e c"
2E;8 ET 2 i i . 8,c1
% to . E t*E C*
ou "u
- 8 *sc c+o "8u 8%E 8 D8 8 A 8 uE oE G p- %22 287 ozE 'w* t: 47 Ea O"c gE 25 e3 e E2 .ct 12G eE3 Ift EE 8E "e s g Penetration *N 5E am zu SE se E
a D2 00$
<m a+a
$0 a<8 SU E s<m
*00
>ma EU m< *D m<
5 us >- Sodlum Transfer Line (In-Cont, to Ex-Cont. Stor. Tank) 9.3 1 Globe 4" N/A N/A Closed Manual Cl osed Manual N/A <30 C Sodlum Transfer Line (EYS Fil1 & Drain) 9.3 1 Globe 3" N/A N/A Closed Manual Closed Manual N/A <30 C NaK DHRS From Fall in Remote Romore Containment 9.3 1 Globe 6" N/A Place Closed Manual Cl osed Manual Manual <30 H NaK CHRS To Fall In Remote Remote Containment 9.3 1 Globe 6" N/A Place Closed Manual Closed Manual Manual <30 H or RAPS to Cold Remote Auto- Remote 61l 4 Box 9.5 2 Globe 1-1/2" CIS Closed Closed Manual Open matic Manual <10 F U CAPS Inlet l Remote Auto- Remote 611 Header 9.5 2 Globe 3" CIS Closed Cl osed Manual Open matic Manual <10 F RAPS to Recycle Ranote Auto- Remote 61l47 1 Argon Vessel 9.5 2 Globe 1-1/2" CIS Closed Closed Manual Open matic Manual <10 F
$N ue CL
~
~
~
O O O
l O O O c ! Table 6.2-5 Lines Penetrating Containment (continued) o_
- 8 %e %S s : a 6 m 8E3 he
- e 3 3 De% 4 35
% s o e #b 8 sc tub c pc 8 t.
e T wi a 3 88 8%3 32 e*8 >,3 3 a8 83
- 3 bh *h . %; 'z B -8.% % .h a8u k% 2% b^ v "-
Et ils Ea 8 38, 88; ;GY as a 3%5 53 83 88 28 Penetration E0! SE NO. 3 EM 33f EMO N#$ s m" E cb d Sid GE e$ ! Failed Fuel Auto- Remote Monitoring 9.5 2 Globe 2" CIS Closed Closed Automatic Open matic Manual <10 F i Supply l Failed Fuel 9.5 2- Globe 2" CIS Closed Closed Automatic Open Auto- Remote 50 Monitoring matic Manual <10 F Return Back
- N2 Supply Line 9.5 2 Globe 2" Pres- *** *** Auto- Open Auto- Remote
- sure matic matic Manual <10 D
! e, Back i m Pres-l h sure
- "o-4
! Emergency l Chilled Water Remote Remote Supply 9.7 1 Ball 3" ** Closed Open Manual Open Manual Manual <10 I 1 i ! Emergency Chilled Water Remote Remote i Return 9.7 1 Ball 3" ** Closed Open Manual Open Manual Manual <10 I ! c'F t 88 m .o- Floor Drain Auto- Auto- Remote
- Sump Discharge 9.15 6"
- y, 2 Gate
- CIS Closed Closed matic Open matic Manual <4 TBD ea 47 1
Table 6.2-5 LIf1ES PEfiETRATIfiG C0flTAlfNENT (Cont'd. ) g
. _. US I cE8 ET W
E 26; . a"
- tu . I t*2 7 ;*
ou u 2 8 *%c c 8 "8u 8t& 8 U8 8 a 8 ut oE G ;- a22 2 +8 2 ore '. tr er E& O"c gZ SE m3 e EE .ct 12E .E3 ISU EE 8E 8e I g Penetration m8 E?
- a. m zm
??
sm E R2
<m 828 888 25 ? %22 25 85 22 im a a*a a. < a s<a > m a. a. < w< or >-
Normal ChlIled Rmote R mote Water Supply 9.7 1 Ball 8" n* Closad Open Manual Open Manual Manual <10 I Normal Chilled Remote R mote Water Return 9.7 1 Ball 8" ** Closed Open Manual Open Manual Manual <10 I Normal Chilled Remote R m ote Water Return 9.7 1 Ball 6" ** Closed Open Manual Open Manual Manual <10 I Normal Chilled Remote Renote Water Supply 9.7 1 Ball 6" ** Closed Open Manual Open Manual Manual <10 i Ncrmal Chiiled Remote Rmote Water Return 9.7 1 Ball 6" ** Closed Open Manual Open Manual Manual <10 I cn m Normal Chilled Rmote Rmote 4 Water Supply 9.7 1 Ball 6" ** Closed Open Manual Open Manual Manual <10 1 E Normal Chiiled Rmote Rmote Water Supply 9.7 1 Ball 6" ** Closed Open Manual Open Manual Manual <10 I Normal Chi lled Rmote Rmote Water Return 9.7 1 Ball 6" ** Closed Open Mar,ua l Open Manual Manual <10 i Containment Ventilation Butter- Auto- Auto-61l47 Air Exhaust 9.6 3 fly 24" CIS Closed Closed matic Open matic Rmote <4 A $N RB
- CL, O O O
O G
. V U )
Table 6.2-5 < LINES PENETRATING CONTAltNENT (Cont'd. )
' 8
. _. Is I . c'8 ET E
! R$r .
aN W c
. E tz2 Z*
ou "u 2 8 c c+8 *8u 8%1 8 U8 g a 8 L2 oE
~
Z- o 2 287 oEE '
. UZ 47 Ee "C gr 23 .5 g 22 .ct t"E 23 23t *: 2 w8 SK RT t9 898 sbt< _" WE* It $"t $".*
1 Penetratim a. m ze se a <w so n. f" <8 "8 > w a. a. < we os 2> Containment Ventilation Butter- Auto- Auto- Remote Air Supply 9.6 3 fly 24" CIS Closed Closed matic Open matic Manual <4 A Containment i Purge Line Butter- Remote Remoto (TFEDB) 9.6 2 fly 24" name Closed Cl osed Manual Cl osed Manual Manual TBD TBD , 'I Containment Purge Line Butter- Remote Remote (TSBDB) 9.6 2 fly 24a es** Closed Closed Manual Closed Manual Manual TBD TBD 3 Ccntainment Vent Line Butter- Remote Remote (ThB00) 9.6 2 fly 24n man
- Closed Closed Manual Closed Manual Manual TBD TBD m
, m Containmed 4 Vent L:.e (TIEDB) 9.6 Cutter- Remote Remote
% 61 2 fly 24" **** Closed Closed Manual Cl osed Manual Manual TBD TBD Containment !
Vacuum Auto- Auto-Breaker 2 14" N/A Closed Closed matic Closed matic Manual <4 TBD Containment Vacuum Auto- Auto-47 Breaker 2 14= N/A Closed Closed matic Closed matic Manual <4 TBD .
$N ue I
te a,
$~ ,
Table 6.2-5 Lines Penetrating Containment (continued) 8 m -m B8 s e cy8 '2 7 N 8 0 sod m SM
% tu m % =2 T B*
b "o -
% 8 <?e cus "8o 8 % 8. 8 D8 8 86 8 ut %E & B- %e3 38B %Bt 'm DB -83 te 'C er FE; e5 a sa med 336 a s 's #3# as sa ao to Penetration
$8
- a. m 58 xa S;E se 3a US em 838 a s o_
888
- o. < v @" m o_
IU
- a. <
MU me 33 or
%S Emergency Chilled Water Remote Re. note Supply 9.7 1 Ball 2" ** Closed Open Manual Open V.anual Manual < 10 I Emergency Chilled Water Remote Remote Return 9.7 1 Ball 2" ** Closed Open Manual Open Manual Manual <10 I Emergeacy Chilled kater Remote Remote
. Return 9.7 1 Ball 2" ** Closed Open Manual Open Manual Manual <10 1 6
4 Emergency y Chilled Water Remote Remote Supply 9.7 1 Ball 2" ** Closed Open Manual Open Manual ttanual <10 I Emergency Chilled Water Remote Remote Supply 9.7 1 Ball 2" ** Closed Open Manual Open Manual Manual <10 1 Emergency Chilled Water Remote Remote Return 9.7 Ball 2" ** 1 Closed Open Manual Open Manual Manual <10 I 5N
.@" Fire Protectior
. Standpipe 2 6" b
"" H2 Sample Line Suction & 2 TBD Remote Return TBD 2 TBD 1"out TBD TBD ,
Open TBD Closed Manual ' Manual TBD iTBD 47 4 9 O O
O 0 0 4 t n e m i n . t a _ n - o _ . C i n - e r l u i a F k 0 n I 0 a 3 T eg a r _ t o _ S m i
- t. -
) o _
s r S _ u y _ o r _ H ( a _ 0 m - I 0 E i 2 r - MI P _ T g _ i n w l o l o F _ e r t u a r e p 0 0 m I 1 e T
)
e c f a r S c ue r n o a l o n e t
- Fn
( i r a _ eM n ig 0 Lnir l eu l 0 0 0 0 0 0 0 0 0 0 0 0 0 CD 8 6 4 2 2 0 1 1 . 9-
> 2
.- u. use= Q mE ! >- 6 e
r u g i F O wv. i " m Y wN il ,:1 1 ,l14 i :l41ll 1 :,j ' :4 jl,I
- llll idliI l1??1. i,i la 41
Ex-Containment In-Coatainment s e N r A N. ( s\ Q Q e >< ( >c C , >( 3 :: "3 ::: e
>< ( ><
DELETED e E E
>< ( x DE!ETED 5 r n >< ^ ,
q t { E J m ( ^ Figure 6.2-10 Containment Isolation Valve Configurations (See Table 6.2-5) 6.E- 38 h Amend. 61 Sept. 1981
i PAGE 7.5.5 Leak Detection Systems 7.5-18 7.5.5.1 Sodium to Gas Leak Detection System 7.5-18 7.5.5.1.1 Design Bases and Design Criteria for the Liquid Metal - to - Gas Leak Detection 44 Systems 7.5-18a 34l 7.5.5.1.1.1 Design Description 7.5-19 7.5.5.1.2 Design Analysis 7.5-22 7.5.5.2 Intermediate to Primary Heat Transport System Leak Detection 7.5-24 7.5.5.2.1 Design Description 7.5-24 , 7.5.5.2.2 Design Analysis 7.5-25 I30 7.5.5.3 Steam Generator Leak Detection System 7.5-25 7.5.5.3.1 Design Description 7.5-26 49 l 7.5.5.3.2 Design Analysis 7.5-27a 7.5.6 Sodium-Water Reaction Pressure Relief System (SWRPRS) Instrumentation and Control 7.5-30 7.5.6.1 Design Description 7.5-30 7.5.6.1.1 Function 7.5-30 61l 7.5.6.1.2. SWRPRS Trip Logic 7.5-30 44 7.5.6.1.3 Bypasses and Interlocks 7.5-32 i 7.5.6.1.4 Sodium Dump 7.5-32 49 7.5.6.1.5 Monitoring Instrumentation 7.5-32 7.5.6.1.7 Sodium Dump Tank Instrumentation 7.5-33 I 7.5.6.1.8 Water Dump Tank Instrumentation 7.5-33 1 1 44 7.5.6.2 _ Design Analysis 7.5-33a O Amend. 61 Sept. 1981
~ - - _ _ - _ _ - . _ _ . . _ . _ _ _ . _ . , _ , - . . _ _ _ . . . _ , - _ _ . _ _ . - . . _ . _ _ - _ _ - _
l l PAGE 7.5.7 Containment Hydrogen Monitoring 7.5-33b l 7.5.7.1 Design Description 7.5-33b 7.5.8 Containment Vessel Temperature Monitoring 7.5-33b 7.5.8.1 Design Description 7.5-33b 7.5.9 Centainment Pressure Monitoring 7.5-33b 44 7.5.9.1 Design Description 7.5-33b O O Amend. 44 April 1978
p 7.1.2.2 independence of Redundant Safety Related Systems To assure that independence of redundant safety related equipment is preserved, the following specific physical separation criteria are imposed for safety related instrumentation. , o All interract 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 for the CIS may be run in conduits containing either primary RSS wiring or conduits containing secondary RSS
,4 wiring, but never intermixed. 2 o Wiring for other safety related systems may be run in conduits containing either primary RSS wiring or conduits containing secondary RSS wiring, but never intennixed, provided that no degradation of the separation between primary and secondary 24 RSS results.
o Wiring for redundant channels shall be brought through separate containment penetrations wit.h only PPS wiring brought through these penetrations. Primary RSS wiring shall not be brought p through the same penetration as secondary RSS wiring. Wiring for the CIS and other safety related systems will be brought d through the same penetration as the RSS wiring with which it is 24 s routed. 57 o Instrumentation equipment associated with redundant channels shall . be mounted in separate racks (or completely, natallically enclosedcompartments). Only PPS channel instrumentation shall be mounted in these racks. Primary RSS equipment shall not be 1 cated in the same rack as Secondary RSS equipment. 57 o The physical separation between conduits, penetrations, or racks containing redundant instrument channels shall be specified 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 r redundant shutdown systems. 57l o The wiring from a PPS buffered output which is used for non-PPS purpose may be included in the same rack as PPS equipment. The PPS wiring shall be physically separated from the non-PPS wiring. The amount of separation shall meet the requirements of IEEE 384-1974. o Electrical power for redundant PPS equipment shall be supplied p from separate sources such that failure of a single power source LJ Amend. 57 7.1-3 Nov. 1980
does not cause f ailure of more than one redundant channel. The power sources and associated wiring shall be separated, as specified in Section 8. The criteria for cable tray fill, cable derating, cable routing in congested or hostile 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 Physical Identification of Safety Related Eautoment The Plant Protection Systm equipment will be Identified distinctively as being in the protection systs. This identification will distinguish between redundant portions of the protection system such that qualified personnel can distinguish whether the equipment is safety related and, if so, which channel. Color coding, cabinet and wire labeling and other techniques as appropriate wIil be used. 7.1.2.4 Conformance to Regt e torv Guides 1.11 " Instrument Lines Penetrating Primarv Reactor Containment" and 1,63 " Electric Penetration Assem-blies in Containment Structures for Watercooled Nuclear Power Plants" There are no Instrument lines as defined in Regulatory Guide 1.11 which p:neTrate primary reactor containment. All electric penetration casemblies in the containment vesseI wilI be designed, constructed and Instolled. In cccordance with Regulatory Guide 1.63 and IEEE Standard K/-1972. 7.1.2.5 Conformance to IEEE Standard 323-1974 "lEEE Standard for Qualifving Class IE Eculoment for Nuclear Power Generating Stations" i AlI Ciass IE equipment wIII be qualifled to con..rm the adequacy of the equipment design under normal, abnormal, and postulated accident conditions for the performance Class IE functions. This will be accomplished through a disciplined program discussed in Reference 13 of PSAR Section 1.6, "CRBRP Requirments for Environmental Qualification of Class IE Equipment." 61 7.1.2.6 Conformance to IEEE Standard 336-1971 " Installation. Insoection and Testing Reauirements for Instrumentation and Electric Eculoment During the Construction of Nuclear Power Generating Stations"
)
The Installation, inspection and testing of the Instrumentation, electrical and electronic equipment during construction will conform to the requirements of IEEE Standard 336-1971. The quality assurance prograsn for the safety related instrumentation and control equipment wilI conform to the requirements of j R:gulatory Guide 1.30. Rrfer to Chapter 17 for a description of the quality assurance program. O 7.1-4 Amend. 61 Sept. 1981
~
+ / s
( ' ' '
) of fan blade pitch and inlet louver position. The fan blades and inlet louvers are positioned by automatic controllers. Manual control of the inlet louver position and fan blade pitch is 54 provided. Manual controls are also provided for the blower motors. The outlet louver is interlocked with the inlet louver.
It opens automatically when the inlet louver actuator is energized. If a high concentration of sodium aerosol in each PACC cell is detected, redundant trip logic generates trip signals to shutdown 59 the affected PACC system for approximately 1h hours. o Pressure Controlled Bypass Valve - To prevent overheating of the Auxiliary Feedwater Pumps at reduced flow, each pump is provided with a bypass line from the discharge back to the Protected Water Storage Tank. The valve in the bypass line is normally open upon initiation during pump startup. After startup, the valve closes and then opens when pump discharge pressure rises to 1970 psig and closes when the pressure drops below 1820 psig. o Auxiliary Feedwater Isolation Valves and Pump Inlet Isolation Valves - The isolation valves in each of the supply lines to the steam drums (AFW Isolation valves) are provided to insure an uninterrupted supply of auxiliary feedwater to unaffected loops following failures in a loop which would otherwise limit ( ) the effectiveness of the auxiliary feedwater system. The
'" isolation valves at the pump inlets are provided to prevent loss of water from the Protected Water Storage Tank (PWST) in the avent of a failure between these valves and the AFW isolation valves and to allow switching suction from the PWST to the condensate storage tank.
o Superheater and Steam Drum Vent Control Valves - These valves are opened upon SGAHRS initiation and depressurize the steam drums to the valves respective setpoint levels. The superheater vent control valve setpoint is 1475 psig and the steam drum vent control valve setpoint is 1550 psig. The valves function to provide steam release during the venting period until the 54 PACC units can remove the heat load in a closed loop manner. ( ')
'~
Amend. 61 Sept. 1981 7.4-2a
7.4.1.1.3 Initiating circuits The Reactor Shutdown System (see Section 7.2) provides redundant primary and 54 48 secondary initiation signals to SGAHRS to sequentially start the three Auxillary Feedwater Pumps and the three Protected A!r Coolci Condensers when either a low steam drum level or high steam-to-feedwater flow ratio occurs in , 47 any one of the three Steam Generator System (SGS) loop subsystems. The PACCs '
,are also initiated for alI reactor scrams. However, initiation of the PACCs upon the occurrence of a reactor scram is not a safety-related function, but is provided in order to reduce the steam cooling of thc. Superheater outlet 61 tubesheet. In each subsystem, the three trip signals for low steam drum level and the three trip signals for high steam-to-feedwater flow ratio are each Isolated and input to redundant two out of three logic networks. The outputs from the redundant logic networks are each isolated within the SGAHRS divisional control system and combined in a one-of-four logic to initiate 48 SGAHRS. If two of three trip signals occur in any subsystem, the SGAHRS is initiated. The sequence of decay heat removal events is shown in Table 7.4-1.
The scheme useri for initiating the SGAHRS is shown in Figure 7.4+1. Since the automatic activation and control of auxillary feedwater flow is necessary to assure decay heat removal, provisions are included in the design to assure that the automatic initiation takes precedence. A startup signal to the feedwater pumps overrides a manual control signal. Similarly, a signal to open the isolation valves overrides a manual closure signal. 7.4.1.1.4 Bvoasses and Interlocks 1 J Bypasses are required on the steam to feedwater flow mismatch and steam drum level subsystems to allow system reset and reactor startup without initiating SGAHRS. These bypasses will be implemented as described in the Reactor 48 Shutdown System (Section 7.2). The following are interlocks provided in the SGAHRS components control circuits: (a) Each auxiliary feedwater pump (preferred) inlet valve may be closed only af ter the associated alternate inlet valve has been fully opened. The preferred inlet valve will open automatically anytime the alternate inlet valve start to close. (b) A switch is provided on the back panel to permit the operator to bypass the sodium aerosol protection circuit of the PACC. The bypassed position is annunciated on the Main Control Room panel. (c) The PACC outlet louver opens automatically whenever the inlet louver i s not f ul ly closed. When the outlet louver is fully open, the PACC 59 blower may be started either automatically or manually. Amend. 61
.4-3 Sept. 1981
\
7.4.1.1.5 Redundancy / Diversity The SGAHRS (fluid system and mechanical components) is de-signed with suitable redundancy and diversity so that it can perform its safety functions following a single failure of an active component for anticipated, unlikely and extremely unlikely plant conditions. The design of SGAHRS relating to these objectives is discussed in Section 5.6.1. Redundancy and diversity are also provided within the initiating circuitry of the SGAHRS control system. As shown in 541 Figure 7.4-1, the system is actuated on twc-nut-of-three trip signals from either low steam drum level, or high steam-to-feedwater flow ratio. 7.4.1.1.6 Actuated Devices All automatic valves and motors in the SGAHRS are provided with remote manual control capability, so that the entire system can be operated from the control room or the remote shutdown panels. 54 All isolation valves within the SGAHRS utilize an electro-hydraulic actuator. All isolation valves are designed to fail 54 to the position of greater safety upon loss of electrical power. All required components of the SGAHRS instrumentation and control system operate on a vital electrical bus. 7.4.1.1.7 Testability Instrumentation and controls for tha SGAHRS are designed and arranged to allow for complete testability during reactor power operation. Bypassing of the actuated components (i.e., isolation valves and motors) is not required during testing as operation of these components during power operation poses no penalty on plant operation. 7.4.1.1.8 Separation The SGAHRS instrumentation and control system, as part of 54l the Decay Heat Removal System, is designed to maintain required iso-lation and separation between redundant channels (see Section 7.1.2.2). 7.4-4 Amend. 54 May 1980
1 .i i
. Instrument Sensitivity e The wcstage rate sMes for jet leaks show dat leah below 10-4 lb/sec persist without major damage for more than one i
loop transit time.6 Thelooptransittimecunbgcalculated from a 13.49 x 10 lbs/hr flow rate ano 4 x 10 lbs sodium 47 inventory in the IHTS loop; the hydrogen generated from the quantity of H 2O leaked in one transit time divided by the total sodium inventory yields an increase of 6.3 ppb in the 50 471 concencration of hydrogen, thus a 6 ppb sensitivity for the hydrogen detectors. e A resolution of.3 ppb change in the hydrogen background con-Sq centration ranging from 60 - 200 ppb (i.e., a change of 3-4%) under steady-state SG operation is a' design goal for the leak detector. e 'he oxygen detector is as sensitive as the hydrogs. detector. Taking into account an oxygen background concentration of 4733 1 ppm (with 2 ppm maximum), the sensitivity is 24 ppb. 13 Instrument Range e Detection capability of leaks up to 10~I lb/sec. 47 13 Instrument Availability e Sodium loop leak detection capability provides continuous monitoring and indication of the impurity level whenever sodium and water / steam co-exist in the steam generator modules. ': 13
) Amead. 50 CJ June 1979 i 7,5-29 . . . . - . - - - . - ~ - - - . . , . - - . - , - . _ . . . - . - . . _. .. .. -
13 operation Reautreme3ts o in order to offect an orderly plant shutdown which mintmize plant unavall ebil ity, the f ollowing operator actions are required. Alarm Leak Size (Ib/sec) Ooerator Action Low < 2 x 10-5 Confirm Icak Monitor leak deta Intermediato 2 x 10-5 to 5 x 10-3 Confirm leek Initiate c terly loop Shutdown High > 5 x 10-3 Confirm leak initiate rapid module blowdown 47 Fer leakages greater than about 0.1 lb/ soc of water, the pressure buildup in the system will occur rapidly, causing the Sodium-Water Reaction Pressuro Rollef System to be activated (See Section 7.5.6). 7.5.6 Sodium-Water Reaction Pressure Rollef System (St!RPRS) Instrumentation and Controls 7.5.6.1 Deslon Descriotion 61l 7.5.6.1.1 Function The ScW om-Water reaction Pressure Rollef System (SWRPRS) Instrumentation and Control System detects the inception of a large or intermediate water to sodium leak in any of the steam generator modules (see Section 5.5.2.6) . For a large leak, three 1E pressure sensors (nino per loop) are provided immodlately downstream from each pair of rupture disks in the superheater and evaporator's (two) reaction products vent lino. The signals are transmitted to the PPS Secondary Shutdown System which inlTlates a reactor trip and PHTS an l IHTS sodium pump trip. Buf fered signals arts transmitted to the SWRPRS trip logic which Isolates the affected loop. A group alarm is transmitted to the Plant Annunciation Systen (PAS). For Intermediate leaks, three pressure sensors are provided in the IHTS sodium expansion tank equalization line to the sodium dump tank, downstream of the rupturo disks. Those signals are transmitted directly to the SWRPRS trip logic via a two-out-of-throo coincidenco logic which isolates the af fected loop. Reactor trip and trip of the PHTS and IHTS sodium pumps is initiated via the PPS Primary Shutdon System as a result of a high steam-to-feodwater flow in iho of fected leop. 7.5.6.1.2 SWRPRS Trlo Loalc Thor 9 are three separate SWRPRS trip logics, one each loop. Thus, only the 59 af focted loop wil . bo isolated leaving the other two loops for shutdown heat 7.5-30 Amend. 61 Sept. 1981
p) y removal. The SU PRS trip logic (Figure 7.5-6) and the remainder of this discussion addresses ono loop only. In parallcl with sendinr signals to the PPS for reactor and sodium pump trip for large leaks, the A S instrumentation send buffered signals to The SWRPRS trip logic. The trip circuit develops a two-out-of-three coincidence logic from each steam generator module (one superheater and two evapora m s). Each module is combined in a one-out-of-three coincidence logic which in turn is then combined in a one-out-of-two coincidence logic. Upon receiving e signal from the large leak detection circuit, the intermediate leak detection circuit or a manual trip from the control room the following simultaneous actions occur in the f aulted loop.
- a. The IHTS sodium pump pony motor of the affected loop is tripped by deenergizing the contactor coil (large leak detection circuit only).
- b. The SGS recirculating pump motor is tripped of f the line by energizing the switchgear's tripping c!rcuit.
- c. The t,ater/ steam side of each evaporator and superheater is individually isolated closure of their respective Isolation valves.
The main feedwater, aui. s lary feedwater, and steam drum inlet and drain isolation valves are closed.
- d. Water is removed from the eveporators by opening the valves between O the evaporator inlet and the water dump tank. Power reIIef valves on V the outlet line of each evaporator and the superheater are opened to provide a s aam vent to the atmosphere.
- e. Water dump and steam vent action is terminateo by closure of alI steam power relief and water dump valvos when the units have been de-pressurized to 250 psig,
- f. The water-steam side is then inerted by opening of the nitrogen purge valves which provide nitrogen to both units in the af fected loop. A regulator on the nitrogen supply maintains the pressure at 200 psig.
In the event of continued pressure buildup, the steam vent power relief valves will open at 300 psig and provide for another depressurization to 250 psig.
- g. SWRPRS piping is purged by nitrogen following bursting c5 SWRRS main rupture disks.
All isolation, dump, power .elief, and purcn valves are provided with controls and status indication in the Main Control Room to provide manual control at the plant operator's discretion. Alarms are provided in the PAS for the SGS isolation, dump and pressure relief valves to warm the operator of inadvertent
' 59 of f-normal operation.
7.5-31 Amend. 61 Sept. 1981
t Q lowers one drip pan pot into an empty, heated position of a rotary table, V directly below the maintenance port in the FHC. The rotary table has 6 59 positions, three of which are heated and three are not heated. The grapple and drip pan pot dwelI for a short time in the heated rotary table position 44 l until any frozen sodium on the grapple has melted, and the grapple fingers are able to retract. The grapple is released and raised a short distance. Next, the rotary table rotates and brings the unheated position; containing three empty drip pan pots, under the EVTM grapple. The grapple is lowered, engages an empty drip pan pot, and is holsted into the EVTM. There the cmpty pot is deposited in the drip pan assembly. The same operation is repeated twice more until the EVTM contains three empty pots, and the FhC three full pots. The EVTM then uncouples from the FHC and resumes its refueling operation. The three drip pan pots with molten sodium in the FHC are picked up by the powered manipulator, one at a time, and poured into a waste container. The FHC operators observe that each drip pan pot has only a minimum of residual sodium left before returning the pot to the unheated position in the rotary table. The drip pan pots are not decontaminated after each emptying since they are used on a repetitive basis. The container with frozen, possibly 44l contaminated sodum is later transferred out of the FHC ar.d turned Radioactive Waste System for further processing and disposal. Procedures for handling and disposing of radiontive metallic sodium are discussed in PSAR Section 11.5.3. 25 9.1.4.3.3 Safetv Evaluation The dose rate from the highest powered spent fuel assembly is Iimited to less
$9 l 44l than the limits given in Sections 12.1.1 and 12.1.2 et the surf ace of the EVTM cask body. A significant dose rate exists only dering the time when a spent 44l fuel assembly is located in the machine. Under normal conditions, this 59l 61l amounts In addition,to less than I hr.
the closest per assembly locations for a maximum where personnel can beofexposed 102 fuel to assemblies. the radiation source are 10 ft. from the cask body for normal operation and 44l 1.5 f t. for Infrequent service operations. These distances result in an 59j attenuation of personnel doses by more than a f actor of three, so that the Integrated dose to personnel is less than the maximum allowable dose. The EVTM has adequate seals to prevent excessive radioactive emissions to the operating floor of the RSB or RW. Radioactivity released from the EVTM will not exceed the limits set forth in Section 12.1 when combined with normal releases from alI other sources in the R W or RSB. The RC8 and RSB have radioactivity monitors to detect accidental releases and to sound alarms. Leakage through the seals has been evaluated in Section 15.5.2.3 for the case of 100% release to the Interior of the EVTM of all fission gas in a high 59lpweredspentfuelassemblyandthereisnohazardtothepublIc. Assessment of the physical constraints to both horizontal and vertical motion 44 of the EVTM with relation to the floor valve and closure valve 9.1-42 Amend. 61 Sept. 1981
indicates adequate assurance for both an OBE and SSE that: (a) the composite component will reseat from much greater than maximum anticipated vertical motion; (b) the clamps will prevent disengagement of the extender from the closura valve; and (c) the lip on the closure valve will limit horizontal motion to one inch. The latter is more than adequate to prevent contact with or damage to a CCP that might be in transit through the plane of the slip joint. The seals between the extender and the closure valve and between the closure valve and the floor valve ensure cover gas containment under normal 44 and seismic conditions. The EVTM cooling capacity of 20 kW is adequate to provide a sub-stantial margin above the maximum normal heat load expected, which is 15 kW The active portion of the cooling system, the blower, is capable of proviaing the specified cooling without exceeding normal temperature limits. In case 44 of failure of the blower or loss of all AC power, completely passive cooling is automatically provided by natural convection. In this case, cladding tem-perature is maintained to less than the limit for unlikely events. Some cladding failures, resulting in the release of fission gas to the interior of the EVTM, might occur; but as shown in Section 15 such a release is well well within acceptable limits. 59 The design heat removal capability of the EVTM has been experimen-tally verified in EVTM heat removal tests. These tests were planned early in the CRBRP program; their purpose and outline are described in Section 1.5.2.7 of the PSAR. The tests have been successfully performed, and the test data have been analyzed. 59l test descriptions,References 7 and 8 of Section 1.6 document test evaluations, and experimental data. Following the review of test data, the EVTM heat transfer computer model was modified to consolidate the model predictions with the experimental data. The main conclusion from these tests is that the EVTM has heat l transfer capability adequate to meet its design conditions for both forced and natural air convection modes. A summary of the tests and major findings is provided below. Full-Scale Heat Transfer Tests Full-scale tests (Reference 7 of Section 1.6) were performed in a HEDL test facility design to simulate the cooling systems of the CLEM for the FFTF and the EVTM for the CRBRP. The fuel assembly was simulated by a full scale, 217-pin, electrically heated " fuel" bundle in a hexagonal duct. The fuel assembly was contained in a sodium filled core component pot (CCP), sur-rounded by aa inert gas filled annulus, and cooled by the concentric cold wall. The test facility and test article design assured that accurate extrapolation could be applied to test data for either refueling machine. Major test results showed the following: 15 O Amend. 59 Dec. 1980 9.1-43
p the Integrated dose at the same level as the remainder of the reacter head ( (see Chapter 5.2.1.3). Activity in the reactor cover gas is contained by plug and cap seals during reactor operation and by adapter and floor valve seals during -efueling. Under all conditions, radioactive leakage and diffusion through seals are in conformance wIth the iimits Iisted in Chapter 5.2.1.3. Mechanical damage to core assemblies is prevented by control interlocks governing RGT positioning during refueling and the RGT cap locking the RGT in place during reactor operation. S.I.4.9.2 Deslan Descriotion The shield plug is so-designed as to limit the total radiation dose rate at the upper end of the RGT to less than 2.0 mr/hr at a distance of 3 feet from the closest accessible surface. Hermetic sealing is provided by both plug seals and seals in the RGT cap. A means to purge the cep-plug Interface volume before removal of the cap is also provided. Control logic interlocks prevent improper sequences of core assembly-RGT movement whenever the RGT is in use. During reactor operation, the RGT end cap locks the RGT in positlo. and prevents all tube movement. Also, no electrical power is provided to the RGT during reactor operation. M 9.1.4.9.3 Safety Evaluation The RGT, RGT plug, and RGT cap are so designed that refueling and/or operating personnel will never receive a total dose greater than 125 mrem / quarter. (Actual allowed dose and leakage levels are shown in Chapter 5.2.1.3.) Double seals and a capability of purging the cap-plug Interface volume ensure [ that gaseous radioisotope leakage from all sources to the head area will never 4g 5 98 cause a dose rate in excess of that given in Section 12. Control Interlocks are designed to prevent mechanical damage to core assemblles contained in core component pots (CCP) and reactor components by preventing the following actions: 4 1) Inadvertent attempt to insert an assembly in an occupied location. 59 20 2) Motion of the RGT with a CCP or grapple extending below the base of the RGT.
- 3) Any motion of the RGT during reactor oporation.
61 4) Positioning of the RGT over any position except one of the storage / transfer locations. 4 Amend. 61 9.1-64 Sept. 1981
59 59l 9.1.4.10 safety Asoects of Soent Fuel Storace in the Fuel Handlina Cell (FHC) The primary functions perf orred in the FHC are to: (1) Recolve Irradiated I core assemblies from the EVTM, (2) provide Interim storage for these 591 I Irraalated assemblies during transfer operations, (3) examine selected Irradiated assemblies, and (4) load irradiated core assemblies into casks f or shipment offsite. Other f unctions, also perf ormed by the FHC, are to provide service and maintenance of radioactive fuel handling equipment (e.g., grapple ropIacoment and drip pan change-out for the ex-vosseI tranfer machine). The FHC also provides contingency-storage for low-heet producing core assemblles, 44 in the event of a complete core unloading (i.e., blanket assemblies, control 53 59l passemblieswer and are and removable coolable radial by natural shield circulation in assemblies argon). that produce little decay These f unctions are impicrnented by the f ollowing features of the FHC:
- 1) Radiation shielding 591 2) Inert gas (argon) atmosphere
- 3) Vlewing capabilItles 44l 4) Remote manipulation t.nd handling of core assemblies and other components 59l 5) Cooling of spent fuel assembiles (described in Section 9.1.3)
- 6) Packaging of liquid and solid radioactive waste.
20l The FHC (locat d as shown in Figure 9.1-2 is a shielded, inerted, alpha-tight hot cell facility located between the EVTM gantry rails below the operating floor of the RSB. The cell design is based on similar f acilities used on other programs (e.g., the FFTF Inspection, examination, and maintenance (IEM) celi, and the National 59 l Reactor Test Sta tion hot f uel examination f acility (HFEF) coll) . The main equipment groups of the facility as shown in Figure 9.1-7 are (1) a spent fuel transfer station for Interim storage of up to 3 (2 during normal i operation) spent f uel assemblies, (2) a ges cooling grapple for hand!!ng bare 20I spent f uel assembiles, (3) a maintenance and service station and pit, (4) a spent fuel examination station, (5) waste container set-down space, (6) CCP storage racks (with no fuel), and (7) a spent fuel %Ipping cask loading station. In addition, provisions are made around the walls of the FHC to 53 144 score Iow-heat-producing core assembiles. O 9.1-65 Amend. 61 Sept. 1981
i ;
+.J
/ ,
SCM*a*!! 0F fftw mtamrt Me mt Ms!935 DE5Ca!PMD OF CNhfG (1) Core assemals (21 Groos t e- s t N (3) Iv*M Greenie $tsa (al Graople F*ager act#atton see (s) Grace 1e F#ager ac.nator Loas Coateel Systea Faesatic b'$1) sectical notion Orsve Plate 181 11M shtold Caiump (9) I'm Guide Sleeve fl01 fim asisco=n $1eeve itti seactee $ mall Rotating Five (f) Wertical Force applied by the Hof sc $ystem GRAPPt,t FINGIR ACTDATION IMTERtMr l'f 5CRIPit04 h Mechanical interlock consistir.g of: a weighted rod bellcrank and be11 crank vetraction spring. Interlock prevents Gr. finger retraction in Region (1) see Figure. h Redundant switches prevent grapple finger actuation in regions h and h see Figure.
@# Redundant switches prevent grapp1 4 fin been fully seated within region 31$ ger actuation
. During coreuntil rore asse41y assr41y insertionhas or witNrawal t ore drag forces ziy exceed the load setting he load con-trol sy tem . The differential sotton of drive p? ate relieves inter-loc k an automatically stors the hoist drive system. I grapple motion is termnated within region the grappl ingers can be actuated. It motion stops within region p events fiaser act2atioa.
/
If action stpps in region @Or h. iaterioc-[b
. interlock interlock (a) prevents grapple finger retraltion.
prevents finger actuation and I (v Op(RATIO 94L DFSCRIPTION gF C AS5IMBLY IMS[RTION: The IVTM is d iven downward from fuli up position. Figde A. After travelling a distance I t weighted rod of the mechanical interlock stops moving when it reaches igwel and the be11 crank spring rotates the be11 crank reljAving interlock (a) ,s Figure 8. Af ter trave.]ing an additional distance Ql) the sultChes ar(tripped relieving interlock ( tl> . see figurgC. When the Core assedly is bottored in its receptacie with14 distance Qly the grapple. stem and load control cyltaders stop sowing, but the drive plafe attached to the load control cylinder pittori, continues downward until the actuator on the piston rod moves away from switch (c) . This action automatically stops down drive motion. With all interlocks 58(1sfied the grapple fingers are retracted. The grapple finger posetton switch (d> indicates that grapple fingers have re-tracted allowing reversal of pressure'in the load control cylinders to activate the up drive system, sse Figure D. CDet aiilT WINsa.at The operationes sequence for core assembly withdrawal is essentially the reverse of the insertion description given above. The basic difference is that the pressure in the load control cylinders is reversed d placing the grapple system and core assedly upwards om drive plate about 1.5 inches. This motion activates switches Jhat allow the ist system to drive upwards. On the jray up the interlock swit (b), aid shortly after the mechaelCal inter. lock (a) 15 activated. All the intIrlocks, that prevent grapple finger re. tractioff,, remain active until t;.s interlocks are relieved on the way down as described in " core asse41y insertion." PARTI AL GRAPPLING INTf RL0rrs Redundant switches h automatically preclude hoist vertical motion, unless the switches are tripped alther in the finger entended position or retracted position. ETE : Letters in diamonds are also correlated to those in Ff g2re 15.5.2.1.1-1. /'l FIGURE 9.1-168 - SCHEMATIC 0F IVTM INTERLOCKS ( ,/ (Sheet 2 of 2) OVER CORE POSITIONS' Amend. 61 9.1-92 Sept. 1981 e
i, 1 SCH[MAf!C 0F IVtM int (#10Cks OVf R INVE55[L 570 pact P0517!0NS CPIRAf!ONAL AC5CRIPi!ON the 1Re<(b) switches and . SRP at 00 activates redundant interlock switches h and deactivates Interlock I C_CD_[ A.._nt"l_lL.Y,__l%. I..D..I_!Gh : l The IVTM is drJ ven downward f rom f ull up fo41 tion. Figure E. Af ter travellinq { a distanta .[lVI t e weighted rod of the mechantral interlock stops moving when itInterloc6 reaches jet (5 and the bellcrank spring rotates the beIICrank retttving i,a) . see figure F. Af ter tra ing an additional distance (V) the swit(hes ad tripped ret teving interloth . seg igure F. When the Core asse%ly is bottop.ed in the CCP within _ance the grapple, stem and load control cylinders stop eoving, but the d we plate $ttached to the load control tyltader pistons, cent)nues downward until the actuatGr on th6 piston rod moves away frofp sultCb(C) . Ihts action automatt(ally stops down j drive motion. With all interlocht %4tisfied the grapple fingers are retracted. 4 the grapple finger position switch so) 6ndicates that grapple fingers have re- ' tracted allowing reversal of pressurf in the load control Cylinders to activate the up drive system see Figure 't. C0ef a$it"Og v;? wean nag The operational sequente for core assepely withdrawal ts essentially the veverse of the insertion description given above. The bastC dif ferenCe is that the pressure in the load control cy* tnders is reversed splacing the grapple system and c're assertly upward'i from drive plate about 1.5 inches. j This motion activates sw+tches (gc) that allow the h st system to delve upwards. q On the.way up the interto(k swittn (e) and shortly af te
- the pechanical inter-i d lock < a
- ts activated. All the interlocks, that prevens grapple finger re.
tracti6n. remain active until the interlocks are reifeved on the way down as ] destribed in " tore asserely insertion." i j NE: Letters in diacends are also correlated to those in Figure 15 $.2.1.1-1. 1 I FIGURE 9.1-16C - SCHE!1ATIC 0F IVTil INTERLOCKS (Sheet 2 of 2) OVER IN-VESSEL STORAGE POSITIONS 2 1-l i 4 6 9.1-94 Amend. 61 Sept. 1981
( -HANDONG Ball
} ,
N ! ' 5 IN GRAPPLE DRIVE HOUSING g.,.NSERVICE PLATFORM sa. h
/ s CHAIN STORAGE TUBE 3d.I;
__ j
) a 7 % CASK BODY Nd
'~ 1 h N ~' J ;
//g N PARKING STRUCTURE a
W WA. ,1 S _.A'%
,i.
~:. l
,eEQUIPMENT CABINET gygv Q
--CASK CONTAINMENT BARREL CLOSURE VALVE - DRIP PAN p,' s.. s, LOWER EQUlPMENT PLATFORM
=
ACCESS LADDER EXTENDER ' % j'- },. . Figure 9.1-17 Auxiliary Handling Machine 9.1-95 Amend. 44 April 1978 b
The RSB argon supply is reduced in pressure in three stages to satisfy the Q interface requirements at the ex-vessel storage tank (EYST) and at the fuel handling cell (FHC). Other reactor ref ueling system compononts serviced in this area are the RSB plug storage facility, RSB floor service stations, EVST 59 seals, FHC operating gallery, and FHC conditioning loop filters and blowers. 9.5.1.2.6 Fresh Argon Sucolv at the Steam Generator Building (SGB) Argon for the Steam Generator Building (SGB) is stored as liquid in two Dewars located on the SGB pad. These Dewars have a capacity of 1500 gal. each and are equipped with fill and vent lines. Normally only one Dewar is in operation. When it is nearly empty, a low-level Instrumentation signal operates automatic controls to shut off that Jewar and to open a full Dewar to the supply header. When the switchover takes place, an alarm signals the operator who is then required to initiate action to fill the nearly empty 59 Dewar. A control override allows drawing on both Dewars simultaneously. Two ambient-air vaporizers on each Dewar can evaporate the iIquid argon at a nominal maximum gas flow rate of 250 scfm each, at 200 psig. With adequately sized piping and regulation, approximately 500 scfm of argon gas, at 93 psig 59 can be delivered to an Intermediate loop expansion tank. The argen flow from these Dewars passes through a filter and into a main header. Branch lines serve the sodium receiving station and the incoming sodium drum sodium sampling oackages. n 9.5.1.2.7 Fresh Argon: SGB Distribution ( l The argon flow from the main header leaving the SGB dewars is divided into several branches and routed toward the three IHTS loops in the SGB. Each loop 59 supply services the fcilowing components: line vents (freeze vents), rupture disc spaces, intermediate sodium characterization packages, intermediate sodium pump seal purge and oil gravity tank, sodium dump tank, and the pressure equalization line between the Intermediate sodium pump and 61l 59l Intermediate sodium expansion tank, providing cover gas for both. Non-radioactive purged gas from the sodium pump oil leakage collection tank and oil gravity tank passes through an oil vapor trap before release to the atmosphere outside of the SGB. 9.5.1.2.8 Vacuum Services The argon distribution subsystem incorporates permanently installed vacuum pumps. Several locations are provided for movable pumps that may be 59l temporariiy conneeted. 59l 9.5.1.2.9 Atmosobere Purification Unit 59l The atmosphere purification unit continuously processes a side-stream of argon gas drawn from and returned to the FHC gas cooling stream. The unit contains two parallel gas-treating trains, each basically consisting of a copper bed to 59l48removeoxygenandamolecularsievedryer. d 9.5-5 Amend. 61 Sept. 1981
WhlIe one of the Ioops of thIs unit operates, the other Icop is regenerated by 59 l fl wing mixed argon-5% hydrogen gas through the molecular sieve dryer and then through the copper bed to reduce copper oxide. The water produced by this 59 l purge and reaction is ranoved by the unit vacuum pump to CAPS. 9.5.2 Nitrogen Distribution System 9.5.2.1 Deslan Basis Nitrogen is to be supplied for (1) cooling and inerting the atmospheres of the colis and pipeways containing radioactive sodium and 1he Control Rod Drive Mechanism, (2) actuating pneumatically-operated valves in the Inerted cells, (3) cover gas for the Dowtherm tanks in the chilled water system, (4) purging the IHTS steam generators and evaporators in the event of a udium-water reaction, (5) primary Na removal and autoclave operations, (6) purging of the RAPS and CAPS cold boxes, (7) a cover gas for the Sodium Water Reaction Pressure Reilef System (SWRPRS), and (8) misee1Ianeous handling and 59 maintenance services. The. SGB nitrogen supply for the sodium-water reaction purge is sized to 59 ] prcvide 250 scfm of nitrogen for a maximum of 12 hours. 59 1 The SGB nitrogen supply rate to be available for the RC8 and RSB cell purge requirements is to be 250,000 scfd. To meet these Iimits the nitrogen subsystem contains two sampling and analysis units, one for the RSB and the other for the RC8 which periodically samples the gas in each nitrogen-Inerted cell and analyzes its atmospheres for radioactivity, oxygen, and water vapor content. The cell is purged automatically by fresh nitrogen whenever the oxygen level exceed 2% or the water vapor concentration exceed 1000 vppm (one of these, by operator selection) as monitored by the respective sampling and analysis unit. If, as the result of purging to reduce the water vapor level, the oxygen concentration f alls below 0.5%, dry oxygen from a gas supply bottle will be 59 Introduced manually into the af fected cell at a tap provided for this purpose. The RSB sampiIng unit causes the celi exhaust gases to be diverted to CAPS If they are radioactive, or to be diverted to heating and ventilating if they are n t radioactive. All RCB inerted cells are normally exhausted to CAPS, and an 59 55 alarm is sounded wnen a high amount of radioactivity is detected. The oxygen content of a nitrogen inerted celi is to be Iimited to 0.5 to 2.0%, and the water vapor concentration to less than 1000 vppm. The oxygen limits are chorma to provide enough oxygen to prevent nitriding of the steel, and yet not exceed a fire-limiting concentration of oxygen. The water vapor is 48 Iimited in order to assure early detection in the event of a smalI sodium 59 leak. O 9.5-6
%end. 59 Dec. 1980
about.10,000 vppm. Reduction from this value to the 1000 vppm limit will be E done by purging with nitrogen. Nitrogen for service maintenance operations is available at service stations located within the R G. 9.5.2.2.3 Nitroaan: RCB Auxillarv Sunniv An auxillary supply of nitrogen gas is stored in high-pressure standard cylinders located within a cell in the tornado-hardened RW. This nitrogen is used to ensure the uninterrupted operability of cor.tain essential valves.In the event of pressure loss in the nitrogen supply header. A control valve automatically restores pressure in the valve actuation circuit when an abnormal decrease.In operating pressure is sensed. A check valve then isolates the valve circuit from the main supply line in order to preclude auxiliary supply blowdown to the remainder of the failed supply circuit. 9.5.2.2.4 Nitroaan: RSB Distribution f The 150 psig RSB header, after providing a side stream.for inerted cell valve 61l 59 operators, branches off into several lower pressure headers that service the needs of other systems as welI as those of the RAPS and CAPS subsystems within the RSB. RSB cells and pipeways containing sodium components are inerted with nitrogen , during normal operation. The cell pressures are maintained by a feed and
- bleed arrangement, and a purge function controls impurity levels. (See O Section 9.5.2.2.2)
The RAPS and CAPS cold boxes are inerted with nitrogen at a continuous low flow rate during operation. These flows are. vented directly to the respective cells so that the cell atmospheres become nitrogen-rich. The RAPS cell pressure is maintained by a back-pressure regulator that bleeds the cell atmosphere to CAPS. The CAPS cold box cell atmosphere is vented to the 50 Heating, Ventilating and Air Conditioning System. The nitrogen requirement to the cold boxes serves two purposes: to inert the cold boxes so that water condensation within the cryogenically-cooled structure is prevented and to provide gas _ for valve operation. The cold boxes would not be. effected adversely by high purge flows nor would there be an impact on the CAPS decontamination process. The only consequence of such flows would be increased nitrogen utilization. Nitrogen for service maintenance operations is available at service stations located within the RSB.. A controlled pressure N2 supply is provided. j 59 separately to the autociave.
- Nitrogen gas is provided as a cover gas for the Dowtherm tanks used in the j- 48 chilled water system.
i
- O 9.5-8 Amend. 61 i
Sept. 1981
9.5.2.2.5 Nitrocen: RSB Auxillarv Suoolv An auxillary supply of nitrogen gas, stored in high-pressure standard cylinders located within a cell in the tornado-hardened RSB, O O 9.5-8a Amend. 59 Dec. 1980 l _ _ _ _ . -
1 P 10 Atuos CAS ritL
%NC ,,t,ER l
~
- =c NC h X W
, .O U
0-Y k;j,
. . O.
<p iI I il l I , ,
J f', l ! ll l l l vAeoainRs S LO. I L O.
?O ATMOS h._.10 to ATwos A1 0.. ,
1:-.10A10..
, NC NC i 5
, l i e i I 1 I I I l 4 e 1 8 8 9 1 1 E 8 M M cM ' CM NC NO w TO A1MOS.
l XNC )(NC FURNISHED BY SUPPLIER l ll "' : y y - gig _, LIOutD ARuON STORAGE VESSFLS
. _ _ _ _ _ . . , . - - , , - -, . ,_2 . . . - _ _ _ . - - - _ - . - - - _ . , - . . - _ - . _ ~ _ . , . . . . . - , _
S
\
NONF e i l NONE+ - I j i jFIG. 0 61 Sh 5 R EGE N E R AT 60N tsOUdDARGONSTORAGE l P IG 9 61 Sh 2
'" E ic 3 sH , j
- ) i NONE e-1 l
l l NONE e-- F . O. NC CAPS ,
- M m =; :% )
NC F . C. FIG. Il 36 $H 1
< > X NC
^
i x NC NC VENT TO CELL M T P N TEST NONE p VENT TO CELL SC 3 4*N
~~
VENT TO CELL NON VENT TO CELL W3 P9 E9 ( $. $: F O. F0
+
+ +
/% /%
,r- -
T, , ,
-T .
( f
'w . .
l L .. t > E N CONT AINME NT E x CONTAINMENT PRIMARY SODIUM PRtMARY SODIUM STOR AGE VEME L STORAGE VEMEL Figure 9.5-1 Argon Distribution In The Reactor Service Building l (Sheet 1 of 5) 9.5-19 Amend. 61 } Sept. 1981 fs
I p._. ~ 1 i ..a3. s s a aa<a t . m _ s ett M ..c e s t p3 g d ) L_P FeG t t 5se 2 w 1 ..
. . . . .. > P".r. ^'t) g .g g ,
q.. q..x- q..
. m
_l A - n.-__ a - - _- _c A. -m- A. .-- --... -l -- - - k - - -- - - f. ! -- . .. ,. . . n n .c ,
.I ,
A l
^ :. 3-.:
- _Jm
1cJ " " " " " " l 1 l x- l , c'e s-me .J . . .a y y y y y. y Mc~.;*...,.... j cy rgg.g,g., l
; c;__m l
,-d.< ,2-I
.,, 13, 9 ,,,, ,, c,
'?u , , , ,
7c -
. _ . _ . . e. .
7.,.,y,,, u. m i 7,;; g , f, . ,n,; , ,.,.,,,,,,,,, m. ,, ,
, .3 c. _
=' ,
{
. s a
?- - .a*_
4 E*2 -
/ T' _n_1. __.e__
. _ {. _ _
n . ....n, ,1' c, , g, 9
/, . ...-1
,e 'C '
^ A vg g----y gl v Y T T.sa T I 1. . ,
V. .
.c. . . ___w
*w,
' j Y
; % s,*,,0g. g,- < ~.co.t.o. g,.
gig . s_._ _ s v .= = s a ..ss. <
.sh7 :-:- mau.(-;
*C vac",ves, *t t *C ]
8.O E W
#Ap5 2 --:
I l l I l I I j 1 6 i 1 1
I i i l i .
) '
, ,, _ ..... . \
. . . . . . . . ,t , 1 i
4 l' p', a X- a X= p a X- g.la X<
- . : -w.---
..\ ,
9 9 o
- l y -*x? ; 7.Fx. l W -; '
, - xc
. , ]I ve47 70 Catt , 7 vigt TO sa t ' '
',-[ v847 TO CSLL
v f'
F - i -> * !E' " '7
~~
' I ~~ +z?iE' ~~ r - * !E~ d.and.9..
, -y - . f-a p-l n a. n -> . n b l i l __q,, p..,- - y s-
' ^
Y Y '* *~v-*:l* y-cu__L-);ur.s. .. L__m:::.%.. _s ,. ,. .. e t ):f ::::.o
. .oo. ,
*c -
car. g- e . . . .. .. > l 4 I _ n R l l t I Figure 9.5-1 Argon Distribution In The Reactor Service Building (Sheet 2 of 5) i t 9.5-20 ' Amend. 61 ! l- Sept. 1981 i I ,
.\ li -e--* *-v-v =c-en a--------,-.-=v-.-,,,,-m--- - . . - , - - --- ,= - - - -- -- - - - - - - - - - -----v- - *
- I l
I v F0 F 0-ND*et x l w LC 4 g P'G 9 S t D* 2 FO l 1 : X < FIG S S t sse 2 NC - - FCV
> , 7 O. esc F0
[ *sC v ,
.m
, ,, ,,,C h[9'1
- -?
~<
F< c _, _ j ej ( ' FI
]- . - -
g3
;
- sc 3 moss T sc 3 mosse
.d ty .
3
- + +
0 ' C c J..
.. F,s , .. ..
L er w .: : h,- t __.. ,
" **c .t ,p f7 E FO t. L 48 2 8 N: b+: tap>
i FO F0 ' -I
. EC2 SC3 U
IR5S 5mc* FLUGII sca .p. sca L - . . . 2 a y t .. _
+
> evs' ' l l - .c i
%._ j p .c:
E N C-
- i l
I i h 1
\
0 1 l l i [Nc
- 1 :.: : i
-i g l .......
x 4 1-FO
^
' ' NC 8 NC j
' O'I ' T_' T ' N -f * ... . y' ' t o
. .~ -- 4
**< l
' * ' ' ' * " ' e. c f
FFc"] ' g l I i L_.1_ i L - __ _ _
-r-- -
c='_'a j o ....in .<u,,,,p. ___ w = F'Y*%J9' JN' RLau r i ; : i [_..f s,w.c - .7su-- F l ro .~ J
...u.
i -
--c >4 g' e j ec ,, c' -+..e, t _'%,y..
~*
j .l:~ p , ,
-J_- - ';'l A i ~., c .c .
l.<=>.{<x->- 2-f ,
- x. X 1 l
i. 4 j Figure 9.5-1 Argon Distribution In The Reactor Service Building i (Sheet 3 of 5) J l 9.5-21 Amend. 61 Sept. 1981 , i l I I o
W t DNONE NONE O-FIG.9 42. SH 3
)'
l ' acO~ l FIG. C.St. SH 2 F.O. X NC e 1 m Y, i e i
4
=
I
\
0 NONEO 1 RAPS PURGE 1
' ' N Y FIG 11.3-5. SH 2 X
F.O. NC Y d b II
==
T FISSION GAS l lg MONIT MODULE (, q p.y A h b M l FC 2 l GAS SAMPLING TRAP l FCl pyg, g g,3 4 3g 4 ,e
, n , N C.
. 1 .
e LJ LJ NONE $
,.4 SC 2 *--* SC 3
/\
SC-2 + T
$ ^
E FC 1 FC NONE ~ FIG. 952 SH 2 SC3 ++ SC-2 gp3 , x : FIG.18.3 6 SH 1 Figure 9.5-1 Argon Distribution In The Reactor Service Building (Sheet 4 of 5) 9.5-22 I Amend. 61 / Sept. 1981 '
f F IG. 9. 51 ( t 8 -- - S.
' NONE -
(o) LO ti
?C
, -g g- . - -
FIG. 9 42 SH 1
-< l ARGON b 1
NONE: F.O. F .C. 3NC X "{ ri g , R EG EN ER ATION _a , , GAS MIM ER pr N C. NC 4 4 - - 4 8'NC NC h FC F .O. FC JL 9F A6 JL 9F
/: ; .
F.O. ., t 1 9 NONE I AL S(RE,g fSh A) REGE NYR AflON fe*FJ " '" V V SS Af. 1 l I' 1
e I I a NC FO FO FIG. 9 52 $H 3 I
.= ..
I
<> ~o~E H j NC gNC 4 @ 8 =- :
ANC U h NONE NC l CAPS I M I FIG.113-6
'F s l4C E
1P JL d' m FC l NC l 4 _
'NC NC f FC Jg 1P J6 9F
~, # ~,N l l l l 5 S 8) 5sA VALVE POSITIONS ARE $HOWN TO DEPICT SYSTEM "A"IN THE PURIFICATION MODE ANO SYSTEM ~B"IN THE REGENERATION MODE tion PURIFICATION UNIT I Figure 9.5-1 Argon Distribution In The Reactor Service Building l
(Sheet 5 of 5) 9.5-23 Amend. 61 Sept. 1981 l I o l
/
l l l
.f NONE e ORGON NONE w l X 1 I FIG 9 52 SH 4 l }
NONE E'O II'3'4 FIG 11.36 SH 1 SC =* SC3 y( SC2 NT CELL
* - NC arf
+ NONE 3 p l f SC3 FIG. 5
, SC3 SCr3 %N FIG.11.14 _ i
.s ,, _#. .
C 1 P J L m A RECYCLE ARGON RECYCLE ARGON l STORAGE VESSEL STORAGE VESSE {~'
- NONQ FIG. 9 52 SH 2 M I SC-3 l ga[i ~
; _tr f-3 NONE z F.O. > z g F .O.
!j %j ?.
sv a X SC-3 Nc d' l t__ NORi " NONE V AC. PUMP CAPm A . , y SECONDARY CRDM
,e M ". m . . ,N
), ARGON SUPPLY FIG.11.3 6 l 9
9 I i iw FIG. 9 S2 SH 2
/ 1 ,
NONE w #RtPUMPS FO FO FC g"*
* =w I F4G. 9 52 SH 3 FO FO W 4 1 F SC2 F.O.
NONE SC 2 PE NE TR ATION 10 Ov E aF'Ow vf SSE' T .. b . . __T t HEAD ACCESS , AREA , , ,,gg ' 2
, . I.
s/ \r \/ \/ RE ACTOR I
-- . - PR E SSURIZATION
_q F-- RE ACTOR VESSEL -d f PRESSURE EQUALIZATION
" Figure 9.5-2 Argon Distribution In The Reactor Containment Building
. .M (Sheet 1 of 5) 9.5-24 ?
Amend. 61 Sept. 1981 )
(" l NoNe .
- sc2 !" "'"'
( NONE FIG 9 5 2 SH 1 [ l MuN - s"8 4 NC
'* 4 90 NC gg, NC NC -
7 s FIG 9 51 SH 4 l G-4 NC 0-X O NC NONE &4 SC2 CONDENSER NT CELL N C. 4 } -. CELL TEST T AP LO
, F .O F IG. 9 5
-2 SH 1 s t
% )
- SC-2 PRESS EQUAL. ' FIG 9 52 SH 1 PRI. PUMPS
- ), '
SC-2 h - - - - - - - - - - - Y - r PRIMARY ( OV E RF L[I l l
?
l l . l t - . . - , - . . , . . - . . - - . - - , - - - - . - - . - - - , - . - - . . . . . . - . - . - - - - ~ - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - ~ - - - ~ - - -
1 9
\
n i l l M ac FC l r.ars 1 x w
&O
- A '
, ,q, , ,,34 l
3 } --$ b s.wIwT% FC FC sC:+- NC (} rs VAFOR NONE e- 4 SC7 CONDE NSF R VENT
-H-Ett D ~c hC5"<'
TEST ( IAP l l X tO
; ,..O.
SC 2 e---e NONE NONE h c"5
- m . . .
'NC N NC VACUUM PUMP F"4G. I1.3 6 $H 1
. . . . T. fvido - _. . . - J l l Figure 9.5-2 Argon Distribution In The Reactor Containment Building (Sheet 2 of 5) l 9.5-25 Amend. 61 Sept. 1981 l . 1 l l t 1 a
?
i i SC2 -.-e esoht F RESM .RGOes F0 RECvCLE Ascom l FfG g 52, gn , A FeG. S E2 5M S ,1, F .O [ 70 I M. e - FC FC g
. .ec t y c '
- , A SCE SC3 i
~ t_
' (Paeasamv etMP] Ost *uPPLY p.._ ._ _ p ia 8 v v
--%. s"a'laYe .!q '
- e: :
p1=.1; sa"" j _ c x y s L.._ I ec FC 7 esc g.. , y y 7.. p
;x . ..
j soo uns. ,l lj.
. a-
.. c.s.su..
sou uzanom A i
, o,o n., asse we t y L_c l l e4e eee i .-'
.'%. )
(
I I i
-Q po esc ,o p =C
?
mvec l o y 1 l esoais mo=e
,,o.,,
l c ac __ X K _7,_ ! dI - _ _ 37 _ tt _ r7 tt ,c ,, d d .;, l og , g___.. __.. _g __.. _;. __ _) s ___ _g
, s s s s_ s d r 1.- 7 r-'1' r~"1_'l
.T_.~
r]
)
bh5I, ",, 2;*"'a [~f_~] l Nd!YNor "d.5'" [di5] r, $EffN" : i 1 78T ~~ - > .,,,..w, em s 1 L y ( , x* en si te sw i i p 3 . ... , sc , *- l Figure 9.5-2 Argon Distribution In The Reactor Containe.ent Building (Sheet 3 of 5) l 9.5-26 ( Amend. 61 l Sept. 1981 i t i I { t I l
i l il l t l I i i P
' RfCyrtt wer ,q555 L "I sC 2 69 eeoast 3
@,, 3 X @ F0 kmC . .u o , ;
. - *E 9 -
' es0ess
---(=>-* e e ,
LO !FC 7C LO _*1 esc
- -: f W 4 ,
8
- Test test u.
VE8sh VE NT WENT hoseE k%C 1 x
. ?rsY
- J OI r T Ap MOast i
.I..-
- h pg m
- pg_ UE' y y .o e
. K3 g3 l
**Aav %
~-
innaev ese 3,: -s S X l . x e ( v
~ r r t
V AChute Puesp t t u y WWv%C r i r . A., sno,. stoaAc.a vessel 3 v4Coues puw 't ,1 w:- $ w-C m v h
I 1 I 90 t i a Xc 5 ,,-
<l F0 k 4C tetAO ACCESS A#f A 'l' '
+ >,
. . . m, : ,
w& S J'49f.*"<t FUELassGt a%. d-- 2 5 Y lM.c ! l 1 x=r : x wc o
~~' -..L 2 6 xnc : x w
k' ,0 , X= X= l {, S s -- c"' y >; Figure 9.5-2 Argon Distribution In The Reactor Containment Building (Sheet 4 of 5) 9.5-27 Amend. 61 Sept. 1981 t l
i k l l W NONE I scom ieG 9 5 2 SM 4 _ i
'. i i
< b 4
[NC lb P O. fNC P.O. hl l- l 4 n > 1 NONE ** SC3 NONE SC3 YEST NONE LELL VE Y d'
*TO I
__, _ _ _ , , _ - _, ,__ , , rett y,,, <
,, INC___ NC R NC a l," gg TO CE LL 1 ,l SC-3 i N. N., N ~
h b b b h TEST T AP CELL i -
%/ %/ %/ %f jg f PRsMARY h COLD ! PRIMARY No COLD l R AP LINE VENT pm,w g TR AP LINE V ENTS 3>,e X C T.
- J MAKf UP PUMP DR AIN VESSEL
, 1 q r P NONE t _ _ _
k, '$NC NC t -- T* ME NC___ M Y NC_ I _ _T{NC__ 4 A A A A A I
%# %# %# %f %f N/ W NC NC l ] l l. l l L.__. _.J L i i _J L _J OVERFLOW RE ACTOR PM' MARY h MAK E UP P8ttMARY No MAKE UP tee AT M AK E-UP PUMP LINE VENTS PUMP Lih! VE NTS EnCHANGER LINE VENT VENT II
.l
..-,--,--.a--....,..-,..w...,_,-,.-..,___ _ ,. .- ,,,, ..,,-,,-n. +- ,-n, ,,,.,-.n,_.,n,,,,-,,---,,,-.nnn.,,--,,.,_.n.,,..,,.,,_..,. ,. . , , . , --
s J t t I J o.-
- l
, , , , , , >, , =' ', = r'
), ..a . u - >
l , , y l l i . , gg [M F 0. kNC F.a ) t . ( ""' o~
- 4. sc2 m__ , , uo . i
=c
. 1 'c =c
--"$l 5%uM TO CELL 8
, , e
,^m
- h fc ,
A*c b , g-A 4 n A t H- Y, - n 4 r T_ _ _T, i , j j
,ky "^P , NvIMs7ariom l
t ho=E A~c e{s)
't i "Es'T"o'vYs'sIL m
a, = , , , , 4 , l Y
- -7 L_ _. , ,
?!g g ,"'^' '
vacuuu ru- , , , , vi T
.g CAPS l
)
FIG 113.g $H g Figure 9.5-2 Argon Distribution In The Reactor Contair. ment Building ; (Sheet 5 of 5) 9.5-28 L (Next Page is 9.5-30) Amend. 61 (' Sept. 1981 $
/s !
l 0
,c w
Xc s , < , iGAS PILL k r.o. k , O. \ T l l l t i I I I I I I I i Jl i I t vAPORt2Eh5 I TOATasOS - TO ATMOS TO ATasCS 70 ATasOS NC NC e I l pumpeSMED
- TO ATasos. Av avPPLIER g : TO ATasOS
- 4y-
,J '
_. .,0,,9 a. p,c
%.c ,=c aOas -
l _ . ..-.. . . _ . vtNT LeOUeO AAGOes STOR AGE VESSE LS 1
4 i i
^ ~
~4 b peg g 63.3M 3 uso AnGo.s ston.GE NC M ),
" F'G S S 9. SM t i Fa NC yn l .y l * * ' ~ ' '
t , , , M 3 4 4 A 4 eia . .. .
- o. xuc. h E M >
M ) FIG. 9 63 SM 2 I l m ). l 1 3_ lF= .1._.l l ih. e."<. x n'*""
.r..=,.
t.. x ._ a 1 s- s, X" X" Xc Xc Xc x x e e , ,
- s s s s 6
o. I T T v.Cuuuev.e P). . .
" C i,'hYiN j a"E' o iI n0s"Ynost.rion IEs'iv5.*[s"T.T om l
t l L Figure 9.5-3. Argon Distribution in the SGB (Sheet 1 of 2) 9.5-30 Amend. 61 l Sept. 1981 U I o
1 i i
..es ..m I 1 j eG 9 5 3 g 5 1. 2.
8 i i H =o <
} . c lw c. j 'O 4 4
...i, I
o,!
' moms
' men. <
=c * " '
.sch. 1
~
, 4e ~.v A gl 1
J'- A f .L
..a g
*.A g
=
- o. sd t => a g T'
- r;
'$ 'h i 5 gi xA <_ _ _ . . - .
5T lE
, .. .rA ' . r .
A- y .=y m. i e.L.s . t 4 A A '---> U_w 2 J I.. ..: :...,.. w~a v v v <' . p l.,_....___. . .. .. <_- .. .. ... . . , _ _ . v. , y v...v.
,.to....
y
,4,.._ a &5 . m
&^ $<3
}W $ $
g .h * - a WW E E
=
I aY lY
- XY Y ,r, ,,g,; ,, y,,g,q r,d;,-]
.. ., . i =,;. .
' p*::: = . , '=,- p!
da" oaf ~j u uu 1. 8
'a.".'_'
%< 85 C.
.e l
-.W ec l
l l l t 1 i I i l
I I a
) %'.777,'
N,
-<-qF>-.c<
**" ~= *
, >X.. 3 &.a,, 4%
t ".. ;
.4
" % 9 ..y..
c ., 5 y J d.ag a k'*
%] f ..,, r,,,,,,,,,
i
, ;)
_, [g_,1
,3 r ] - ! *= l :
i g' l l E. l
=>_
{--- ;- - - i i !. f ; c H-. <
- : ;-=:,.
# . - . . , ; -t- -+-
- P, a. , [l AT:: 'J,',sl~
- x P..J. . :
1
~ ~ . ~ ~ - 1 k : "'
!;=:,. '
g m... L. . , gs .
/ _
x.. .. A T ' Xc m
+
~
1 y Y
!?-G
(=,77 ) i. l ! b" . . Figure 9.5-3.(Cont.) (Sheet 2 of 2) Argon Distribution in the SGB (Typical of Three IHTS Loops) 9.5-31 Amend. 61 3 Sept. 1981 {
a , 's
- .:. E .,
FO w NC T7 g NC
.O9 J >
I TO ATM - - ~ Tl f LO y
, NC TO ATM NC HLL 3 X I 8
NC NITROGEN SUPPLY SYSTEM
-% NONE 4
RCB SUPPLY FIG. 9.58 l NONE: , RSB SUPPLY FIG. 9.55 SH t 1 {, FO NC F ip i NONE ( I
, VALVES g.j ty[N (qq I
) NS DE RAPS ----> VE NT TO CE LLS o, FC i FIG.11.3-5, SH 2 NONE I i EVS PTl 4e'
' ' Y' - CONTROL ----> V ENT TO CE LL ,
PANEL
} ,
y, .4 t l VALVES FC g-~,a I I i INSIDE CAPS 4 VENT TO CELL COLD BOX FIG.11.3-7 a i
+ -.
( ] VAPORf2ERS TO VALVE OPERATORS AUX. LIO. MTL SYS. ATM I
%l
. . - TO ATM p,.[ _. ( , l
' {*%; to FO l ) l
! I j /\ e i
4 , M - s FO l n f n i W # i n_ _ A ' 1 l e ' NC , .. .. l M ! I n.... - ---- - - - - - i
- N M .
NONE : i AUXILIARY N2 GAS SUPPLY Figure 9.5-4. N Supply to RSB and RCB 2 t Amend. 61 8 9.5- 32 Sept. 1981 .I
?
i i i I ;
; F.o. S u NE FO FO (L x NS
. . _hmt. 4 xN< xN< . _ xm_<
L _ ,_ A.. F1 F1 rg x
\/
Y, Y Y La N C-
,N C. [NC SERvrCE St ATIONS IN R$8 SERVICE STATIONS #4 DECON F AC- FO Ft F1 a
l CA*S I F IG.11.34 NONE h l l l l i
\
e I I J l y g .
.,q,
,o ,o j . . . . . .
x
= c.
.C .
- .: _.A .
NO g' C p eg g ug g,, , a
; y son C tt 4.c .
g___.L. 3 3 p
.g. .1 90 00 I
_y. _ _. _ 9
- SON
! I C.,5 CoLO,URG.
C.LL G883 r------ e , w,, y
. . . . . . .u, .
i;;Y'3",','O. O
- . E.
=a no a o.
3 l l t W - o.- u M . . . .. )..
-. 4 Figure 9.5-5. U2 Distribution of RSB (Sheet 1 of 3) 9.5- 33 Amend. 61 Sept. 1981 l
, / I 1
F l F W IIOlet t I
'**** . e s s s,. s I
feO4E
.c .c g,c apossE ascess A__________________ _ _ _ ____ _ _ _ _ _ _ _ __ A__t A
4,,,.Aq .. g NC, 4
,___Y_ w _________________y ___ ]_ _ , ,_---_L., ,-
7aE,['
%c, icEu. a; A. E r Gi n= J. C.6u J.
C t_ _p % {1-*-------------------I.__%-. L_ e t_I___ e _ ---1-> s, s, ,, s- m4. {.- l x me-SC3 9 9 0 SG3 Sc 3
- V 2':
i.(3 ' r; ' ! s' s n,
- m
) *c eposet peans ,, -l em e s,. *c ?mc
{.s v . __ e[ g.< _. J em .su : , - k T T L b l i 'o **
- i 9 i., ... i T acae M l C M- *'.---?'e se gs, l
o 0
( ( 6 6 b w v K Y _. A { .A,. A
-1 3,, I 'l,c =>
eI c 1'. . I g __4__t______.g___..._____ __ 3,_g w;. r. F .. t --c4_a
-"t* 4 me
_ _ _ _ _ , ,_ _ _ J- ] rY-H-F-------f_gy* r___Y %
"'" 1,
.;,7 _. .._1 r7,7;;p t_t' m g %)f,rv$ L. 4LJ F .J.F"1
- e a._ a rd?{
o *>e.__ J E.'_.*.*3 _ _ _ _ _ - _ _ 2- s t ._ . w _ _ _ _ _pr_7 - -[ - * ,____,.; y o v v .c . .t. F v _e.> l .c,. F ,j
.1e r~ . -
2 _V_y.,v _____.______y,__t q> h _ _ _ _ _ _ _ . e_ ._ - _,y,.,____
~
of, w . . ..
- n. . g
- e. 3,1, ,y =L _ . .v I :: c *---{
c !
- 1 s-
?.
r, e).,c
-d l
s N $ -- 3 l l l(
- m. -
Figure 9.5-5. N2 Distribution of RSB (Sheet 2 of 3) 9.5- 34 Amend. 6k Sept. 1981 1 2
- - - - - - - - - - - - r
9 I l i FIG. 9.55 SH 2 I W NONE F.C, r.........
?
l F AN/ COOLER l 6........c F.C. FIG.9.55 SH 2 8 NON E +--- 1
i. a i J mm um F.C. EVST 3rd LOOP AND PIPEW AY CELLS 3 ..........i ,N/ l
- A t X
8 s g Nor;E g t, F.C. N C. M._ NONE h
*W Y FIG. 9.55 SH 2 Figure 9.5-5 N Di s tri t,. -- in RSB 2
(Sheet 3 of 3) l 9.5-35 Amend. 61 Sept. 1981 ) i 1 e
P l 6
... X = c. pr,. o paAh p. .. o
..o... .
5
=c.. =c _ =c =c _ *c 'C
('*
^ 1 1 1 1
..... Y Y Y
, Y
,, c, .,
x .,_1. . _ eX.c f,.. x.c - .= . c.
. .. s &X.c M > "* H
[ I l To atu (m t afw .4 .. . ~ ~ 3
- To t .es .c .'. Y'0 "N h
d.a CW U
%d
= c.
f '0 *" i
),
..a >-c=:
* =.v.oota .v tv .vstru 1_. _. 4 e
k
[ (
- .wf,s. :
4.r-
- x. 3 y l Tita"'.". !.i"*
-d. *
).. ~4. .
.4 r----.
- -i
- t. __. _ _ ;
c.+ t. , w,.yme,s
. _ X s_
- -{'* ~-
{ s
.g__E c l
1 . . t t 4 A Y Y
.. ., v v 8
i
" -- X '. c-
. N 4-f- - f;,,-
- 3. 4~-~ , ,, ,, );.
===:-
o ._
~
) w. . ~
l o .. J 6 l l Figure 9.5-6 SGB N S"PPIY 2 9.5-36 Amend. 61 Sept. 1981 I f 6
r i et 4-4 SC-3 NONE iC.:
, u l
NT Se C4-- > LINE TO TANK E AST FO LeNE TO TANK E AST po FROM SUPERHE ATER FROM SUPERHE ATER M: 9 FC AwW > .7: FC
- =:-ew-- )
UNE TO T ANE WEST m UNE TO1 ANK WEST ( W =:WN ---4 7 -4 *:4y a
)
UNE TO T ANK E AST LINE TO TANK E AST W=W M l ) -4*>-+ y--4 > SWRPRS LOOP g, SWRPRS LOO 1P yf X - N.C. h
- =: x :
_. _ E F .O. no..S. V APORIZERS
\ yM -- -- -.. . . ._.-.
TO ATM ; ; s' [h TO ATM N C. ,Lo T NITROGEM SUPPLY Sv$ TEM v
, , -T- F.R_EO .T _,E N
. FIG. 54 l l
)
, I I ,
I r 4 i NONE SC-3 l g tEl 70 b>v L'NG TO T ANg g Ag7 FRuM SUPE Hwt Af g n
- ~ D,b b shI ] f NONE
- j
- SC 3 LINE TO I E V APOR ATOR WE ST
,9 g q~~~
LINE TO TANa WEST FC 0C
- I )
( 3
- LINE TO TANa E AST NONE h--* SC-3 WEST ,
abma { y y LINE TO Io tv4PORAT0ma l
)
SWRPRS LOOP 3 IVA ATOR E AST g, 39 '*
,c 0FC 'M~ >2 :N -'
, i E k,1* ec 4 ac i M.IC j). ,
g j b ?*. t ec ,
- I t '
q I N c.
- LINE TO 8
, tb APORATOR A l ' LiNe TO Q ,9 sc 1
- -WN : .j _
1 t ) %: ? '9:'c
- - :x :.r_
g suPenws A n n
)
t h, isc .AT ' 1
' ' [ k( $.tf A. GEN CELL - LOOP 2 P
I LINF TO
'fmx :.i -
NON .-p sc 2 (t$ in Au ciN ui L - tOO, ,
- .. L N. ,0 sv A=ATOa West J,e .?.
~
y NON, .. .. ,0, L_,
- 4(:,e l SUPER NTR LINE
) f
~
F0
- s T " ' ' '
I'- ",'*"
- c 37
,' ~ %, ,t ,i?
i
- l su_ren,NTn L t,Na i l 4HW >
l ml ~ ,
,' EdueAnn I Q !.f'C ,
a }
'c g sisAuce seit_i4 ',
, 4
- suPea NTat N.
-iHM M I
n -. n Figure 9.5-7 SGB fia Reaction ft 2SUPP IY 9.5 37 Amend. 61 Sept. 1981 9
/
I
(m l
~>~"
+.C. 4. .o 1 M: -+-<<---.
, , . . .x. -
.C 40=. m C. 4C nong y
.-x SC 9
NC y : : : : : ;. 2-;g.- 1 1 1 1 1 1 'o *o Y l y v v v
. , !, l .
4C 9 C. V9C U4C n 3," ** 4C f f f { it
W <r g , teO8eE h 80 FC
^
,ag s r
E PR588 Pfl C04?a06 g#44f t
'N
~:- >
[ej
- s E 4 " II M '
" SuP*LY TO G8
' ACFuAftp v.eygytt s SUPPLv TO G4,
.CTvATED WAtvil aumuaa, => msv ,g m -
- ; s.uPPo.r Cvo viotov.tv.
c, morrum rCrv.v.o , 1 rio v tvi. L i
_ _ _ . . . _ . ~ . . _ m ._ _ .___ .._. _ _ _ _ _ _ .___...m___ - . - . _ _ _ ._m.____.mm_.. - e 1 J 8
) . . ..!,".. .".a l E
CSLL 49te est ADE A g,'** l suone l Y c c, f , 4.. '
,% , c.
F;*
, Edcm 3
i .. _'s'vsrYu. . _ q Cui mc senses +-- e r--1 T. , b= , i = -ac
= :: :: :x
. x - -
_L j L=: L . .J j >i i W -H J. L-- r pan .1:coten i _ _J + i .I I I i i i Figure 9.5-8 RCB N2 Distribution High Pressure 9.5-38 f Amand. 61 Sept. 1981 e d
/
i
l t i
--+ NoNa s
g / RIG. S 5-8 X X r .c. I s .c. .r.c. $r.c. i N.c. X N.c. _[ Y v' r..-.--."
,...J...... ~~. 7 ., C -, '
3 -3
. . r,... .
; rAN/cooten ;
j eea7 etLL 123 .7Amecoottn; j mnme2,' a, , . enL 122 a, l.... .....J 6.
. . . . . . .. p 6... ...
' - - - - - - _ .g . J s
N N/ W .Y-V 4 V 4 o o NONE 7 0 a .C. ,W,N C. C 9 a .C. N.C.
- V_ _ . I V- - -
4 l , , 4 '4 -- e I
.l
e i I o NONE flG. 919 SH 2 v y
& 4>
. F C. F .C. F C. F .C.
I 45 1 d N.C M.C. [ -
..___.._ V Y
,.. . C. , r**--*******i*] ,,,,c, a
~ ~f_ =-------~'-- L-i l l l
, PHYS LOOP 1.y l
- FAN / COOL ER a e, CE LL 121 ,' FAN / COOL ER l f RE ACTO 4 CAVITY l
- CELLS 101 A THRU 101E .
'.,,.l,_,,
, ..., 6----- s
-- - r** , L L. -
= - * * * * *] ' ~
1 1'
\/
\/
- .... .- ' ~ ~ ~ ~-
s Cal - 7, NONE 9 d NONE 7 9 NONE F.C. N C. F .C. N .C, r T v_.. -- A NONE +--- q yn . L 4 FIG.11.36 SH I
, l HVAC
- H ),
N.C. I NV W1 ZONE 6E l NON E +--- Figure 9.5-9. N2 Distribution RCB, Low Pressure (Sheet 1 of 4)
- 9. 5 - 3,.'. Amend. 61 Sept. 1981 7 1
/
s a
l I l. I I w NONE SC3 n , 7.------. x F AN/C
;e; __{ Wl-~~--
f l
/ X LD FC p sG. 9 E9 l SH 1 1 I
-:= :--I
= l l
I I I F .C. l j i l l 1 l t l i
~ON 4. SC. l
!. r- - m c_ ____e I fl ra-aar NONS - ,e 7 g\g i L -_ y,*gEuc
+++ _.. _
a y l FIG. 9 5 9 SH 3 I I I PRIMARY S. F- Rf AC10R Ovfd CEC l l 1 I I
- 1 i
NONE++SC3 .k 6 au r g PTl C N C. f I i )
\
l i SC 3 eje f.oNE t i 4 C. ons .-- SC 3 d NONg , l
- ia
____._sk m, I y . .
;x -
Cm
),
' no. iu.
i ,c , l . V 1 - 't i none 9 ) no. iu. i I I I I 9
'.;"]
- .,N l
I I l l l l sC 3 *~o~. l gojw Il I c-- - s c .. -_ _s ..! n , _ . _ J};;i m ;
-i I
I I I u no..c , ! " "' d _.i.< q . ,"o " " g i l T l g HVAC i 1 E .,<, l >> no~. go, . . .
, c >-
.._i
; y ,
y l aoat - I _) r .o,. s.
---* aoa' Figure 9.5-9.
(Sheet 2 of 4) N2 Distribution RCB, Low Pressure Amend. 61 9.5-40 sept. 1981 a 3
. . - - - . _ _ b
I I I l .i l
.4 NONE l .
& AL E FIG. 9 5-9 F IG
- .a
-> - .9.5 9
.c
- >_r
.q OM m .., . x i ej a1 mc i i I
- .r oc : > 'i d j
d' I l l l 1 t. haK ( l fQUIPM] t.
- .: &, c >-
sc
- ; ,I I
d I I
&,.c- I
.: x I a e.c :
L .I 't.^ "-' NC I L
\
I t 4 wost e =,m.
+ ' .. :
- ) :.:
sc. 1 'Y
.~..__ ,,oii3.. i 1 ::. - ;
l 2: " y I + " d'* h l'! : : :.: u l S h b2 , i s
;,.c.
.; =o~. .M i
I
-t.: l 9
, ,e , 3.
I I I I m,~c ,,, l >1 < m
, .+.
Fi r s S :.: :- i' ;i ac I l 1 1 :< N C. waa. G a!
?l :.-7c- .: :-
1 NONE 3 F IG.1, 3. Figure 9.5-9. (Sheet 3 of 4) N Distribution RCB, Low Pressure 2 9.5-41 Amend. 61 Sept. 1981 I e
e NONE l x 1 s FIG. 9.5-9 SH 3 FC NONE
~
N.C. ,, SC-3 r- - 1-bfF * - r PRIMARY f.a MAKEUP PUMP--1---------i g L._ANP VALVE GALLERY CELLS ), l- w- H#F, wL4 ( SC 3 NONE l FC NONE c s b V_ _ ._
r i I i 4 l l FC FC NONE a> l . . .. i a 1' N.C.
$_-.-_. Y_ .. .__ g i
_ _. _. I 6 COOLFR j T 4 p _ __ _ ___J SC.3 4-+ NONE N ON E *--
.4 l :
M: y SC 3 F 1G.11.3 8
,, SC.3 NONE FC . C,
, NONE l
'5 NONE +-
CAPS i
- FIG.11.3 6 SH 1 NONE4 ]
Y N.C. l,'?; S5 9 Figure 9.5-9. (Sheet 4 of 4) N2 Distribution RCB, Low Pressure 9.5-41a Amend. 61 , Sept. 1981 l t
c ( F AN/COOLE R b I >- ::< ,@d-c=0 F
- y55 . --
CELL GROUP L - + *iv
^
6 9.55 3 gg, ; t F AN/ COOLER b W [ } X ghCL%0 > Mk-f FIG. 9.55 SH 3 e ,
"" ~
EVST CAVITY b
! P N NdM ^
b [g, /. j;'o 3855 xs i y
~ ~~
F AN/ COOL E R l ) C=0 h;1 h t*3 > G 9.5-5 Ca}-V ; CELL GROUP b I >- D< 'h;3Dd-t=0 ->- I FIG. 9.5-5 -C*$-F t--- ty l SH 2
- SPARE [ X g 1-{ e; b-I
%>1 ; - _ _
s W5 p-SPARE [ Om}--- $ QQ l CS)-b ; ... .[ ~
~ I SPARE [ X g 1._{gj y--
I l X->v c aL SPARE e 4 3 t _gg p_
--*NONE Ct}->< -
Q3 l_ .. .. ._ _ _ _ _ _ . . _ _ _ . N2 SAMPLING & ANALYSIS t i
i 4 U v 4 l / FICV FICV O2 , 9 v b
><l-- - -
~
Z l F.O. NON E 4---- F.O. g CAPS I
- 4*CxH )'
F.C. FIG.11.34 SH 1 sir "* N.c. INSTRUMENT Aia F a 2 catis Figure 9.5-10 RSB N2 Sampling and Analysis 9.5-42 Amend. 61 Sept. 1981
?
9
t 9.14 DIESEL GENERATOR AUYJJ.lARY SYSTEM g () 9.14.1 Fuel Oil Storage and Transfer Svstem The standby diest,I generator f uel oil storage and transfer systems are shown in Figure 9.14-1. 9.14.1.1 Design Sasis The plant is provided with two standby diesel generators separately driven by two diesel engines operating on No. 2 fuel oil with each engine supplled by a separate diesel generator fuel oil storage and transfer system. The diesel generator f uel oil storage and transfer systems are designed to ASE Section III, Class 3 and Seismic Category I requirements. The entire system is designed and f abricated to Seismic Category I requirements. The diesel generator f uel oil piping and oil tanks are designed, constructed, and tested in accordance w ith ASE Boller and Pressure Yessel Code, Sw: tion Ill, Class 3. The fuel tanks also meet all applicable requirements of N> 'onal Fire l Protection Association (NFPA) and Underwriters' Laborator % (UL). Mechanical portions of the diesel generator units and accessories associated with non IE S3 components are designed to ASE Section Vill, Division 1. The redundancy provided by the separate fuel oil storage and transfer systems ensures that malfunction or failure of an active or passive component will not Impair the capability to supply fuel oli to at least one of the diesel engines. h V A cross connection with two locked-closed valves is provided between the fuel 61lolltransferpumpdischargelinesfromeachstoragetank. 1 9.14.1.2 pfster Descriotion The diesel generatcr fuel oil storage and transfer systems, (Figure 9.14-1) located in and adjacent to the Diesel Generator Building, consists of the following:
- a. Two bur!ed diesel fuel oil storage tanks constructed of SA-285, Grade C material. The buried depth of the storage tanks and fuel transfer iIras is suf fIclent to provide protection against tornado generated 28 missiles. The description of tank foundation is contained in Section 3.8.4.
- b. Four full-capacity, electric motor driven, gear type fuel oil transfer pumps - two pumps for each diesel generator are fernished. Should one pump f all to start, the other automatically starts. Each pump is provided with relief valse discharge line back to its associated fuel oil storage tank. Eac h diesel generator fuel oil transfer pump is located InsI(., the d'esel generator building.
O V 9.14-1 Amend. 61 Sept. 1981
- c. Two fuel oil day tanks, one for each diesel engIno, are provided in 53 l the diesel generator building. Each fuel oII day rank is supplied 53 l with a flmo arrestor on the vent and is sized for 1100 gallons.
- d. Fuel oil storago end transfer system associated with each generater is j tested twico each month for one hour duration.
9.14.1.3 Safetv Evaluation As a result of the redundancy incorporated in the system design, as described above, +he diesel generator fuel oil system provides its minimum required safety function under any of the foilowing conditions:
- a. Loss of offsito power coincident with failure of one d!esol generator.
- b. Loss of offsito power coincident with maintenance outage or failure of one diesel generator fuel oil ..ansfer pump associated with each diesel generator.
Each fuel oil storage tank is sufficient for at least sovon (7) days operation of one diesel generator at the largest actual operating load, as indicated in Section 8.3, following a loss of all offsito power sources: Fuel oil can be delivered to the sito within 24 hours. The sulphur content of the No. 2 diesel fuel oil is specified at 0.5 percent maximum, by weight, to minimize corrosiveness of sulphur compounds in the diesel engir.3 exhaust gas. Any corrosion within the f uel oil storage tanks is considered in the design by providing either adequate corrosion allowances and/or coatings. Each fuel oil storage and day tank is supplied with a fl me arrestor on the vent. Tho capability to operate and monitor the diesel generator fuel oil system Is provided from olther the control room or the local diesel control panel. Fuel oil quality will be monitored by sampling and degenerated fuel oil will be j discarded. 9.14.2 Cooling Water System 9.14.2.1 Design Basis The diesel generator cooling system is designad to limit the temperature riso 61j f h ja ket water through the engino. Each of the two diesel engines has its own Jacket cooling water system. The engine Jacket cooling water pumps, heat exchanger and piping are furnished according to ASfE Section lil, Class 3 requiremen+s. 9.14.2.2 System Descriotion Each diesel generator cooling water system (Figuro 9.14-2) consists of: O 9.14-2 Amend. 61 Sept. 1981
- a. A shaf t-driven Jacket water circulation pump.
f'T Q b. One water temperature regulating valve, which maintains the engine Jacket cooling water at a uniform temperature, and includes a method of bypa::, sing the heat exchanger for f ast engine warmup. The temperature of the Jacket cooling weter - high or low will be annunciated !ocally and in the control room.
- c. One water expansion tank of suf ficient capacity to replace water evaporated in the Jacket water system. The low level of the expansion tank is annunciated locally and in the control room. Makeup to the i expansion tank is through a manually operated demineralized water 61l line. Corrosion inhibitor is used to control corrosion. I 53l d. Auxillary electric heating elements, thermostatically controlled to maintain the engine Jacket cooling water at a constant temperature of 1250F when the engine is not running.
- e. One AC motor-oriven water circulation pun.p to circulate the Jacket cooling water through the heating elements to maintain a keep-warm 53 mode when the engine is not running. j
- f. One heat exchanger suitable for maintaining the engine Jacket cooling water at the desired temperature. It is of the shell and tube type with the Jacket water flowing through the shell and normal or emergency plant service water fIowing through the tubes.
The diesel generator engine cooling water system is a completely self-N contained closed loop, with the engine Jacket water used for cooling the various engine components. The emergency plant service water system, Section 9.9.2 Interfaces with the engine Jacket cooling system at the engine Jacket Cooling wafer heat exchanger. 9.14.2.3 Safetv Evaluation The engine Jacket cooling water heat exchanger is furnished in accordance with ASE Boller and Pressure Vessel Code, Section ill, Class 3 requirements. The emergency plant service water system up to and including the engine Jacket cooling water heat exchanger is designed to meet Seismic Category I and ASE Section iI;, Class 3 requirements. (m 9.14-3 Amend. 61 Sept. 1981
9.14.3 Starting Air Su tems 9.14.3.1 Desion Bases Each diesel generator set has two independent, redundant air starting system 61l designed to be capable of starting the diesel engine without external power and also to moot the single f ailure criterion. Each independent air starting system is of sufficient volume to be capable of cranking the engino for 30 seconds or for 5 automatic starts, whichever is greater, without recharging 53 the tenks. This system is safety related in that the air storage tanks, valves, and piping between tank and air starting solenolds are governed by the ASE Boller and Pressure Vessel Code, Section lil, Class 3. All other portions of this system are not safety related. 9.14.3.2 System Descriotion Each diesel gor,erator is provided with two independent and redundant starting air systems (Figure 9.14-3). Each Independent starting system includes the following: i ,
- a. AC motor driven air cooled air compressors ll
- b. Two Air storage tanks
- c. Inlet air dryer
- d. All necessary valves and fittings 61 e. Instrumentation and control systems Each motor driven air compressor has sufficient capacity to recharge the associated air storage system in 30 minutes, from minimum to maximum starting 53 l air pressure (130-260 psig). Motors are furnished with automatic start and stop control. Power for the air start solenoids motors will be available from the IE power distribution sources in the event of f site power is not available.
Tha Instrumentation and controls on compressed air Tank up to the D.G. connection are seismically qualified. 33l 1 11 61l9.14.3.3 Safety Evaluation The redundancy incorporated in the diesel generator starting system design provides its minimum required safety function under the following conditions:
- a. Loss of of f site power, by putting into operation the standby diesel generator.
- b. Maintenance outage or f ailure of one of the two air starting systems associated with each diesel engine.
O 9.14-4 Amend. 61 Sept. 1981
The diesel generator starting air system including the air storage [. tanks and piping between tanks and engine is designed to meet Seismic x Category I requirements. 9.14.4 Lubrication SvstAm 9.14.#.1 Desfon Bases Each diesel generator has its lubrication system Integral with the diesel engine. The diesel generator lubrication system is designed to have sufficient capacity to ensure continuous lubrication of main bearings, crank pins, camshaf t bearings, valve gear, rocker arms, and all other oil lubricated wearing parts. The system is designed to Seismic Category I and AS>E Section ill, Class 3 requirements. 9.14.4.2 System Descriotion Each diesel generator lubrico+ ion system (Figure 9.14-4) includes the following equipment
- a. One direct engine driven lubricating oil pump.
} 53 b. One AC motor driven lubricating oil circulating pump to supply warm lubricating oil to the engine 7 ump and other necessary components during normal plant operation, when the engine is not running. 53 l c. Lubricating oil filters and strainers of the full flow replaceable cartridge type. 53l61 d. One lubricating oil cooler of the shell and tube heat exchanger type, capable of controlling the lubricating oil temperature at the required levels using the emergency plant service water at a maximum inlet 53I temperature of 900F. 4.14.4.3 System Evaluation The AC motor driven lubricating oil circulation pump ensures that the lubricating oil is always maintained at the desired temperature, even wh*le the engine is not running, by circulating the lubricating oil through siie lubricating oil heater. l The engine drl'ven lubricating oil pump discharge pressure is continuou.cly 61 monitored.
- (Q 9.14-5 Amend. 61 Sept. 1981 L
The lubricating oil cooler is f urnished In accordance with ASE Code Section lil, Class 3. The lubricating oil cooler is serviced by the emergency plant service water system, Section 9.9.2. During routino engine testing the lubricating oil 33 cooler Is serviced by the Emergency Plant Service Water System, Section 9.9.2. O l l l l l l 1 l O l 9.14-6 Amend. 61 ! Sept. 1981
i
.- l uje' , i COP.4 C f I ON LC M &
- m. ~
q b 1 vet ANSih
,, s uis
* ' I 2 v. c*a
- 2 +. Ma 7 %,? _
b v20 v2e i v24 ves Oft t, e
.' F W 6.
> >. m.e 1..wra m-
.e.
,m. .e . , C , ,,
t,
; E,: . . ,
Sn! r e \ .
?
~
!~ s-.-one s ?
a L -_ d W 34 VJs
- .~ m_._ r<t on vs via taabsif 5 be ..
*IM sov ses Q.:p set aos G 'p 1 mi us d
~
is' . ., ,i3 .~ 1 -6ec !" I 4 +et._-Cl_4C y 01 EEL OFNERAf0R SUILDING v.s v v., vs0 Ga _
--n: - -n; . m ,,
p.itwo run orrut rtat yn~ c e,a % ,,._
-.ece- 4isa1 r- i.eJ...oisc ). s 2NTY5c.
4,',,A si
. -vss vs4 E-4 3 vsr vse
;; 7 3 g
e
- -" e
- u. . l
)
]
g Nl- i i T_
=
O.
@" @@" @T @@
@ vss '.'
g o
" .g.
trwt rgv r<t .o.i is
.- ,.. n if55i'i- I s B o oitsct Ocwa.70= witois. -
O s T Amst Suup k o;'%- 2-+er.-oi e s ., , u'Q 7 ')
"}" #,,
4. 45 2 - c- I 2 - "a m.: :.: : - ;.: a, v21 v28 di v3' V 32 rWL Ott .sv faaWijTWhwP 5' [ -~ihidG5~i4~ 2->ec.-o o o , - ' *Q= i I - t i 4 f
, ~
2-.. o2o \ 96 :V37w.: + di v 3e f r- r , ?he%
' '"# D s., .. c. 9 == => c 9 l u2 u4 f
;4 .. Fu g'=
- l-6.C.4148 l-*40 DIESEt GEPs Oa BilltDING ,p
" E'?.0P' L-n - -g -- - . -
m ggvct j
'* M '*".*'.'7 m, I' Ws,s vse vs. vb0
+
E)L_ ' o 4m.4ias"4 , , , uANWOL (
+ o T i iT - O-
[@@@@ @@ H SH SMI _ 3 i girwt rurt Ort T
. @ '68Os'.' . . -
=
( ~swaarYir. 7hif55'iW- T f 9ttsrs Orwaafon suitoiNG , g am Taw suiA l 2 , e,. NN505A
GENERAL NOTES i
~"
r-- n i. sv mas annas:tvia s I l *l
l Ba c DOC D-0038 $
I , *. , L 2. ALL EQUIPw47 seamteg ese p XE0 12h a m T m .T m 3. at t twstinesut .nsieras a.e
,d b E E P5ILF afD 0 T UOS estee 88.r844, I ,.4.'E' , * !M,m"Fa'"O.'!llr1=f up.,
t4 L OIL Pvt.P 4-* elf e- 12mi -O l a 13 an!TTLm at 4-seeDa-014
;N S. u CLiaft.IICS$ CLASSIFICattopa E a
~~
ir S. SE15MIC CATE00RVs i UDO OIL Pts.p a 1. RADIATION ZODEnUNRESTRICTED U0S
, w. (" a , a S. SAFET, CLASSIFICATIONS 9C 3 (A19 l fmab H I; 2 - ma ' ' ' .,
x
- n!JP',to i cassana av **#, "#! '5,'f'l'!!!,J' 2 - c' m "
n
! u
, --} _ , '$ nt,J,.M,i, m g H. Jfa sa, L 0 ..
fj -Y[ fS4#5 1- g], 18-QIjfQcgaan asa Caos 1, u i{ .- __- -_ __ -
-a
$F--_ __ 1,"
=
--~~ 1 . .. i ro - . m . u .
.- -! . i m.i.rA.T.=;:Sif.la.t, au $ wuto ..'n.
a y 4 m ~s. m a '&.h m { l+ ,,
._1
* **.. { .
,Tt"Or R1" = '
a 3 e i u e i H y
@l,1
.,,t.wo,ca< ..
x 5 , v--
-H arla ab. [--- !
g[N " b lo lo.u..<=aH $ , l 2 _. - . 1._,o,,,,<<_- r - L- ' w -=c
---l
,- ee "
0 igs N f e ,V m fam a a
,,.b,.., l jg.n .,ggg 1 - i e r 3 . _. m "
}"
h cafpgi. den
'#,'!," " REFERENCE DRAWINGS
- i. rmL O s it.. ,g ,,,, ,,,,,
2-.- S' 5 l
--c*
- v
_H fmah.H E 2 -* -c'a : : l -ain +<acca ,~ , H ~e h4g p L.. __ . " ~--d20 " --- - -'
- - - ---j
~{ __ -- -- --
l y .
-. i e _ i m a
'-A 4.>-
} ;" '
,,,h i * '
l
.. ,T?OTRO l
-*> I ra 3 ^m . e m
'., m 3' ;
. .jp, T,,A TO "
... y ,
( - . " . - .m ~ pl= ai'g [' l 2 .e.i-e - i l
--..m.,m i.csrLo L._.
cm __E" s... - _jl _. p
. - - : L- _ _ _ ___- . __ -- --_
Figure 9.14-1 Diesel General Fuel Oil _m " ? Storage ced Tronsfuel Systein 9.14-7 Amend.6d Sept. 1981
?
r m % ir E8 db o g... . . - - . - . A A e,
,me.
s 8 di g I O
- 4 il i 1 i e I e
, T. Tt Tse rest vu m I
.e .
O G r (p u - - p o h~fdEI. o . C G Y
- 4
' "s'.? x %g Casicutafeces puter i F .etaess a ,
f l L
i i t.
- s. ... m. =n = 4 w
1 :::: '::!::".?.'.".".* '"#".W"'
. " *l.::=,"A" 4
y .. ~ g.g_ ... ... If T,'." ::!'". I vv I b "" I _g, w
- - W %
h m
,~ 6 I :
j F
-*-r m, l
G u l Figure 9.14-2. Diesel Generator Cooling Water System Amend. 61 9.14-8 Sept. 1981 3 a I o L
P l
.in T.fM b,' 7 II II 11 to $$n NW WN men 1 .in cristirsson
[ couparsson ccaril T2NT.Dson 56ir TfwT166TC T26.Tia6Ti-sf :: : : :: -}-g af
,, ,, ,, ,, -2 ,,
x x x x x ir ir : : ir m ir ir m r r , ~ l e ,,-
--6 , , , , -
-~6 s,..,,-
4 s,..,,,- 4- l
.m .. m ..
firin GiTu fIiiii fiE fia
,2hTf5o' s. e2wTTEoac 12wTf503. 12NTI6i33G 82hlf 0038 4 4 4 4 r s - s - s S 6 6 6u-Y i x . . x x .
Jy , x . e e i L. __ __ _ ___I l________ __ _ _ __J L_ _ _ __
- ?
? ?
9 9 E
! 1
_ , , , , __q _ , , , , ,
..m. : , 22.
E..ES' E._.
'E..EU' E...,
s y y y y . . .
""M$
"'" h ""16 6 a
~
~
-J -
~
g 66a - 6 -6 G> : 6 66 + 6 E 6 - 6 af'!a + + af'!5
= = .fis72WW = ; = '
'*. !?%1wW <
L__ _ _ I ____] l____ _
\
flN504A
\
l GENERAL NOTES a
- __ __ aim t- snro 7__ __ __ __ ,' TJe g)47 "**
__._.____l WJbfisi'"Ji " u = ..w .. a .~_._ i ,
= = . = =
- m. ,.,0. . .. . . E. ..
> g3pyg] N W. Z Z ^& M g)p yen A IR l g SCR g. 3(gggfC CATEGORY: 9 UDS g 1. R&Ola,10N ZOPE: UDfES,RICTED U0S I S. SA8Ef 7 CLAS.,FICATions 9C-3 UD5
= = = x % -
=am---
, .n = .= g= .r.
_u y i a; pag p;fa 'g,=,=gaa u l __ ___u,_- _. . _ 1 r
- < :< x o u o r m r 4 ,,,,-
4 ,,..,,. 4 ,,..,, 4' iTF aTA- aia fia fia i2NTF00m s2fra T T603r i >Tf60am. 4 4 4 4:
- s -
D ve,e 6 m x . 6 6 , x ' Y., , ,
. s.
E o i ? t p AEFERENCE DPAWlNGS { _ l
+
I 66 ._mi -:_. + 66___st.._sx_::__ V - 4 V - ex - 3x_:
~
~ O- 6 6 66 6-~E ,
!!!'!fd + + L E'!fa +
= l '
'" !f a Y'J s ?
= i
(#2?'J&P __ l ._ _ _ _ _l Figure 9.14-3 Diesel Generator Starting , Air System
/
9.14-9 Amend. 61 Sept. 1931
F-t
~
LM 1 a 1 gg W 4GHS G4898 EDUAG eseTCMS P90w8080 SV GWR 98LL ~ s _ Wet.t.e.e.tt.e... A es stA".esYaas x ....._. 1P 1I db db 1P dk i " i
- l(
N tusf Ost puaar b use g,3 -. 'S"L X 4 L 0 av.=n ev iu DeCuteflose puesP fi k d b i o 1 n l x t I t !' i l l s------A--------_.--._----._------_------ --_N _ - _ . - - - - _ - - - - - - . . . _ - - - . - - - - - - _ . _ _ _ - - - _ . . . - - . . - - - - - - . . - - - - - - . . - - . _ _ - - - - - . -.._x._ _ . - . . - - _ . . - - . _ . _ - - - - - - - - .
e i I J
"I'",'"' g".,:g.,*.'lX'%'*".*4".'
.-..m.
- OR
.8s SE av . OUTLE f L;".'.T.'"
"l','.'~
;;.". ;L. ,
..'.".'.L.
g "
! ! & 4
) - 8886 $6 -. T@~ 2.'. J
- ;L x vv y y ,o T C T V i A
E V= A l x p:
..lL',':L.
v a, Figure 9.14-4. Diesel Generator Lube Oil System 9.14-10 Amend. 61 Sept. 1981 5 1 i e
9.15 EOUIPMENT AFD FLOOR DRAINAGE SYSTEM
/ \
C) 9.15.1 Design Bases The plant Equipment and Floor Drainage System (EFDS) is designed to collect 1he 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 tho maximum postulated flooding event such as a pipe rupture, tank rupture, or sprinkler discharge and limits water accumulation on the floor to no more than 31/2 Inches. All safety related equipment is mounted on pads et least 4 inches high. 9.15.2 System Descriotion Seperate 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 full 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 waste treatment system sump drains contair,Ing potentially radioactive fluids are routed or pumped to a main collection sump in the RSB Radwaste Area and
) monitored for level of radioactive contalmination. If the sump Influent is contaminated, it is pumped to the radioactive lIquid waste sump where it can then be processed thru the radwaste treatment system. If the sump Influent is not radioactively contaminated, then it is pumped to the waste water disposal ,
system. Treated water and other process water treatment wastes which do not have 1he potential to be radioactively contaminated, are routed to seperate sumps for transport to the waste water treatment system. i Where there is a potential for oil spills, the drainage is routed to oil Interceptors prior to discharge into the waste water disposal system. Oil spills are not allowed to drain in areas that contain radioactively i contaminated equipment or fluids. In this case, the oil spill is contaminated with curbs and dikes and removed manually. Oil routed to the oil Interceptor is collected in a waste oil tank and removed from the site for subsequent disposa!. 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 deiectors. These leak detectors are provided to detect and identify the 61 location of any water leak. O V Amend. 61 l 9.15-1 Sept. 1981
61 9.15.3 Safetv Evaluation The plant equipment and floor drainage s> stem is designed so that it is not reasonably possible for any radioactive ucoinage in these systems to be discharged out of the plant without undergoing the required treatment or processing. 61l Evaluations of radiological considerations for normal operation and postulated spilis and accidents are presented in Sections 11.2.5 and 15.0 respectively. The plant Equipment and Floor Drainage Systems is not safety related except for the piping and valves required for containment isolation (Section 6.2.4). EFDS piping within areas containing safety relcted equipment is supported with 61 Seismic Category I supports. 9.15.4 Tests and Insoections EFDS pipes embedded in concrete are leak tested. I.Il EFDS piping is tested for leaks after installation. All leaking pipes or joints are repaired before 61 the concrete is placed. All pumps are tested to ensure that their perf ormances meet the required design flows and pressures. A check source will be provided with the radiation monitor to ensure its operability. 9.15.5 Instrumentation Acolication Each sump is provided with automatic controls to start and stop and alternate operation of the sump pumps. In case the lead pump falls to start, a high level switch automatically starts the 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. l l 9 Amend. 61 9.15-2 Sept. 1981
9.16 Recirculating Gas C991tng System ks _/# The Recirculating Gas Cooling System (RGCS) provides heat removal capability for primary CRDM, primary Na makeup pumps, EVS Na pumps 6nd cold trap and for inerted cells in the RCB and RSB and maintains the cell temperature below a level which would be deleterious to concrete, electrical wiring, Instrumentation, compon>nts or equipment. The RGCS is comprised of 13 Independent subsystems, 8 of which are located in the RC8 and 5 in the RSB. Tabl e 9.16-1 lists these subsystems and their seismic and safety classification, Table 9.16-2 lists major system parameters. 9.16.1 Deslan Basis The RGCS is designed to provide the following capability:
- 1) Provide heat removal capability and maintain the following required temperature in the inerted cells of the RC8 and RSB.
a) 1200F nominal cell gas temperature under normal operating conditions. b) TBD OF cell temperature under off-normal operating conditions. c) 1500F concrete temperature, except local hot spot which shall not exceed 2000F during operating condition, d) less than 3500F concrete temperature for a period of 24 hours. A ( ) 2) Provide cooling gas direcTly to the primary Na makeup pumps, EVS Na pumps, EVS Na Cold Trap and the primary Control Rod Drive Mechanism.
- 3) Prevent leaked water from the cooling coil entering the cells containing Na or Nak components.
- 4) Isolate the RGCS components in the event of a Na or Nak spill or leak.
- 5) Maintain the independence of the served redundant system.
47 6) Isolate Individual cells f or maintenance. 61 59
- 7) Provide accumulators for air operated valves, sized with sufficient margin to perf orm their safety related f unction f or the duration required.
9.16.2 Svstem Descriotion All RGCS subsystems except f or the Primary Control Rod Drive Mechanism cooling (CR) subsystem operate at approximately ahnospheric pressure. The subsystem (CR) operates at 100 psig. A typical low pressure RGCS subsystem is shown in-Figure 9.16-1. All the operating equipment such as fan, cooler and valves are located outside the Inerted cells in normally accessible areas. The return 47 gas from the cell is drawn through piping embedded ( .
\. J 9.16-1 Amend. 61 Sept. 1981
in shielding concrete, an isolation valve and a cooler by a fan located downstream of the cooler. The cooled gas is supplied to the cell through an isolation valve and piping embedded in the shielding concrete. Inside the cell gas is distributed by the ducts. The isolation valve is located close to the cell and the inerting and deinerting connection are provided on the component side of the valve to facilitate inerting and deinerting of the cooler and fan without deinerting the cell. served. Figures 9.16-3 through 9.16-7 show the P&ID's for the various RGCS subsystems including the identification of their safety, seismic and code classifications. Table 9.16-2 lists major parameters of these subsystems. The low pressure fans are direct-driven vaneaxial fans with manually adjustable blades. The motors are totally enclosed, nominal 460V, 3 phase induction motors with NEllA class 'H' insulation. A typical cooler contains commercially available cooling coils in a factory fabricated AT Code rated casing and is shown in Figure 9.16-2. 9.16.2.1 Primary Heat Transport Systems (PHTS) Subsystem PA, PB, PC Each of the PHTS cells group is served by a separate RGCS subsystem to maintain the independence of the PHTS loops. Each sub-systems consists of one 100% cooler and one 100% fan, as shown in Figure 9.16-3 and cools (1) the PHTS cells, (2) the associated hot leg pipe chases and (3) the associated check valve cells, up to and including the reactor cavity bellow seals. The supply duct distributes cooled gas to each of these cells. The return from the hot leg pipe chase and check valve cell is through the clearance around the pipes as they penetrate the neutron shield wall. The return gas from the PHTS cell is drawn h from a high point and passes through the shielding block located within the cell. The return gas is drawn by a fan through the cooler and supplied back to the cell. 9.16.2.2 Control Rod Drive Mechanism (CRDM) - Subsystem CR The subsystem CR consists or two 100% coolers and two 100% blowers, is shown in Figure 9.16-3 ano nools only the primary control rod drive mechanisms. A separate subsystem is used for this purpose because of the high system pressure (100 psig) and high system pressure y drop (8 psi). The pressure relief valve is set at 150 psig. The return gas passes through a manual butterfly valve, the operating blower, the cooler and an automatic butterfly valve before returning to the PCRDMs. Unlike the low pressure subsystems, in this subsystem the blowers are located upstream of the coolers. This is necessitated by a large energy input in the blower resulting in approximately 45 F rise in gas temperature. The blower operates in 59 47 l a gas envir onment at a maximum temperature of 175 F. Amend. 59 Dec. 1980 9.16-2
?
i
.. t L c, e
dib .A e L_-
& I a e' JM' ::
FO P 0. 40 m as ConsPass F8 Lit A L___ h K3 l
< OZ 90 '
FtG. 9 63 SM 1 NYS ##sts Co VEp VEBSEL X sas saw s svf 4 I s
= c. ;, a t .
I
,,
- c. .K= L e-.
+
- 40. 8 mao Ga5 0 PtLTER { ,,C04AP8f_
SC 3 I l W ss. n se su i W E2 pett Dutsp Lust OL SYS OL Tmap g noon na si FIG. 9 Ft Sp 3 e l Cas putsp Lute WL SYS OL Taar g stor =o a p
,FHl. 9 43 EM 3 e
PRtiLoo CUhI,P noLtJGt n o0L SY$ @L TAM l FM 9 62 EM 3 e h pasasang see ovenFtour vesset l W em vs sw a ,, 1 1 I i l I I 1 I l I 1
a I I a
- p. _.
,, .p f.. . . . .>.M.c ___
ac C GC .
.~~3 :s: ... ,,
f
' a
,. i
- _, 'rJ c
!=:: . ;c .,
li . . _ _ _ j_f_. - , 10.':".. = c-I o. r + w .. - . --
- 10. 40 e
., -dV' ,
y,~ c
-. ,, T i
,,,, .s_ -- %.I> .
__ _, '..I ---<
'<=f. C;;%
=0 W " "'""
______4-j , >p.%
, c
+ " ' "
l'; ,,.h s=~ :n
; c
~~
( so.. . l 1, p
- I ,, ,, o t F_--+= 4 c.n ..co. .. .u ,,
I
\
.o.....
o . .. a ;
,c,,__
f, o ,, ..
-.m e
- g. ,,,
=c t. . .- : , ]* g m m.=v
. c. . .... .
_.=> =cy
+- .c I
'. con t j va Figure 11.3-4. Flow Diagram - RAPS Amend. 61 11.3-43 Sept. 1981
- I
+
) < 1...... I ) ' [ F AEEse AaGOpp
, 9 us. .gg _ -us I . :-:
"O
- I: 7
- x .:
sc ac g en.o.v., ears sumos _ .-} ro t Fe
,8.... y ,
( % d*.
' c-I I. vacuu.
vesess I wi.-4-. un 1 x L__ a'"" se
'l j? j l I. ... . ;"o'#,T'5 '
( I..c,ct oo.
- wATwT-aa i ec ec *o ..asi as :
1 =c L,?,
, , ,"go,,,I x l svoaaos vines x l W * *'
f"8 m.... a,'o',',,7f M,*, j cans v acuu= vinre 6 (
.< F-- W ; e ii n
.
- SC 3 --
=.
p=. A+---y
- . .. .. X
< 7E r....
es,AC guneaugT
- o. aTo seef tRSACG vtni 70 ctLL p ._. . _.
1 I i h'
P 4 1 I a'Ohd SC-3 i b k I i i 6otaf EMCeentsGan m_ _ g 9.6 4-7 O. c$ k M +--g E E #ELIEF f
' esf aygn asOset h ATA [D wvn
]
( / Cav0GEhtC STILL [ -, 7,.g. ..L .
-e
, O.e. -
valves i smis vaLvs actuar,On RAPS COLD 00m Figure 11.3-5. Flow Diagram RAPS (Sheet 1 of 2) 11.3-44 Amend. 61 Sept. 1981 -
/ t l
l
+d FUEL F AILURE MONITORING I
-1 1
! l F.O.
i e
- NONE NONE W-> SC-3 VENT TO OUTDOORS SC-3
[.C__. _ LO . F.C. . ...
~;E ^ ),
F.O., F.O. p ig, ,3,3,7
}(N.O. SC.3 + d NC LJ SC-3 $ .'?-
o NONE
~.
i : e:N. - l : 1 i l i I Figure 11.3-5. Flow Diagram - RAPS (Sheet 2 of 2) 11.3-44a ' Amend. 61 i Sept. 1981 i o
s I i j I i r;1, ! 1 r
- m...1. A+, : P.
2a.L. ov}---- o [~~ a ?* . . . .
>- ' r l g I
- =y :
'c .,i.
[ . . . , . l f >-
..e..s .
I >- I w
.. ;"," 2' l -- n Sc 3 m.
.s 0 I
,m . . . . . i > %c E-3 Am ,1G . 3 S.4 9 l p I, C.
KE
,eG t13. p 3 ,,
,4 f t ) l }
l
.c a
...F >
j 4=>0-X'c- ..
. , , . i > %
. * ,o ' > Q' -~
..o n sa l > .s o{ -
*a
..... , > =
.. ... > l > -
ri ,,o , , , , j >_ o . ... . i > ,,, ,
'***'**I F l
; "5 % H. , ,.
, - . . - . i >
.. n .......i > .. ... . ; r : : :
III ****'** >
,.c... . i y ,,,,,,,,,,i y ,1
~ *
~
.m . .. I >-
Toe ,
.E GAL
.. . . ; . . . i > . .o . .. - i i w ..
- w s_.J . b d
. . . . i i
i figure 11.3-6 Flow Diagram - CAPS (Deet 1 of 2) Amend. 61 11.3-45 Sept. 1981 1 3 I l
I e I _.N O.
, x s N C. nsv
'=: = MJ l 3 i e
N O. RAD GA5 FILTER ; M COMPRESSOR N C. I __..e, % g g u. N O. MAD gas FILTER ,
- .m i x X
O b. 1 N O. :
-3 COM*RESSOR I
V ACUUM PUMP
- M [1 N O. NC I )
[ e RADW ASTE DISPOSAL SC 3
, WASTE
( V ACUUM VE S$E L pq NC ] yg,9, ,gy,, gg FIG t114 $w 1 NC a( NO i i R.EMO,,V,U,
.e NIT SC-3 SC3 w
T
- xy o.
NO NC FIG 11.3 7 a f RAD WATE R x MOLDING VESSEL i SC 3 e C i l 4
I
\
a
) CHtLLED WATER OUT IN r CHeLLE D g WATER IN Esc
. NO$ ,
NONE 1P db F O.
-,N SURGE VESSEL A ) CHILLED WATER OUT l WATER N b
c th i HEATER NC OntDeZER HE AT EXCHANGER TReTIUM WATER
. .. REMOV AL UNf 7 '
y ) g'O.
,, ND.
TRITIUM W A TER 88' 0 ' RF MOV AL-UNIT f ,,q , , $ y R AD W ATER g F fG.11.3- 7 NC ( g I HOLD. vessels < g3 % ! 1 , TRITIUM OxtDtEER , i g SC 3 e sca H HE ATING & VENis TION
. 3 F.O -.F C. g.o, 4 ,
' RAP 3 HgAT E XCHANGhR$
1 F G. i s . , Iigure 11.3-6. Flow Diagram - CAPS (Sheet 2 of 2) I I ' 3 #'d ' Amend. 61 Sept. 1981 t I o 1
l' m, h GN1N 1 W
. ... ' l 8eoha esONE
. . p- --
l ** g 9 C. m- _
~ ;
e 4 SGS F0 NE.7 D=YE 8 x.CM.4GE.
" .C3 h-* MOsse
...,0...... ,, g, i .,G ,, ,.
go,, I
- Q,c =
1r I e cp -- ., F0 w . ,. pg l 348088 TOnessG ST.7:080 . oc l d " g f sG 5 3g
# G-
- ' ! b_.J f
. I' h4.....
. .c.
G.. i ,, ,,
< 'c8*--* *cae l
,' ato ..,en on..
j , TO seotosesG wegget l Fe6 54 34
- v. twas
[ ~ - .
. - . . . . , _ . ~ ,
F,,,., , , , , _ t
t
\
7 1 4) I ros es no t e p to te u K3 i H ** -4 L_ .. a. ( 1 F4 913.See 3 l l} }% t 80 0 ip
#0 h yC OI 80 db MJg IR t.T. ?r""
-.3 { _, _
3., : .. 1 _ pp PC so +C
- 1 i W At W63
~
. cap.~
WatW9 ACTVATeces (
~~
M wesef 90 tsit (
%<sume.
' M. . .. . .I I Figure 11.3-7. Flow Diagram-{APS Amend. 61 Sept. 1981 11.3-46 i
a
/.
_ _ - . - _ _ . _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ . _ . _ _ _ _ _ . _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _]
O M Nore! -- **
, REACTOR CAVITY C &
l
. F 10. 9.59 SH 1
> X hW N HTS LOOG 1 CELL I
i *iG. 9 b9 SH 1
) X hW M
+
HTS LOOP 2 CELL ' X ' 4 [ } IFIG. 9.59 SH 1 '~ i HTS LOOP 3 CELL " l > X #
, FIG. 9.59 SH 1 -
1r
' R')
.CE LL GROUP l
i FIG. 9 E9 SH 4
) X
- hW >
F AN/COOLE R JL [ e FIG. 9.59 $H 4
) X ' W >
CE LL GROUP
- I ) X b!P ' '
# FIG. 9 69 SH 2
' N ~
F AN/ COOLER D C I ) X W > F .C. F eO. 9 59 SH 2 M -
- F A. COOLE R *
{ } X FIV . I COMPRESSOR I FIG. 9 59 SH 2 N - I
. CELL GROUP g JL l ) N ' F
' FIG. 9 59 SH 2 N -
e 1 .. .. -- -- N2 SAMPLING & ANALYSIS UNIT 1 1
1 I 6 l l FICV FICV 902 e-HO2 l v E e & DG--- l l , ~ c-E n F.O. F.O. NONE CAPS l c-N--ec4-4 > F.O. FIG 11.3-6 SH 1 l1 O EA'F8s"$' c^ 8-Figure 11.3-8. RCB N2 5 mpling and Analysis Amend. 61 11.3-47 Sept. 1981 t i 1 e
CHAPTER 13.0 COPOUCT OF OPERATIONS O i V TABLE OF CONTEPRS EEE 13.0 CHAPTER 13.0 COPOUCT OF OPERATIONS 13.1-1 13.1 ORGANIZATIONAL STRUCTURE OF THE APPLICANT 13.1-1 13.1.1 Project Organization 13.1-1 13.1.1.1 Functions, ResponsibilItles, and Authorities of Project Participants 13.1-1 13.1.1.2 Applicants' in-House Organization 13.1-1 13.1.1.3 Interrelationships with Contractors and Suppllers 13.1-1 13.1.1.4 Department of Energy (DOE) Participation 13.1-1 13.1.1.5 Technical Support for Operations 13.1-1 13.1.1.5.1 TVA's Technical Staf f 13.1-2 61 l 13.1.1.5.1.1 Division of Nuclear Power (NUC PR) 13.1-2 13.1.1.5.1.2 Other TVA Orgsnizations 13.1-4 13.1.1.5.2 Project Technical Support 13.1-4 13.1.2 Operating Organization 13.1-5 13.1.2.1 Plant Organization 13.1-5 13.1.2.1.1 Plant Operations Section 13.1-5 13.1.2.1.2 Plant Engineering Section 13.1-5 13.1.2.1.3 Plant Mechanical Maintenance Section 13.1-6 13.1.2.1.4 Plant Electrical Maintenance Section 13.1-6 13.1.2.1.5 Plant instrument Maintenance Section 13.1-6 13.1.2.1.6 Health Physics Unit 13.1-7 61 13.1.2.1.7 Nuclear Plant Management Services Section 13.1-7 O 13-I Amend. 61 Sept. 1981
TABLE OF CONTENTS (Continued) em O 13.1.2.1.8 Division of Power System Operations (PS0) Engineering Unit 13.1-7 61 13.1.2.1.9 Nuclear Plant Quality Assurance Staff 13.1-8 13.1.2.2 Personnel Functions, Responsibilitles, and Authorities 13.1-8 13.1.2.2.1 Plant Manager 13.1-8 13.1.2.2.2 Assistant Plant Manager 13.1-8 13.1.2.2.3 Plant Operations Supervisor 13.1-9 13.1.2.2.4 Assistant Plant Operations Supervisor 13.1-9 13.1.2.2.5 Plant Engineering Supervisor 13.1-9 13.1.2.2.6 Plant Mechanical M3Intonance Supervisor 13.1-9 13.1.2.2.7 Assistant Plant Mechanical Maintenance Supervisor 13.1-9 13.1.2.2.8 Plant Electrical Maintenance Supervisor 13.1-10 O 15.1.2.2.9 Assistant Plant Electrical Maintenance Supervisor 13.1-10 13.1.2.2.10 Plant Instrument Maintenance Supervisor 13.1-10 13.1.2.2.11 Assistant Plant Instrumentation Maintenance Supervisors 13.1-10 13.1.2.2.12 Health Physicist 13.1-10 13.1.2.2.13 Supervisor, Nuclear Plant Quality Assurance Staff 13.1-11 13.1.2.2.14 Safety Engineer 13.1-11 13.1.2.3 St.ift Crew Composition 13.1-11 13.1.3 Qualification Requirements for Nuciear 61 Piant Personnei 13.1-12 j 13.1.3.1 Minimum Qualification Requirements 13.1-12 0, Amend. 61 i 13-11 Sept. 1981 i
TABLE OF CONTENTS (Continued) 13.1.3.2 Qualifications of Plant Personnei 13.1-17 61 Table 13.1-1 13.1-18 Figure 13.1-1 13.1-24 13.2 TRAINING PROGRAM 13.2-1 13.2.1 Program Description 13.2-; 13.2.1.1 Program Content 13.2-1 13.2.1.2 Coo. <tination with Preoperational Tests and t-uel Loading 13.2-2 13.2.1.3 Practical Reactor Operation 13.2-2 13.2.1.4 Reactor Simulation Training 13.2-3 13.2.1.5 Previous Nuclear Training 13.2-3 13.2.1.6 Other Scheduled Training 13.2-3 13.2.1.7 Training Programs for Non-Licensed Personnel 13.2-5 13.2.1.8 General Employee Training 13.2-5 13.2.1.9 Responsible individual 13.2-5 13.2.2 Retraining Program 13.2-6 13.2.3 Reptacement Tralning 13.2-6 13.2.4 Records 13.2-6 13.2.4.1 TVA 13.2-6 13.2.4.2 Plant 13.2-6 Floure 13.2 -; 13.2-7 61 O Amend. 61 13-III Sept. 1981
TABLE OF CONTENTS (Continued) 13.3 EMERGENCY PLANNING 13.3-1 13.3.1 General 13.3-1 13.3.2 Emergency Organizatfor. 13.3-2 13.3.3 Coordination with Offsite Groups 13.3-4 61 l 13.3.4 Protective Action Levels 13.3-4 13.3.5 Protective Measures 13.3-5 13.3.6 Review and Updating 13.3-5 61 l 13.3.7 Medical Support 13.3-5 13.3.8 Drills 13.3-6 13.3.9 Training 13.3-6 13.3.10 Recovery and Reentry 13.3-6 13.3.11 Implementation 13.3-6 13.3.12 Radiological Analysis 13.3-7 13.3.12.1 Projected G.ound Level Doses 13.3 -7 13.3.12.2 Accident Assessment, Warning and Evacuation Times 13.3-7 61 13.3.12.2.1 Assessment 13.3-7 13.3.12.2.2 Warning 13.3-8 13.3.12.2.3 -Evacuation 13.3-8 TABLE 13.3-1 13.3-10 TABLE 13.3-2 13.3-11 13.3-12 TABLE 13.3-3 13.3-13
^U ' ' ~# * ~#
61 O Amend. 61 13-IV Sept. 1981
i TABLE OF CONTENTS (Continued) Figure 13.3-1 13.3-15 Floure 13.3-2 13.3-16 Floure 13.3-3 13.3-17 Floure 13.3-4 13.3-18 Flaure 13.3-5 13.3-19 Floure 13.3-6 13.3-20 61 13.4 REVIEW AND AUDIT 13.4-1 13.4.1 Review and Audit - Construction 13.4-1 13.4.2 Review and Audit - Test and Operation 13.4-1 13.5 PLANT PROCEDURES 13.5-1 13.5.1 General 13.5-1 13.5.2 Normal Operation Instructions 13.5-1 13.5.3 Abnormal Opersting Instructions 13.5-2 13.5.4 Emergency Operating Instructions 13.5-2 13.5.5 Maintenance Instructions 13.5-3 13.5.6 Surveillance Instructions 13.5-4 13.5.7 Technic:! Instructions 13.5-4 13.4.8 Section Instruction Letters 13.5-4 13.5.9 Site Emergency Plans 13.5 -4 13.5.10 Radiation Control Instructions 13.5-4 Figure 13.5-1 13.5-5 61 OG Amend. 61 13-v Sept. 1981
TABLE OF CONTENTS (Continued) en O 13.6 PLANT RECORDS 13.6-1 13.6.1 Piant History 13.6-1 13.6.2 Operating Records 13.6-1 13.6.3 Event Records 13.6-1 61 13.7 RADIOLOGICAL SECURITY 13.7-1 13.7.1 Organization and Personnel 13.7-1 13.7.1.1 Division of Property and Services 13.7-1 13.7.1.2 Office of Power 13.7-2 13.7.1.3 Employee Selection 13.7-2 13.7.1.4 Employee Evaluation 13.7-3 13.7.1.5 Industrial Security Training 13.7-4 13.7.2 Piant Design 13.7-4 61l 13.7.2.1 Design Features 13.7-4 13.7.2.2 P' Arrangements 13.7-6 13.7.2.3 Owner-Controlied Area 13.7-6 13.7.2.4 Protected Area 13.7-7 13.7.2.5 Vital Equipment and Vital Areas 13.7-7 13.7.2.6 Alarm Station 13.7-8 13.7.2.7 Security Barrier 13.7-8 13.7.3 Security Pian 13.7-8 13.7.3.1 Access Control 13.7-8 13.7.3.2 Control of Personnel by Categories 13.7-9 13.7.3.3 Access Control During Emergencies 13.7-9 61 O Amend. 61 13-vi Sept. 1981
TELE OF CONTENTS (Continued) 13.7.3.4 Surve!Ilance of Vital Equipment and Material Access Areas 13.7-10 13.7.3.5 Potential Security Threats 13.7-10 13.7.3.6 AdelnistratIye Procedures 13.7-10 13.7.3.7 Test and inspections 13.7-11 Floure 13.7-1 13.7-12 Figure 13.7-2 13.7-13 Floure 13.7-3 13.7-13 Figure 13.7-4 13.7-13 Figuro 13.7-5 13.7-13 . Floure 13.7-6 13.7-13 Floure 13.7-7 13.7-13 F laure 13.7-8 13.7-13 F loura 13.7-9 13.7-13 Floure 13.7-10 13.7-13 [?gure 13.7-11 13.7-13 61 i i 'O Amend. 61 13-vil Sep t. 1981 I
CHAPTER 13 CONDUCT OF OPERATIM LIST OF TABLF$ TABLE NUMBER TITLE PAGE 13.1-1 Technical Support Summary 13.1-18 13.3-1 Participants in CRBPP Radio- 13.3-10 logical Emergency Plan 13.3-2 Summary of Data Utilized 13.3-11 for Source Term Radiolog; cal Analysis 13.3-3 Projected Maximum Resident + 13.3-13 Transient Population Distribution Within 5 Miles of the Demonstra-tion Plant for Census Year 1980 13.3-4 Projected Maximum Resident and 13.3-14 Transient Population
- In Evacua-tion Sectors Within 5 Miles of 61 CRBRP O
O Amend. 61 13-vill Sept.1981
CHAPTER 13 COEUCT OF OPERATIONS LIST OF FIGURES FIGURE NUMBER TITLE PAGE 13.1-1 CRBRP Organization Chart 13.1-24 13.2-1 Proposed Training Schedule 13,2-7 13.3-1 Elapsed Exposure Timo to Reach 13.3-15 Specific Bone Dose Versus Downwind Distance (Based on Site Sultability Source Tenn) 13.3-2 Elapsed Exposure Time to Reach 13.3-16 Specific Lung Dose Versus Down-wind Distance (Based on Site Sultabilliy Source Term) 13.3-3 Elapsed Exposure Time to Reach 13.3-17
- Specific Thyroid Dose Versus Downwind Distance (Based on Site Sultability Source Term)
IOi 13.3-4 Elapsed Exposure Time to Reach 13.3-18 i Specific Whole Body Dose Versus l Downwind Distance (Based on Site Sultability Source Term) 13.3-5 Project Area - North Hal f 13.3-19 13.3-6 Project Area - South Half 13.3-20 13.5-1 Plant Procedures 13.5-5 l 13.7-12 13.7-1 CRBRP Security Organization l 61 i3.7-2 to 13.7-11 Proprietary 13.7-13 l i l l !O ( 13-Ix Amend. 61 Sept. 1981
CHAPTER 13.0 COPOUCT OF OPERATION 3 (d 13.1 ORGANIZATIONAL STRUCTURE OF THE APPLICANT Contract AT (49-18)-12 has been established to design, construct and operate a 61l Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. The parties to the contract are the Department of Energy (DOE), the Tennessae Valley Authority (TVA), the Commonwealth Edison Company (CE), and the Project Management Corporation (PMC). The organizational structure of the applicant (DOE, PNC3 and TVA) is covered in Section 1.4. TVA, as part of its lead role responsibility described in Section 1.4, will be responsible for the safe operation of the CRBRP. 13.1.1 P_tg.jggt Oroanization 13.1.1.1 Functions. Resoonsiblifties. and Authorftles of Protect Particloants The functions, resporsibilItles, and authorities of Project participants are described in Sections 1.4 and 1.4.2. The qualification requirements of Project participants are described in Section 1.4.4. 13.1.1.:: Anolicants' In-House OraanIzation This material is covered in Section 1.4.2. 13.1.1.3 Inter-Relationshins with Contractors and Sunollers Tt.Is material Is covered In Section 1.4.3. 13.1.1.4 Deoartment of Enerov (DOE) Particloation The participation of DOE in the Clinch River Breeder Reactor Plant (CRBRP) 61 Project is described In Section 1.4. In addition, DOE participates in R&D In support of the CRBRP Project through its LMFBR base technology programs described in Section 1.5. 13.1.1.5 Technical Sucoort for comrations TVA's Office of Power will be responsible for carrying out the operator role for the Clinch River Breeder Reactor Plant. Within the TVA Office of Power, the Division of Nuclear Power (NUC PR) will be responsible for the operation and ;nalntenance of the CRBRP. The TVA technical staf f supporting the operation of the CRBRP will consist of the Nuclear Central Offica (NCO) staff in Chattanooga and also support from other divisioni, and of fices within TVA (Section 13.1.1.5.1). in addition, technical suppsrt will be supplied by the CRBRP Project Office, the Lead Reactor Manufacturer of Westinghouse Advanced 61 Reactor Division (W-LRM),and by Burns and Roe (Section 13.1.1.5.2). [ Amend. 61 13.1-1 Sept. 1981
13.1.1.5.1 TVA's Technical Staff 13.1.1.5.1.1 DIvlslon of Nuclear Power (NUC PR) The Division cf Nuclear Power is responsible for the operation and maintenance of all TVA nuclear electric generating plants and will have this responsibility for the CRBRP. The Director of Nuclear Power is responsible for plant operations and maintenance, and determines the area of responsibility assigned to each branch in the division. Table 13.1-1 provides a summary of the number of personnel in the NCO technical support branches and staffs as well as their educational background and technical experience. The responsibilitles within the N(X) are provided as follows: The Assistant Director of Nuclear Power (Ocerations) is responsible for the overall operation of the nuclear generating plants and the nuclear training f acilities within the TVA power system. He is responsible for ensuring that the planning, organization, and control of division activities related to operation of the nuclear generating plants is adequate to provide plant safety. The Assistant Director (Operations) meets both the definition and qualifications of " Engineer in Charge" as set forth in " Selection and Training of Nuclear Power Plant Personnel," ANSl/ANS-3.1-1978. That portion of the NCO Staff to which the Assistant Director (Operations) provides direct cupervision, meets the " Staff Specialist" definition of ANS l/ANS-3.1-1976, and consists of the following: Nuclear Operations Coordinator Supervisor, Nuclear Operations Staff Supervisor, Preoperational Test Staf f Supervisor, Nuclear Security Staff Chief, Nuclear Training Branch The Assistant Director (Operations) wil1 direct the operation and maint. nance of the CRBRP; review operating and engineering data, regular and special I reports, test results, anti other information pertaining to the operation and l maintenance of the CRBRP; anc review cnd coordinate operating, maintenance, j and surveillance procedures to ensura that the CRBRP is operated to provide maximum safety along with achieving project operation and demonstration goals. He is responsible for providing input to the Office of Power for operational i 61 aspects of the plant design. O t ' Amend. 61 13.1-2 Sept. 1981 l
During Initial operation of the plant, the Assistant Director (Operations) assists in coordinating activities through the plant manager including (V_) pr eoperational, startup, and acceptance tests with the CRBRP Project Of fice, Nuclear Regulatory Commission (NRC), reactor manuf acturer, and the equipment suppliers. He will coordinate the activities of the CRBRP with other branches and divisions within TVA in such areas as onsite fuel management and waste disposal. The Assistant Director of Nuclear Power (Maintenance and Enalnearina Services) is responsible for the overall planning, organization, control, and implementation of division support activities related to nuclear generating plant maintenance, engineering, and outage management. The Assistant Diredor (Maintenance and Engineering Services) meets both the definition and qualifications of " Engineer in Charge" as set forth in
" Standards for Selection and Training of Personnel for Nuclear Power Plants,"
ANS I/ANS-3.1-1978. That portion of the NT Staf f to which the Assistant Director (Maintenance and Engineering Servicas) provides direct supervision, meets the "Staf f Special ist" def inition of ANSl/ANS-3.1-1978, and consists of the following: Chief, Controls and Test Branch Chief, Outage Management Branch Chief, Reactor Engineering Branch Chief, Nuclear Maintenance Branch s Supervisor, Design Review Staff Personnel within the above organizations provide technical support to nuclear facilities in accordance with the following: Controls and Test Branch The Controls and Test Branch develops engineering standards and provides a verlety of rrechanical, chemical, controls, instrumentation, environmental, and metallurgical engineering services for the division. The branch recommends desirable changes Indicated by engireering studies, furnishes technical assistance, and acts in an advisory capacity within the division on the more difficult systems engineering problems. Outaae Manacement Branch The Outage Managoment Branch is responsible for planning, scheduling, and implementing refueling outages, modifications, and forced unit or equipment outages. Other maintenance, tes^ lng, or inspections requiring expertise or resources beyond those contained within the normal plant staf f may be assigned 61 to the Outage Mancgement Branch. O Amend. 61 13.1-3 Sept. 1981 i _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ ._ )
Reactor Engineering Branch The Reactor Engineering Branch assures the adequacy of engineering plans and methods used in the operation of TVA's reactors and reactor-related systems. e The branch develops requirements for fuel accountability and has responsi-bilIty for TVA's low-level radioactive waste management program. The branch coordinates the division's safety analysis report review; performs operational safety analyses; and evaluates the adequacy of design and operation of safety-related systems. The Reactor Engineering Branch provides safety-related systems engineering expertise to the plants. Nuclear Maintenance Branch The Nuclear Maintenance Branch develops programs, standards, and procedures for the mainteniice of electrical and mechanical nuclear plant equipment. The branch provides technical assistance and guidance within the division on 61 diffIcuit maintenance engineering probiems. 13.1.1.5.1.2 Other TVA Oraanizations Other organizations within TVA which supplement the CRBRP and the Division of Nuclear Power Staff are as follows: Division of Power System Operations (PS0) Division of Transmission Planning and Engineering (TP&E) Division of Energy Demonstrations and Technology (ED&T) 61l Division of Engineering Design (EN DES) Division of Constri.ction (CONST) Division of Chemical Development (CHEM D) Division of Medical Services (MED SV) Division of Property and Services (P & SYS) 61l Division of Occupational Health and Safety (OC H&S) Office of Power Quality Assurance and Audit Staff (QA&A) A description of the duties of these organizations is given in Sections 1.4.2.4.1 and 1.4.2.4.2. 13.1.1.5.2 Project Technical Sucoort Project technical support for the operatico of the CRBRP will be provided to the Division of Nuclear Pover in TVA by W CRSRP Project OffIcw W-LRM, and Burns and Roe. Areas of support will be in accordance with the responsibilitles described in Section 1.4. 13.1-4 Amend. 61 O Sept. 1981
13.1.2 Ooeratiac Oraanizatlon 13.1.2.1 Plant Orcanizatlon The plant organizational chart is shown in Figure 13.1-1. The principal groups that function directly under the supervision of the Plant Manager and Assistant Plant Manager are the Plant Operations Section, the Plant Engineering Section, and the Plant Maintenance Sections (Mechanical, Electrical and !nstrument). Staf f services are provided by Management Services, a Quality Assurance staf f, and the Health Physics Unit of the Radiological Hygiene Brsnch. The latter is under the administrative 61 supervision of the Division of Occupational Health and Safety. The CRBRP organization follows the pattern developed through experience and used at all TVA fossil and nuclear generating plants. Plant employees are selected primarily from existing TVA conventional and nuclear plant staf fs and NUC PR's central of fice. Personnel quellfications 61 shall meet the criteria set forth in the ANSI /ANS-3.1-1978. 13.1.2.1J Plant Oneratlons Section The Plant Operations Section is responsible for all plant operations, it provides operating personnel to support the preoperational testing, fuel loading, startup testing, startup, and plant operation. It is responsible *~- coordinating and scheduling the training program for all operations persos. . It provides the nucleus of emergency teams such as the plent rescue and fire-fighting organizations. The Plant Operations Section is under the direction of the Operations Supervisor who holds a valid NRC Senior Reactor Operator (SRO) license. He is assisted by an Inline Assistant Operations Supervisor who also holds a valid NRC SRO license. Within the Plant Operations Section are five shift crews. The minimum shift crew will consist of the Shift Engineer who holds an NRC SRO license, one Assistant Shif t Engineer who holds an NRC SRO license, two Unit Operators who hold an NRC Reactor Operator (RO) license, one Shift Technical Advisor (STA), , and two Ass!stant Unit Operators. One Health Physics Technician will also be 61 assigned to each shif t. Additional operators are assigned as necessary. Plant management and technical support will be present or on call at all times. 13.1.7.1.2 Plant EnaineerIna Section l The Piant Engineering SectIon Is under the dIrectIan of the Eng'neering l Supervisor. He is assisted by a complement of engineers. The Plant Engineering Section is responsible for providing technical direction and staff assistance in the areas of nuclear and chemical engineering. ResponsibilItles 61{management, of thIs sectIon inelude waste plent and management equipment and chemistry control.performance tests, inplant f ue1 Amend. 61 13.1-5 Sept. 1981 l
. ~ , - - - - - . . - . - - - - . - - _ --.
The Plant Engineering Section carries out a comprehensive program of plant tests, studies, and investigations for 1he purpose of monitoring tile reactor, engineered safeguards, and plant operating conditions to assure compliance with the operating license and technical specifications and to improve the efficiency of the plant. This includes the coordination of the survelliance test pegram with other plant sections. 13.1.2.1.3 Plant Mechnical Maintenance Section The Plant Mechanical Maintenance Section is under the direction of the Mechanical Maintenance Supervisor. He is assisted by an iniine Assistant Maintenance Supervisor. The Plant Mechanical Maintenance Section is responsible for mechanical maintenance work and Inspections in the plant. This includes scheduling and conducting periodic inspections and tests on the systems assigned to this section associated with the reactor and engineered safeguards, as required by the technict.1 specifications and operating license. This section develops and carries out a p; eventive maintenance program that assures that the repair and replacement of parts are consistent with the intent of applicable codes and basic requirements of the original equipment. A record file is maintained by the section on all equipment, inservice tests, inspections, and maintenance 61 reports. 13.1.2.1.4 Plant Electrical Maintenance Section The Plant Electrical Maintencnce Section is unc%r the direction of the Electrical Maintenance Supervisor. He is assisted by an InlIne Assistant Malntenance SuperyIsor. The Plani Electrical Maintenance Section is responsible for electrical maintenance work and inspections in the plant. This includes schedu!Ing and conducting periodic Inspections and tests on the systems assigned to this section associated with the reactor and en3 neered
! safeguards, as required by the technical specifications and operating license. This section develops and carries out a preventive maintenance program that assures that the repair and replacement of parts are consistent with the Intent of applicable codes and ba,1c requirements of the original equipment. A record file is maintained by the section on all equipment, inservice tests, inspections, and maintenance
- 6) reports. ,
13.1.2.1.5 Plant Instrument Maintenance Section The Plant instrument Mainvenance Section is under the direction of the Instrument Maintenance Su>ervisor. He is assisted by two In!Ine Assistant Maintenance Supervisors. The Plant Instrument Maintenance Section is responsible for instrument maintenance work and inspections in the plant. This includes scheduling and conducting periodic Inspections and tests on the systems assigned to this 61 section associated with the reactor and engineered O Amend. 61 13.1-6 Sept. 1981
I safeguards, as required by the technical specifications and operating license. This section develops and carries out a preventive maintenance program that assures that the repair and replacement of parts are consistent with the :
. Intent of applicable codes and basic requirements of the original equipment.
l A record file is maintained by the section on all equipment, inservice tests,
- 61 Inspections, and maintenance reports.
Health Physics Unit 61] 13.1.2.1.6 The Health Physics Unit of-the Radiological Hygiene Branch is responsible for all health physics activities at the plant, it develops and applies radiation standards and procedures; reviews proposed methods of plant operation; participates in development of plant documents; and assists In the plant training program, providing specialized training in radiation protect!on. It conducts comprehensive onsite environmental radiation monitoring before, . during, and after plant startup and provides radiological health coverage for all operations including maintenance, fuel handling, waste disposal, and ] decontamination. It is responsible for personnel and inplant radiation monitoring and maintains continuing records of personnel exposures, plant l radiation, and contamination levels. This unit is under the administrative supervision of the Chief, Radiological
- Hygiens Branch in the Division of Occupational Health and Safety and under the 1 61 functional supervision of the Plant Manager.
J 13.1.2.1.7 Nuclear Plant Manaa== ant Services Section The purpose of this section is to enhance plant rollabiiIty, avalIabiiIty, and nuclear safety by providing the plant staf f with an integrated, automatic system for assisting the staf f in managing plant business, and to perform clerical services for the plant. It consists of three units: Document Control, Plant Services, and Administrative Services. Each unit is under direct supervision of a unit supervisor with the composite 61 reporting to the section supervisor. 1 61113.1.2.1.8 DIvlsion of Power System Operations (PS0) Enaineerina Unit The PS0 Engineering Unit, of the Division of Power System Operations, is 4 responsible for the maintenance and testing of the relaying associated with the transmission system. They are also responsible for maintenance of all i external come.unications systems at the plant (with the exception of the Bell Systems Equipment). They are responsibie for malntenance of portions of the onsite distribution and bus protection relaying. I This uait is under the administrative supervision of the Knoxvills Area Office in the' Division of Power System Operations and under the functional 61 supervision of the Plant Manager. O Amend. 61 13.1-7 Sept. 1981
,-..x .~...~.-,.m._.
.,.%,__,. . , ,,m _ . . , _ _ . , _ , _ _ _ , , _ _ _ _ __ _ , , _ , _ _ _ __ _, _ , ,
Nuclear Plant Oualltv Assurance Staff 6d13.1.2.1.9 The nuclear plant quality assurance staff is under the direction cf the Quality Assurance Staff Supervisor. The nuclear plant quality assurance staf f is responsible for developing, planning, initiating, and directing a comprehensive nuclear plant quality assurance / quality control program in the plant. Responsibilities include informing and advising other plant sections of the applicability, requirements, and implementation of the quality assurance program. The nuclear plant quality assurance staf f is responsible for coordinating, scheduling, and verifying surveillance monitoring and Inspections of safety-related structures, systems, and components. 13.1.2.2 Personnel Functions. Resoonsibilities. and Authorities During normal plant operations, the plant manager is responsible for all plant activities. In the event of absences, incapacitation of personnel, or other emergencies, tha following persons will be responsible in the ceder listed for all plant activities: Plant Manager Assistant Plant Manager Plant Operating Supervisor Plant Engineering Supervisor Shift Engineer 13.1.2.2.1 Plant Manager The Plant Manager has direct responsibility fcr all plant activities. He is l responsible for safeguarding the general public and plant employees from l hazards assortated with the operation of the CRBRP through implementation of the TVA hazard control standards and requirements, applicable DOE and NRC l rules and regulations, and plant procedures, and for adherence to all requirements of the operating license and technical specifications. He receives direction and supervision from the Assistant Director of Nuclear Power (Operations) and staff assistance from the Division of Nuclear Power 61 Central Office. l 13.1.2.2.2 Assistant Plant Manager The Assistant Plant Manager assists the Plant Manager in planning, coordinating, and directing the plant activities, in the absence of the Plant Manager, he is responsible for management of the plant activities. O Amend. 61 13.1-8 Sept. 1981
I. i l l l 13.1.2.2.3 Plant Ooerations Suoervisor i The Plant Operations Supervisor is responsible for the safe and ef ficient i operation of the plant in accordance with the operating license, technical specifications, and approved procedures and TVA hazard control standards and requirements. He is responsible for the preparation and maintenance of 4 611 up-to-date operating instructions and the preparation of operating records. He is also responsible for operator training programs and operating personnel schedules and is charged with the responsibility of keeping the Plant Manager fully informed in all matters of operating significance. 13.1.2.2.4 Assistant Plant Ooerations Suoervisor , The Assistant Plant Operations Supervisor assists the Plant Operations i Supervisor in reviewing, coordinating, and planning the activities of the plant Operations Section. in the absence of the Plant Operations Supervisor, 4 he assumes the responsibilities of that position. i 13.1.2.2.5 Plant Enalnearina Sunervisor 4 The Plant Engli.aering Supervisor serves as supervisor of the Plant Engineering Section and as a staf f engineer in providing engineering advice and assistance to the Plant Manager. He is responsible for initiating, planning, and
- coordinating the technical training programs. His experience and training must provide him with a good understanding of nuclear reactor technology, hazards, saf aguards, and licensing requirements and a knowledge of the control systems used in a nuclear plant. He is responsible for analysis of the perf ormance of the reactor and turbine cycle and associated equipmwent during the test, startup, and operation of the plant.
j i 13.1.2.2,6 Plant MacFanical Maintenance Suoervisor i The Plant Mechaalcal Maintenance Supervisor is responsible for all mechanical maintenance work and inspections in the plarJ. He is responsible for maintaining safe working conditions for his employees and for their adherence to saf e working practices. He is assisted in his work by an- Assistant Supervisor with experience in mechanical maintenance. He is also assisted by l foremen of the various craf ts within the section and engineers who will be )
- assigned to the plant as the workload demands. The Plant Mechanical Maintenance Supervisor must have a thorough knowledge of the operation and 61 maintenance of all plant mechanical equipment.
1 13,1.2.2.7 Assistant Plant Mechanical Maintenance Sunervisor The Assistant Plant Mechanical Maintenance Supervisor assists the Supervisor In planning, coordinating, and directing the maintenance work and inspection 61 in the plant. I l O Amend. 61 13.1-9 Sept. 1981
13.1.2.2.8 Plant Electrical Maintenance Suoerviser The Plant Electrical Maintenance Supervisor is responsible for all electrical maintenance work and inspections in the plant. He is responsible for maintaining safe working conditions for his employees and for their adherence to safe working practices. He is assisted in his work by an Assistant Supervisor with experience in electrical maintenance. He is also assisted by foremen of the e ectrical craf t within the section and engineers who will be assigned tc the plant as the workload demands. The Plant Electrical Maintenance Supervisor must have a thorough knowledge of the operation and 61l maintenance of alI plant electrIcel equipment. 13.1.2.2.9 Assistant Plant Electrical Maintenance Suoervisor The Assistant Plant Electrical Maintenance Supervisor assists the Supervisor in planning, coordinating, and directing the maintenance of work and 61 inspection in the plant. 13.1.2.2.10 Plant Instrument Maintenance Suoervisor The Plant instrument Maintenance Supervisor is responsible for all instrument maintenance work and inspections in the plant. He is responsible for maintaining safe working conditlens for his employees and for their adherence to safe working practices. He is assisted in his work by two Assistant Supervisors with experience in instrument Maintenance. He is also assisted by foremen of the Instrument Mechanics within the section and engineers who will be assigned to the plant as the workioad demands. The Plant instrument Maintenance Supervisor musi have a thorough knowledge of the operation and 61 maintenance of all plant instrumentation. 13.1.2.2.11 Assistant Plant Instrumentation Maintenance Suoervisors The two Assistant Plant Maintenance Supervisors--one an Instrument Specialist, the other an Instrument-Computer Specialist--assist the Supervisor in planning, coordinating, and d!recting the maintenance work and inspection in 61 the plant. 61l 13.1.2.2.12 Heelth Physicist The Health Physicist is the onsite supervisor of the Health rhysics Unit of the Radiological Hygiene Branch and !s responsible for direction of an adequate program of health physics surveillance for all plant operations involving potent,lal radiation hazards. He keeps the Plant Manager informed, 6t all times, of radiological conditions related to personnel exposure and potential contamination of site and environs. His duties include training and supervising health physics technicians; planning and scheduling monitoring and surveillance services; maintaining current data files on radiation and contamination levels; personnel exposures, and work restrictions; and ensuring that operations are carried out within the provisions of developed radiological hygiene and procedures. He provides monitoring assistance and technical advice to plant operations and provides assistance to the medical staf f in emergencies where radiation and contamination hazards are involved. O Amend. 61 13.1-10 Sept. 1981
61l13.1.2.2.13 Suoervisor. Nuclear Plant Ouality Assurance Staff The Nuclear Plant Quality Assurance Staf f Supervisor serves as supervisor of the nuclear plant quality assarance staf f and as a staf f advisor to the Piant Manager. He is responsible for advising the Plant Manager of unresolved quality assurance problems and trends significant to plant operation and safety. He is responsible for review and approval for plant procedures and instructions. He also advises the Plani Manager of f ailures of plant equipment to meet technical specification requirements and other nonconforming aspects of operations. He is responsible for the inplant quality assurance / quality control training programs. 61l 13.1.2.2.14 .Safgtv Enalneer The Plant Safety Engineer provides consultation to plant management on all fire safety matters; coordinates and evaluates testing, maintenance, and repair of all fire-related equipment and systems; conducts periodic safety and fire inspections to identify deficiencies and recommends corrective actions; conducts fire training and evaluates fire drills; provides on-the-scene advice to fire brigade leaders during fire emergencies as appIIcable. He reviews pre-fire plan and emergency planning documents and coordinates fire safety 61l matters as required with the Safety Staf f at the Central Office. 13.1.2.3 Shift Crew Comoosition Normal Ooerations The Shift Engineer on duty is in direct charge of the plant including startup, operation, and shutdown of the reactor and turbogenerators. He may institute immediate action in any given situation to eliminate difficulties or remove equipment from service to preclude violation of the operating license or technical specifications or to avert possible injury or undue radiation exposure of personnel. The Assistant Shift Engineer is under the immediate supervision of the Shift Engineer. He follows established procedures in doing his work. However, it a particular situation is not covered by a procedure, he may seek adv!ce from the Shift Engineer; or, if the situation is critical, he may use his own judgment to prevent damage to equipment, injury to personnel, or undue radiation exposure of personnel. He performs operations in the electrical switchyard, diesel generator building, and other areas inside and outside the main powerhouse structurm The Shift Technicsl Advl.;or (STA) is under the immediate supervision of the Shift Engineer. The STA serves in an advisory capacity to the Shift Engineer in matters involving engineering evaluation of day-to-day plant cporations from a safety point of view; accident assessment dedicated to concern for the safety of the plant; and in so doing has a clear independence from duties associated with commercial operation of the plant. He receives his technical 61 guidance from the Supervisor of the Plant Engineering Sectisa. n Amend. 61 13.1-11 Sept. 1981
The Unit Operator is under the immediate supervision of the Assistant Shift Engineer and the general supervision of the Shift Engineer. He follows established procedures in operating the plant. The Assistant Unit Operator is under the immediate supervision of the Unit Operator and the general supervision of the Assistant Shift Engineer. He follows established operating instructions in doing his work and does not deviate f rom those Instructions except as directed. He performs assigned routino inspections and manipulative operations without close supervision. He assists in the operation and performs work requirements within defined areas such as the Control Building, Reactor Containment Building, Reactor Service Building, Turbine Generator Building, Diesel Generator Building, intermediate Building, Steam Generator Building, and intake Structure. When on shif t, the Radiochemical Analyst is under the f unctional supervision of the Shift Engineer. These duties consist of periodic sampling of the various systems, such as feedwater and main steam, water makeup, waste condensate, and periodic monitoring of the primary and secondary sodium coolant. When on shif t, the Health Physics Technician is under the f unctional supervision of the Shift Engineer. He perfurms routine radiation surveys, personnel monitoring activities, and other assigned duties. He keeps the Shift Engineer informed of radiation hazards and performs special surveys as requested. 13.1.3 Oualification Recuirements for Nucle r Plant Personnel All personnel at the CRBRP will be required to obtain and maintain qualification standards equal to or better than those spectfled in i 61) ANSI /ANS-3.1-1978. The personneI seiectIon and tralnIng program that assures fulfillment of these qualification requirements also satisfles NRC Regulatory Guide 1.8. Specific minimum qualifications fcr alI those personnel discussed in Section 13.1.2 are given below. 13.1.3.1 Minimum Oualification Recutrements Plant Manaaer At the time of initial core loading or appointment to the active position in the licensed plant, the Plant Manager shall have ten years of responsible power plant experience of which a minimum of three years shall be nuclear experience. A maximum of four years of the rcrnaining seven years of exportence may be f ul filled by academic training on a one-for-one time basis. I This acade.wic training shall be in an englaeering or scientific field l generally associated with the production of power. The Plant Manager shall l have acquired the experlance and training normally required for examination by l NRC for an SRO license whether cr not the examination is taken. O Amend. 61 13.1-12 Sept. 1981 i L
i i If the Assistant Plant Manager meets the nuclear plant experience and NRC l examination requirements established for the Plant Manager,the requirements of O' the Plant Manager may be reduced so that only one of his ten years of experience need be nuclear plant experience, and he need not be eligible for NRC examination. The Plant Manager or the Assistant Plant Manager should have a recognized baccalaureate er higher degree in an engineering or scientific field generally associated with power production. Assistant Plant Manaaer At the time of Initial core loading or appointment to the active position in the licensed plant, the Assistant Plant Manager shall have a minimum of eight years of responsible power plant experience of which a minimum of three years shall be nuclear plant experience. A maximum of four years of the remain!ng five years of the power plant experience may be f ulfilled by satisf actorily completing academic or related technical training on a one-for-one time basis. A degree in science er engineering is desirable. He or the Plant Maneger shall be capable of fulfilling the requirements of an NRC SRO license whether or not the examination is taken. if the Plant Manager has the required three years of nuclear experience, the requirements of the Assistant Plant Manager may be reduced so that only one of his eight years of experience needs to be nuclear plant experience. Plant Ooerations Suoervisor () At the time of initial core loading or appointment to the active pos! tion in the licensed plant, the Plant Operations Supervisor shall hold an NRC SRO license and shall have a minimuu of eight years of responsible power plant experience, of which a minimum of three years shall be nuclear plant experience. A maximum of two years of the remaining f!ve years of power plant experience may be f ulfilled by satisf actory completion of academic or related technical training on a one-for-one time basis. The required nuclear experience for this position may be reduced to one year if the Assistant Plant Operations Supervisor has the required nuclear plant experience. Plant Enalneerina Suoervisor At the time of Initial core loading or appointment to the active position in the licensed plant, the Plant Engineering Supervisor shall~have a minimum of eight years of responsible power plant experience or applicable industrial oxperience of which two years shall be nuclear plant experience. He should huve an engineering or science degree. Plant Maintenance Suoervisors (Newhanical, Electrical. Instrn= ant _4 At the time of initial core loading or appointment to the active position in the licensed plant, the Plant Maintenance Supervisor shall have a minimum of seven years of responsible power plant experience or applicable industrial 61 experience, Ireluding at least one year of nuclear A U Amend. 61 13.1-13 Sept. 1981
plant experience. A maximum of two years of the remaining six years of power plant or industrial experience may be fulfilled by satisfactory completion of academic or ralated technical training on a one-for-one time basis. He further should have faciliarity with nondestructive testing and maintenance of sodium containing components, as well as with industrial maintenance or an understanding of the practical and theoritical aspects of electricity, or an understanding of industrial and nuclear instrumentation. Suoervisor. Nuclear Plant Ouality Assurance Staff At the time of initial core loading or appointment to the active position In the licensed plant, the Supervisor of the Nuclear Plant Quality Assurance Staff shall have seven years of responsible power plant experience or applIcabio Quality Assu ance experience of which a mini:um two years shalI be nuclear power plant experience. He shall be a graduato with a degree in engineering. A maximum of two years of the remaining five years of power plant or quality assurance experience may be fulfilled by satisfactory completion of academic or related training on a one-for-one time basis. If the Staff Supervisor has not had the quality assurance experience, he shall receive training from the Office of Power Quality Assurance and Audit Staff relative to basic quality assurance theory and practice. This training shall include an orientation to the Office of Power Quality Assurance Program as 61 defined by the Of fice of Power Quality Assurance Manual. The Safety Enalneer The Plant Safety Engineer shall have a sound understanding and thorough technical knowledge of safety and fire protection practicos, procedures, standards, and other codes relating to electrical utility operations. He shall: be ablu to read and understand engineering drawings; possess an analytical ability for problem solving and data analysis; be able to communicate welI both orally and in writing; be able to write investigative reports and prepare written procedures; have tho ability to secure the cooperation of management, anployees, and groups in the implementation of safety programs; and be able to conduct safety presentations for supervisors and employees. He shall have experience in safety engineering work at this level or have three years experience In safety and/or fire protection engineering. It is desirable that the Incumbent be a graduate of an accredited college or university with a degree in industrial, mechanical, electrical, or safety engincering or fire protection engineering. Haalth Physicist The plant Herith Physicist shall meet the qualifications as specified in NRC Regulatory Guide 1.8. Plant Ooerations Soction Fmoloyees At the time of !nitial cora loading or appointment to the active position in the licensed plant, the Assistant Plant Operations Supervisor shcIl have a minimum of six years of responsible power plant O Amend. 61 13.1-14 Sept. 1981
experience, of which a mininum of one year shall be nuclear plant experience. A maximum of three yet.rs of the remaining five years of power plant experience O may be fulfilled by satisf actory completion of academic or related technical training on a one-for-one time basis. At the time of initial core loading or appointment to the active position in the licensed plant, the Shift Engineers shall have fulfilled all the requirements of TVA's Division Nuclear Power Operator Training Programs for this position and have a high school diploma or equivalent and five years of responsible power plant experience, of which a minimum of two years shall be nuclear plant experience, six months at the plant where license is issued, and 61 he shalI hold an NRC SRO IIcanse. At the time of initial core loading or appointment to the active position in the lIcensed plant, the Assistant Shift Engineers shalI have fulfilled alI the requiremants of TVA's Division of Nuclear Power Operator Training Programs for this position and have a minimum of a high school diploma or equivalent and four years of responsible power plant experience, of which a minimum of two years shall be nuclear plant experience, six months at the plant where the 61 license is issued, and he shall hold an NRC SRO license. At the time of initial core loading or appointment to the position in the i licensed plant, the Shif t Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipi:ne and have received specific training in the response and analysis of the plant for transients and accidents. The Shift Technical Advisor shall also have received training in plant design and layout, including the capabilities of instrumentation and O 61 controls in the control room. At the time of initial core loading or appointment to the active position in the licensed plant, the Unit Operators shall have fulfilled all the requirements of TVA's Division of Nuclear Power Operator Training Progesns for this position, a high school diploma or equivalent and two years of nuclear power plant experience, of which a minimum of six months shall be at the plant where the license will be issued. The candidate shall successfully complete 61 the cold license or hot license programs and obtain an R0 license. At the time of init!al core loading or appointment to t'he cMive position in' the licensed plant, the Assistant Unit Operators working within the plant shall have a minimum of a high school diploma or equivalent, completed all requirements of TVA's Division cf Nuclear Power Operator Training Programs to this position, completed a basic nuclear course and plant systems course, and had several months of onsite plant f amiliarization. This position does not 61 require an NRC RO Iicense. Plant Enalneerina Section Fmolovees At the time of initial core loading or appointment to the active position in tha licansed plant, the Chemical Engineer shall have a bachelor's degree in science or engineering and a minimum of one year's experience in radio-chemistry. O Amend. 61 13.1-15 Sept. 1981
At the time of initial core loading or appointment to the active position in the licensed plant, the Reactor Engineer shall have a minimum of a bachelor's degree in engineering or the physical sciences and two years of experience in such areas as reactor physics, core measurcrnents, core heat transfer, and core physics testing program. At the time of initial core loading or appointment to the active position in the lIcensed plant, the Mechanical Engineer shalI have a minimum of a bachelor's degree in engineering or the physical sciences and four years of 61 experience in such areas as power plant performance, testing, and rollability. The Radiochemical Analysts and other engineering aides shalI be high school graduates with a m!nimum of two years' experience in their respective fields. At the tima of initial core loading er appointment to the active position in the licensed plant, the Nuclear Engineer shall have a minimum of a bachelor's 61 degree in engineering or the physical sciences and two years of experience. The-Shif t Technical Advisor (STA) recieves administrativo and technical guidance f rom the Engineering Section Supervisor, and serves as advisor to the Shift Engineer. Refer to Plant Goerations Section Emolovegg for STA 61 qualifications. Plant Ma!ntanance Section Emoloveet At the time of initial core loading or appointment to the active position in the licensed plant, the Assistant Maintenance Supervisor shall have a minimum of five yaars of responsible power plant experience or applicable industrial experience, including at least one year of nuclear plant experience. The position requires famillarity and understanding of electrical, or mechanical, or instrument maintenance. At the time of initial core loading or appointmen+ to the active position in the licensed plant, the Mechanical Engineer shall have a bachelor's degree in science or engineering and a minimum of one year's experience. l At the time of initial core loading or appointment to the active posit!on in the licensed plant, the Electical Engineer shall have a bachelor's degree in science or angineering and minimum of one year's experience. At the time of initial core loading or appointment to the acilve position in the licensed plant, the Instrument Engineer shall have a bachelor's degree in science or engineering and a minimum of one year's experience in the field of instrumentation. Six months of this experience shall be in nuclear instrumentaton and control. l Each Instrument Mechanic shall have a minimum of three years' experience in 61 his craft and shall be a skilled journeyman. O Amend. 61 13.1-16 Sept. 1981
t Each TVA craf tsman shall be a skilled journeyman. These experienced journeymen will prec'ominantly be transferees from other TVA generating plants O and installations. Craftsmen shall have a minimum of three years in one or more crafts. The primary source of new journeymen is the TVA apprenticeship program. This program, jointly administered by a TVA labor-management council, normally requires in excess of four years for completion. The program requires assignments designed so that he will develop skills equal to the recognized journeyman standard. Related classroom and correspondence lesson assignments provide the technical - aformation needed in the actual work being done on the job. 13.1.3.2 oualificat!ons of Plant Personnel The positions listed in 13.1.3.1 have not yet been filled. These positions w11I be filled as Indicated in Figure 13.2-1. i O O Amend. 61 13.1-17 Sept. 1981
--e, ,-------r - , , - - --r- - r m--- -r-,,r-----,.m--,- ----~~e-<,--,,-- r--,.-y y,-.-,------,,- -
w,a---
O O
~
i ; t i Table 13.1-1 TECHNICAL SUPPORT
SUMMARY
OPERATIONS 1 1 Nuclear Nuclear Operating Tetal Preoperational Operations hbclear Staff Tecleical Test Staff Coordinator Security 3rervier Ertwr innee s 1 1 1 4 7 Number of Personnel ,
- 1. Engineering (man years)
A. !bclear Power Field 7 15 5 49 76 D. Engineering Management 9 8 17 C. Total Utility Experience 7 15 6 78 106
- 2. Field (man years) j A. Reactor Physics B. I
- cactor Controls and Instru. 15 C. Chanical Engineering 15
] y w Electrical Engineering 5 5 D. Itchanical Engineering j w E.
,' F. Civil Digineering co G. Instrument Engineering I II. Other(s) Specify ;
Aero Space 2 2 Physics 3 3
- 3. Education DS NE
$N BS NE 4 Ng nS EE 1
i *a. M ac 1 1 ! .g BS Engg. Physics /Engg. Science 1
. "co - K3
}3 PhD Other(s) Specify 1 1 4 2 BS Physics (BS Sociology) individuals ABC SR OPR License i 61 i i
Table 13.1-1 (Cont'd.) - TECHNICAL SUPPORT
SUMMARY
CONTROL AhD TEST BRANCH Performanco Chenical Metallurgy Instrtncnt Tbtal Engineering Engineering and Standards and Controls Technical fhmt. Group Groun Creun Groun Bmerimee Nteber of Personnel 14 19 47 7 54 141
- 1. Engineering (man. years).
A. Nuclear Power Field 135 67 111 43 136 495 B. Engg. Managenent 43 27 6 9 85 C. Total Utility 167 87 177 54 138 623 Experience
- 2. Field (man years)
A. Reactor Physics 1 B. 1 Reactor Controls
- and Instru:nents 36 3 13 51 F C. Chanical Engg. 157 157
- D. Electrical Engg. 4 23 O*
E. Mechanical Engg. 52 75 20 23 5 27 175 P. Civil Engg. 4 G. 4 Instrument Engg. 16 64 92
!!. Other(s) Specify 31 1 20 41 100 Ca puter Anal. Metal Engg. & Instru. Ihgg.
Welding Engg.
- 3. Education
, BS NE 1 1 2 BS !E 5 14 3 2 BS EE 5 26 3 45 48 W@ DS ChE 3 29 35 3@c. DS Engg. Pirfsics/Engg. Science I4S 3
' 9 1 3 16
~ PhD 4 4 Other(s) Specify 3 4 13 , 3 1 31 Asno. degree SS Qian Fetal Engg. BS Math 61 1 BSB 1 CE 1 BS Cmn e G G
O O O i Table 13.1-1 (Cont'd. ) TECHNICAL SUPPORT
SUMMARY
OUTAGE MANAGEMENT BRANCH Long-Range Modifi- Short Range Outage Total Fanage- Planning cation Planning SLyrt Technical ment Section Seution %crion SecH nn Prnerience Number of Personnel 7 5 3 4 5 24
- l. Engineering (man years)
A. Nuclear Power Field 49 33 12 24 17 135 B. Engg. Management 24 17 2 15 10 68 C. Total Utility Experience 53 4 9.5 23 23.5 113
- 2. Field (man years)
A. Reactor Physic.s 3 2.5 5.5 B. Reactor Controls &
~ Instrtreent 2 3 3.5 8.5 ta C. Chcnical thgg.
D. Electrical Engg. 3 7 1 11 22 h E. Mechanical Engg. 26.5 11 13 31 3.5 85 o F. Civil Ehgg. G. Instru:nent Engg. 20 12 32 H. Other(s) SR eify Project Control 5.6 Licencing 5.2 1 2 4 18 Materials Fast Flux Test Control Rocun Facility Operator Design ,
- 3. Education ES lE BS lE 2 1 1 1 3 8 DS EE 2 1 1 3 m M OE 1 1 Q ,g BS Eng. Physics /Engg.
r+ s Science 1 1 2 F MS 1 1 5 o, PhD
$~ Other(s) Specify BS Ehg. Tech BS Nath BS Industrial Assoc. -Elec. 7 Industrial Management Technology 61 M n gonent BS Math /Chan/Canp Science BS Physics
Table 13.1-1 (Cont'd. ) TECHNICAL SUPPORT
SUMMARY
REACTOR ENGINEERING BRANCH Reactor Ios Icvel Reactor Nuclear Radiological Systens Radwaste Analysis Safety Dnergency Total Mc-+ - Groun Manag e nt Groun Staff Plannin<T St aff Ernerience inzrber of Personnel 13 7 3 17 1 1 42
- 1. Engineering (ann. years)
A. Ibclear Ibwer Field 115.5 48 11 38.5 5 3 221 B. Engg. Pqnt. 28 0 1 4.5 33.5 C. Tatal Utility Experience 108.5 24.5 7 32.5 5 3 177.5
- 2. Field (man years)
A. Reactor thysics 53 19.25 2.5 74.75 D. Reactor Controls
- & Inctru. 2 5 7 F C. Clxsnical Engg. 5 5 3 13 7
N D. Electrical Ergy. 0.5 0.5 g E. Pechanical Engg. . 3.75 4.5 12.25 F. Civil Engg. G. Instrunent Engg. 0.5 6.5 1 8 II. Other(s) Specify 32.5 (NE) 11 (NE) 8 30(NE) 3(NE)- 98.5 4 Navy 10 Radwaste Mgnt.
- 3. Education tog BS tE 8 6 1 11 1 27
-@ m DS !E 1 1 2 Pg ME 1 1 BS ChE 1 1 2 5m M~
ES Engg. Physics /
~ Engg. Science 2 1 1 2 6 11 S 3 1 1 6 11 PhD Other(s) Specify 1(BS @an) 1(BS Physics) 4 61 1(BS thc. Sci. & W .)
1(AEME) e O O
i O O O 2 , Table 13.1-1 (Cont'd.) ' TECHNICAL SUPPORT
SUMMARY
NilCLEAR MAINTENANCE BRANCH Stationary Rotating Electrical Special Total , Dpipnent Pquipnent Djuipnent Projects Technical Manac_omprit Groun_ Groun_ Grmn Staff Pwnarience ) tbnbar of Personnel 2.0 30.0 11.0 18.0 11.0 72.0
- l. Ehgineering (man years)
A. Nuclear Power Field 18.0 162.5 22.0 68.5 44.5 i B- 315.5 Fogg. Management 6.0 600 6.5 4.0 1.0 77.5 , C. Total Utility j J Experience 20.0 140.0 69.0 108.5 23.5 361.0 C 2. Field (man years) l A. Reactor Physics 1.0 t e B. Reactor Controls 1.0 S$ & Instru. 5.0 5.0 C. Chanical Dvgineering ] D. Electrical Dyjg. 9.0 10.5 100.0 38.5 158.0 1 E. Mechanical Dxjg. 11.0 182.0 32.5 17.5 243.0 j F. Civil Engineering 39.5 1.0 43.5 } G. Instrument Engg. 5.0 5.0 II. Othor(s) Specify 52.5 33.0 8.0 1.5 135.5 mg Nuc Const (6.5) Engg. Aide (7) Elect. (3) Draftenen (5) O ro Craft (3) Plt. Maint. (6) Elect. Tech.(4) j FR - IE (3) Turb Maint. (20) l OA (3)
; I$cn Electronics'(16).
$" Res. Engr. (3) 61 Des. Fagr. (15)
Electronics (3) 4 i
Table 13.1-1 (Cont'd. ) TECHNICAL SUPPORT
SUMMARY
NUCLEAR MAINTENANCE BRANCH Stationary Rotating Electrical Special Total RI uipnent Dpipnent Dpipacnt Projects Technical Manaa m t Crotm Groun Creco Staff Bmerimee
- 3. Education BS NE 1.0 1.0 BS II 9.0 6.0 15.0 BS EE 1.0 14.0 15.0 DS Che DS Ihgg. Physics /
Engg. Science 1.0 3.0 4.0 ftS 1.0 1.0 1.0(Dy. Sci.) 3.0 Fic Other(s) Specify 17.0 5.0 4.0 2.0 28.0 BS-Arch. Engg. BSIE(1) AC' (2) BS Math (2) - CE(1) ASME Tech (1) BEE (1) F Physics (l) ASPE (1) PPMS (1) 7 Eng. Sci.(2) Navy (20) y ASFZ (4) ICS PE (1) BS CE (2) FEA (1) PE (2) Asso. Nuc. Ttch. 61 Tech School-2 yr. (2) BS Ind. Tech. (1)
$N R8 P
O O O
I i ( PLANT MWs4R I I Quality Assurance Ass't Plant Supervisor (1) Manager (Il r oA i sare,. l EngWrlag (7) I l 5taf f (4) i Plant Service E btaff (3) {h NUC Pit Central Office i i
- t. - irui a
.....a...
'** [ncluttales Operatl "NUC .'A
{ ..........,........... Per ggl gig OP;Ti$v itM] sortavis0R supervis l l 10 (NGINttt st i ! CitRK b L.. .U*l5ES...l !
- e. .. ....Jlsi.vist ansl (Am. Rsserelces)
! Assistant Operettons tupervisor r
l l . - -
. Cheetcal Reactor Mrchanical Engineer (1) Engineer (t) tagineer (1) 3,,,,
Engtseers Chemical %uclear thenanical
~
(ngineers (*; tagineers (I) tagin'eFS III Ass htant
$hlft themicay Engineering Engirrering angineers E ngi n,,, - Aide Alde -
trainees (2) ( 54M) III (Statistics) (g} Unit Cheetcal EnginMr ~ Operators Engineer Associate (gg Associate (1) Assistant Cheetcal Engineering Unit taMeatory Aldes Operators Analysts (6) I8I - I i
1 i s 1) I (D"O% ices _mnse ne a o-. ;
~
S!'J'S lc.l 5.,vic.s P50 l Plant Security a=-t weith Physin~ Servics (1) h.lD* (I) thy (1) ser $! f t1 I_ Of fic? St ff (20) d, (4)) I 11$1 i . t ul l Nchanical Elec trical Instrument y- Maintenance Malatenance Malatenance ll) Supervisor (1) Supervisor (I) Superviser (1) F' (14 rt g gl,rg l l 6 ( Adn. Sys.) i ( m , 5,g.) i 1 Assistant Assistant ~ flectrical issistant last rume nt Supervisor supervisor (ngineer J(p (I) (l) (II Supervisors (?) I"9'"" g,3 mhanical g angtneering Inglwr l
- i Aldes l II' Crafts flectrical [ng*r, te trument Clers
[ ~ Foresen Forearn
~
Alge teresen (m. Sus.) $ (4) 110fE5: (l) (I) III ., i Mministrative 5ctrillon [ g l
- Shif t Technical Engineer (tert Sr. Instr. t
_ Craftsarn p,,,,,nsel err included la ~ (* I'S*I *Ch*"i" g (M)g the Nuclear ingime Group IIcctricians (10
] (12l p
~' '
** No; included in the I~ '
g,,,, Plant Perwearnt 5taff. Jr. Instr.
- f gggg
~- gg) behanics y 8 (8) 3:) FIGURE 13.1-1 CRBRP Organization Chart (1 year after power operation) 6717-1 13.1-24 i i Amend. 61 i Sept. 1981 l u? ; e o _ . _ _ _ _ _ _ _ _ _ _ J
13.2 _IPAINING PROGRAM G 13.2.1 Procram Descriotion Yho basic objectives of the training program are:
- a. To assure that all plant personnel are properly trained and qualified to perform their assigned tasks in a safe and efficient manner.
- b. To assure that the CRBRP is operated in accordance with NRC regulatory requirements and guidelines.
- c. To assure that all training is formally documented.
- d. To meet or exceed NRC lIcensing requirements.
In achieving these objectives, Individual training needs are established by comparing job requirements with Individual experience. The training program, as it is initially constructed, is approved by the 61l Assistant Director of Nuclear Power (Operations), af ter being approved by the Plant Manager. This ensures that the content and the Intent of the training program provide the necessary training for personnel associated with reactor operations. The program is designed to train personnel both with and without previous nuclear experience. T The effectiveness of the training program is evaluated by the performence of employees on TVA and NRC examinations in carrying out rheir assigned duties. In addition, periodic audits of the training program are performed by designees within the Office of Power, but outside the Divisicn of Nuclear 61l Power. 13.2.1.1 Procram Content At the time of manning the CRBRP, TVA should have highly trained nuclear plant
- operating personnel at the Browns Ferry, Sequoyah, and Watts Bar Nuclear Plants. These plants will be the primary source of personnel for the CRBRP.
Those positions at the CRBRP which require an NRC lIcensed SRO shalI be filled with personnel who have or are eligibie to sit for an NRC SRO license on a commercial size light-water reactor. Other positions shall be filled with personnel within the TVA organization as available and selected from competent applicants from outside. AlI CRBRP personnel wilI be given comprehensive l training to produce personnel who have that combination of educetion, ! experience, and skills commensurate with their level of responsibility which provides reasonable assurance that scisions and actions during all normal and off-normal conditions will be such that the plant is operated in a safe and offIcient manner. l I C Amend, 61 13.2-1 Sept. ]S81
l The TVA student operator training program and replacement training at operating TVA nuclear plants shall ensure no loss of operator efficiency at ! those plants becaust. of transfer of personnel to the CRBRP. Individual I training needs shalI be established by carefully examining the individual's experience and previous training and comparing these with the job require-ments. The formal program to be provided for candidates seeking NRC R0 license or NRC SR0 license as well as the length of each aspect of the program j is discussed in the following paragraph and depicted in Figure 13.2-1. The l training program for a student who has not had nuclear experience shal! 61 consist of the following phases: ' 61l Nuclear Physics Plant Tecnnology and Specialist Training Reactor Operations (LWR and FBR) Actual Fast Reactor Training 61l Onsite Work-Study Program and Simulator Training Nuclear System Special Training to include Sodium Technology A final training plen based on the needs of the individual staf f members will be prepared after personnel selection and shall be included in the Final Safety Analysis Report. . 13.2.1.2 Coordinarfon with Preocerational Tests and Fuel Loadina Figure 13.2-1 presents a proposed training schedule for the CRBRP which 61l satisfies the requirements of ANSI /ANS-3.1-1978. It is planned that the j following personnel shall be licensed in accordance with the requirertents of 10CFR55 befors initial fuel loading: Operations Supervisor, at least five l Shif t Engineers, an 1 at least five Assistant Shif t Engineers. The Plant 1 Manager or the Assistant Plant Manager shall obtain the treining required for i an SRO license. It is planned to obtain R0 licenses for at least five Unit l Operators during startup testing of the plant. The various phases of training available are outlined below, along with descriptions of personnel participa- i tion in each phase. 13.2.1.3 Practical Reactor Ooeration Practical training at TVA's Browns Ferry, S%uoyah, and Watts Bar and at an operating sodium-cooled fast reactor is anticipated in order to provide the experience required for applicants for cold licenses. The persons who will initially obtain SRO IIcenses shalI participate in this training. Training ! requirements shall be individually determined and training will be supplied to fili these needs. The program involving the detual participation in the operation of a sodi'.n- l coolod fast reactor to gain experience in the areas of liquid metal systems and the characteristics and performance of fast reactors O Amend. 61 13.2-2 Sept. 1981 ;
shalI be Integrated into this phase of the total training program for iIcensable personnel and management personnel. 13.2.1.4 Reactor Simulat f or: Training A simulator for the CRdRP to be located onsite and operational at the time of commencing preoperational testing to serve as a mode for procedure checkout as well as a vital component of the operato.- training prog. am will be available. The training program for all candidates seeking NRC SRO and R0 licenses will 61 include significant time at the simulator. 13.2.1.5 Previous Nuclear Training Figure 13.2-1 presents the tentative training schodule showing the relation of the training to the plant schedule for construction, testing, and operation. The actuel training schedule wilI be dependent upon the background and experience of the Individuals chosen for positions requiring a cold license. Hence, the schedule is tentative and subject to change before submission as a part of the Final Safety Analysis Report, 13.2.1.6 Other Scheduled Trainina 61l _ Nuclear Physics The nuclear courses for operators will consist of basic atomic and nuclear physics; nuclear reactor principles, including neutron and reactor physics, reactor kinetics, reactor control, reactor instrumentation, and reactor materials, with special reference to f ast reactors; reactor core thermal-O3 hydraulic characteristics, such as hot channel factor, sodium boiling and voiding, linear heat rate; and radiation protection and radiation safety. In addition, the course will include work on sodium technology. Other personnel whose duties require basic nuclear training, such as the Chemical Technicians and Instrument Mechanics, wilI receive more abbreviated instruction in nuclear fundamentals as part of their on-site specialist training. 61l The prerequisite qualifications for participation in the nuclear physics and succeeding phases of the operatcc training program are:
- a. High school education or equivalent.
- b. Knowledge of mathematics through high school algebra.
- c. Satisfactory completion of medical examination.
61
- d. Satisfactory performance of the work in his present classification.
l y Amend. 61 13.2-3 Sept. 1981
Plant Technoloav and Soecialist Tra b ag A design lecture series covering the function, design and operation of nuclear systems and components shall be conducted at the plant site. The persons ; requiring NRC SRO licenses shall participate in this plant technology training. In addition, the plant management end engineers shall elso participate. Various specialist training, consisting of work-study assignments, shall be conducted for plant engineers, technicians, and maintenar.ce personnel. This training shalI be specifIcally talIored to the Individual's needs. It is planned to use the Browns Forry, Sequoyah, and Watts Car facilities for as much of this training as feasiole. Specialist training in LMFBR technology is 61l Planned for the Supervisors, the Nuclear Engineers, and certain Central Office staff engineers. Specialist training is planned to varying degrees in instrumentation and controls ir.cluding process computer programming and maintenance for the Instrument Engineer, several Instrument Mechanics, Plant Engineering Supervisor, and certain Central Office staf f engineers. Training assignments at TVA nuclear plants of varying lengths are plannef for plant staf f personnel such as the Instrument Engineers, Reactor Engineers, 61 Mechanical Engineers, Electrical Engineers, and Chemical Engineers. After completion of this specialist training, the approprite personnel shall organize and conduct necessary specialist training onsite for the Chemical Technicians, Maintenance personnel, Instrument Mechanics, and others. Onsite Work-Studv Proarem This phase, which begins before fuel loading., Integrates personnel into their plant assignraentre !t is conducted under the direction of TVA supervisory personnel. Darlag this period, plant personnel participate in the preparation of procedures and manuals, preoperational testing, preoperational checkout of the operating procedures (the simulator will be used to verify the adequacy of the procedures), initial fuel loading, and initial startup program. The applicants for NRC RO and NRC SRO Iicenses w11I participate in further training and examination preparation related to obtaining the required NRC license. This will include Simulator Control Room experience with selected mal functicn exercises. All plant personnel will participate in a plant 61 indoctrination and radiation-protection course. Fire Brigade Training Although Fire Brigade Training is not a prerequisite to sitting for an NRC Operator License exam, it is included as a portion of the operator, assistant shift engineer, shift engineer, assi: tant operations supervisor, and operations supervisor training. The objective o; this training is to ensure that plant fire brigado members, leaders, and other plant mployees perf orming f ire-related f unctions receive comprehensive fIrst ald fIrefIghting instructions and application that w11I be instrumental in the prevention, control, and suppression of plant fires. This training will be updated or revised as necessary to ensure that current, acceptable practices are included. O Amend. 61 13.2-4 Sept. 1981
.c 13.2.1.7 Trainina Proar=== for Non-Licensed Personnel O '
TVA, on a continuing basis, plans and administers training programs for the professional and managerial development of its employees. Relationships are maintained with both local and state educational Institutions as well as with the vendors of vr.rlous items of equipment. Advantage is taken of appropriate seminars, specialized courses, and training activities offered by these groups ( to keep employees abreast of new develop.nents in power production and safety. 2 13.2.1.8 General Fmnlevee Training Personnel with specific duties and responsibilities in the plant she? l receive Instruction in the performance of these duties and responsibilities. All pc sons having unescorted access to the plant areas shall have completed either (1) Intensive nuclear training, which will include radiation protection techniques and the site emergency plan, or (2) a brief plant Indoctrination-cnd radiation protection course which will Include discussion of plant organi-zation and layout, controlled zones, ; adiation and contamination hazards, exposure Iimits and controls, elementary health physics, and pertinent sections of the site emergency plan. When persons wW have not completed either (1) or (2) above enter the plant areas, they will be escorted by an employee who has received training in radiation protection and plant emergency procedures. AlI permanent p! ant personnel shalI receive training perlodically in the e plant's fire protect!on policies. Temporary personnel who have responsibilities in fire protection will also be included. Training and Indoctrination relating to quality essurance will be provided to all employees as applicable in conformance to RDT-F2-2, Section 7.3.2. The training program for all regularly employed persons shall comply witt-l 61 Section 5.4 of ANSI /ANS-3.1-1978. Periodic retraining of plant personnel regarding radiation hazards with emphasis on individual actlons, will be conducted sit monthly meetings attended by all available plant personnel, and at short wewKly meetings held within the
- various plant groups. The emergency plan will be discussed at least once j annually in these meetings.
13.2.1.9 Resnonsibic Individual The individual responsible for conducting ac.d adninistration of the nuclear power plant training program is the Assistant Plant Manager. The plant
~
l Quality Assurance Staf f Supervisor shall be resDonsible for develonIng and directing the Nuclear Plant Quality Assurance Program which compi ses with RDT-F2-2. l i iO i A'nend. 61 l 13.2-5 Sept. 1981 l
13.2. Retrainino Proatra This information will be included in the FSAR. 13.2.3 Reolacement Trainina This information will be included in the FSAR. 13.2.4 Encords 13.2.4.1 11b Of ficial records of employeo qualifications experience, training, and retraining of each member of the plant organization are maintained in the of ficiel TVA Personal History Record (PHR) by the Division of Personnel. The PHR providas in a standardized arrangement, the information oiticially recognized in recording and supporting employee status. The PHR is maintained in current and cccurate status and is controlled as to availability. The material admitted to this record is restricted to items for which authecticity has been confirmed through established procedures; e.g., of ficial TVA forms, signed statements f rom the employee, management representatives, etc. 13.2.4.2 Plant Reucrds supporting cequests for NRC SRO and NRC R0 licenses are maintained in the pl ant master f ile. These records include training courses attended, retraining classes, number of reactor startups, and other Information necessary to ensure that training requirements have been met. Some of these f records are duplicated in the PHR. A training file for each member of the plant organization is maintained in the plant master f ile. Information regarding participation in training and retraining activities and records of employee participation in training activities leading to promotion to a higher level of competence will be maintained in this training file. P O Amend. 61 13.2-6 Sept. 1981
1 H014THS FRIOR TU CRITILA'.lTV
~ _ . .
PLANT MANAGER ASSISTANT PLANT MANAGER OPutATIONS SUPERVISOR ASWENT OPERATIONS St'PERY13OR SHFT ENGINEER ASSISTANT SHIFT ENGINEER
- UNIT OPERATOR AS$1STANT UNIT OPERATOR EN3NEENNG SUPERVISOR O CEMCAL ENGNEER O NUCLEAR ENGINEER _
C Mechanical Fnginone O SHFT TECHNICAL ADVISOR STARTUP CREW FROM CEMTRAL OFFICE CEMICAL ENGNEER NUCLEAR DJGNEER EQtANICAL ENGINEER MANTENANCE SUPERVISOR (ELECTRICAL) . O ASSISTANT SUPERVISOR MAINTENANCE SUPERv:SOR (MECHANICAL O ASSIST;u.T SUPERVLSOR MAINTENANCE SUPERVISOR (INSTRUMENT) C ASSISTAliT SUPERVISOR (INSTRlkdNTATIO OASSISTANT SUPERV150R (INST.-COMPUTER OINSTRLMENT ENGINEER QUALITY ASSURANCE SUPERVISOR QUALITY ASSURANCE ENGINEER PREOPERATIONAL TEST TEAM TEST COORDINATOR TEST DIRECTOR TEST ENGINEER SAFETY ENGINEER INDUSTRIAL ENGINEER I- N!aYe tU $ Etwo sanuu cootto rasi mac-
- ta",*4*c,Ja.i'TJ7a=AT"""
E: n=0 lW",A'Zffi"a^KMPivT,E1 r L*,,Ju???% RISnF,'aA'""#, - J:ln?"MJ ?#u"v'##' R%'77 AAE rlLLED WITH TVA PUtSONPEL f L
i i
\
LWA (82) ONSTRUCT!0N PERMIT (68) PRE OP TESTING (25) START Na LOAD (11) VA t%NAGERS STAFF ONSITE (62) f CRITICALITY (0) i i tA lB C i lJ l L i ! tA Bl C ' lJ l L l tA C 'O l L l B f , t A Bt C ' D I L l jF ' l E B' C D I L
- !F l H ll Bt C L D ! L t ll F G lH BT C i D j L
- i F tC ! !J l L l l l' A lB C l lJ j L I I l E '0] C i lJ l L i
E *B C ! lJ j L
~l '
~ i E 'R C ! .11 L i
- lE 'Bl C I !J i L i l ! !
i l l ! K M N
! ! ! l K M N j l ! K M N l i 'A l0 C ! 'J.
e L i l j B1 C I lJ ! L l 3 Al B C Jl _ _ _ L I l Bi C I !J ! L l lA B C 4e L
- 1) ; l j B' C lJ L l f ,B' C 'J L E{ _ fB ,
C 'J ' , L
! 3 l A lB C ! lJ ! L h ll E
Bl C l TJ ! L
. t l l l P Q N P Q N l l N
j l l lP Q L { ; AC L i AC L H - AS$sGtMNT AS AS$r$ TANT COm(TROL OtTRATOR Ar TVA Nuf. JAR PLANT (PRACTCAL Rf ACTOR OfY. RATION) OR I - A$sGM.NT AS CONTROL OttRATOR AT TWA NtRIAlt FUNT (PRACTICAL REACte OPERATION) FIE PREF 9R% TION, ETC J - f40lt.NLOAD , FUEL t.DADING ANO STARTLP YtSTS-PROG [URE PRff% RATION MLLCONTHLE FROM ITEM *C' It - AS50aeff De SITE FOR 51ARTUP FAMILIADR ATON (PL Atti TEONCLOGY Arc $PECW. TRA#TeGI O L - OttRATIP8 OURr4 DEMONSTRAflON ftRIOD lAR PLANT M- AsspeCNT As STARTUP CREW ASSISTNG OPERAfl0N WITH A$9CNED OUTIES N - RETuRio To CtNTRAL OFFICE
) GRAM $ ASSOf(NT TO CRDRP PROICT n Q NTIAL TVA EMPLDrMO(T TO FILL TtESE MMilflDNS OR REfW TVA FY>tSONFEL MO WILL FILL TDESE FO$tTIONs MSE POSff 0f3 P- Assigned onsite for fai'illiarization arid preparation E f Test Procedures 1- Perfoming pre-operatlOnal Tests FIGtJRE 13.2-1 Proposed Training Schedule T
2 Amend. 61 13.2-7 Sept. 1981 ( i _ _ _ _ . _ _ _ _ _ _ _ _ __ i
. - . . . - . . . ~ . _ _ .
13.4 REVIEW AM) AUDIT O 13.4.1 Revlaw and Audit - Construction The review and audit function during plant design and construction will be accomplished as an integral part of the quality assurance program described in Chapter 17. 13.4.2 Review and Audit - Test and Ooeration l This will be included in the FSAR. i
- [
Amend. 61 13.4-1 Sept. 1981
I - 13.5 PLANT PROCEDURES 13.5.1 General Day-to-day operations will be carried out by the various plant sections. Each section, within its assigned area of responsibility will operate with an optimum degree of independence, but with an adoquate degree of supervision to assure suf fIclent coordination between sections io best achieve the common purpose. Standard practices will be iss"ed governing employees' actions and establish-Ing standards for plant operation. These Instructions will contain adminis-trative restrictions and plant requirements in conformance with ANSI-N18.7 established to ensure safe operation of the plant in accordance with the-provisions of the f acility licenses, technical specifications, and TVA hazard control standards and requirements. They will provide that plant activities be conducted in a manner to protect the general public, plent personnel, and equipment. A formalized system of written instructions confor:ning to the requirements of the Operational Quality Assurance Program Plan will be employed in support of the standard practices. Figure 13.5-1 shows the organizational structure of these instructions, instructions covering all plant operations, maintenance work, tests and equipment changes, will be written by ihe plant operations staf f and/or the technical support staf f and will be put into ef fect only af ter review and approval by specialists in appropriate disciplines and written authorization of the Plant Manager. It is his responsibility to O ensure that required reviews and approvals are completed before authorizations are issued. The Plant Operations Review Committee composed of the Plant Manager, Assistant Plant Manager, Section Supervisors, and the Health Physicisi, will initiate and review all proposed changes to plant instructions. On the basis of the recommendations received from this group, the Plant Manager will determine whether further review is required before approving a change. There will be in addition to planned changes in the plant and instructions, the area of accidental or gradual changes in plant equipment charactei-Istics ! or conditions. Each supervisor and employee wilI have the responsibility to l be continually alert for such changes and for reporting them upon detection. I The periodic inspection of plant equipment and the continuing review and analysis of operating data from plant logs, instruments, and tests will provide regular sources of Information on plant conditions. 13.5.2. Normal Ooeration Instructions y Instrudions will be prepared for integrated plant operations, system operation, and instrument operation. The instructions for integrated plant operation wi:1 outline the principal steps required for startup and shutdown of the reactor, turbine-generator, and supporting auxillaries as an integral unit. O Amend. 61 13.5-1 l Sept. 1981
The systm Instructions will c.ontain requirements for startup, operation, shutdown, and anticipated abnormalitlei of the system concerned. The Instrumentation instructions will be similar to systern proci.dures. They wilI cover the normal checkout, operation, and calibration of the nuclear and process control and monitoring systems. Instructions will be supplied for all operational systems and instrumentation which have a significant bearing on nuclear safety. RefueiIno Instructions Detailed ref ueling instructions will be used to ensure a safe and orderly refueling. The instructions will specify or make reference to other system operation documents that specify periodic shutdown margin checks, fuel-handling techniques, and other precautionary steps to assure that the facility license and technical specifications are not violated. When fuel is being Inserted, removed, or rearranged in the core or when control rods are being installed, removed, or manipulated, licensed operators will be in the control room and in the refueling area supervising the operations. Technical personnel wilI provide guidance where necessary and will verify that all fuel has the proper orientation and is in the correct location. An essential part of the plant nuclear materials control and of refueling outage requirements is to have complete knowledge of the Identity, location, composition, and condition of all fuel and other core components. 13.5.3 Abnormal Ooeratina Instructions Abnormal oporating instructions will be prepared for abnormal operation of systems or equipment. Instances where operation of systems or equipment under abnormal conditions could af fect nuclear safety will be given, and the Iimiting conditicas for continued operation as set forth in the technical specifications w11I be iIsted, in addition, indications of the abnormality, possible causes, automatic actions that may occur, and subsequent operator actions will be given. 13.5.4 Emeroency Goeratino Instructions Emergency operating Instructions will be written for conditions which may possibly lead to injury to plant personnel or the public or to the release of radioactivity in excess of established c= rating iimits. These instructions will contain information describing the 9nt or conditions, probable indications, automatic actions which oct. rnediato operator action, and any subsequent operator action necessary to correct or control the situation. The primary responsibility for initiating the corrective action will rest upon the operator who first becomes aware of the situation. He will notify his supervisor of the existir.g cor.dition and the action he has O Amend. 61 13.5-2 Sept. 1981
taken. All operating personnel through training and experience will have learned to recognize and evaluate Impending f ailures or malfunci!ons and to O. initiate proper corrective actions. The emergency instructions will be used to train the operating personnel and make them aware of the accidents or situations that could occur and the proper course of action. Eautoment Clearance Instructions The clearance Instruction is the method use.d by TVA for protection of workmen, the public, and equipment, whether it be electrical circuits, hydraulic equipment, mechanical equipment or other devices. No work on such equipment will be performed except under the applicable clearance instruction. The Shift Engineer will be responsible for the tagging of equipment within his area of responsibility. The TVA load dispatcher will issue tagging instructions for the high-voltage switchyard including the transformers. The Health Physics Unit will be responsible for determining the existing radiation hazards. Clearances will be issued only to those persons whose names appear on official clearance lists. A clearance will be established by the use of colored protective cards placed to indicate the boundary of isolation or special operating iimitations. Protective tags shall not be applied, altered, or removed except under applicable established procedures by authorized employees. Every person working around equipment that is involved in a clearance will be responsible for recognizing the boundaries established by protective tags and the conditions imposed by the protective tags and must in no way violete the areas and conditions outiined. 13.5.5 Maintenance Instructions The plant maintenance program will be designed to safely and efficiently provide maintenance and repair to keep the plant in good operating condition. Maintenance work will be initiated through work requests and/or by the preven-tative maintenance program. Safe working conditions will be assured by the use of TVA's hold order, clearance, and special work permit instrucilens. Complex and critical maintenanco operations which require step-by-step performance will be detailed in written instructions. These instructions covering mechanical, electrical, and instrumentation maintenance will provide information to assure proper coordination of operating and maintenance employees as welI as step-by-step procedures to be folIowed by the craftsmen doing the work. Amend. 61 13.5-3 Sept. 1981 l l -
13.5.6 Surveillance Instructions Instructions will be prepared covering the conduct of all surveillance tests and inspections designated in the plant technical specifications. These instructions wil l specify prerequisites, precautions, refer ences, acceptance criteria, necessary step-by-step actions for conduct of the tests and return to normal, data sheets, and signatures of those conducting and reviewing the tests or inspections. De talled test schedules and records will be maintained to assure that all streelllance requirements are conducted in a timely manner and the results are properly documented. 13.5.7 Technical Instructicns instre::t ions covering routine technical operations will be prepared as required. Examples of these operations are chemical sampling ar:d analysis, chemistry control, and calibration of vital instrumentation. Fuel accountability instructions delineating the requirements, responsibiii-ties, and methods of nuclear material control from the time new fuel is received until it is shipped from the plant as spent fuel will be utilized. They will provide detailed steps for physical safeguards, inventory, accounting, and for preparing records and reports. 13.5.8 Section Instruction Letters Each section supervisor will, as the need arises, prepare numbered Instrucrion letters pertaining to administrative routinos, responsibilities, and methods to oe foilowed by members of his section. 13.5.9 Site Emeroency Plans These plans are discussed in Section 13.3. 13.5.10 Radiation Control Instructions Radiation control Instruci ans are written and made available to all plant personnel. These instructions include permissible personnel exposures consistent with 10CFR Part 20 and other requirements and guidelines to minimize radiation exposures. All plant personnel will be required to follow these procedures. O Amend. 61 13.5-4 Sept. 1981
O O O i 3 , - _ _ _ _ i d FIRE BRIGADE HANDBOOK
, [ HEALTH mYSICS MANUAL j ADMINISTRATIVE RELEASE MANUAL DIVISION PROCEDURES MANUAL i
0PERATIONAL QA MANUAL
- TVA HAZA RD CONTROL MANUAL j NUCLEAR MATERI ALS MANAGEMENT GUIDE l
RADIOLOGICAL EMERGENCY PLAN i l l * ! STANDARD PRACTICES 4 - i F Y' ) ui i GEtlERAL I OPERATING NORMAL INERGENCY HAlNTENANCE S ECT 10 N I INSTRUCTIONS OPERATING OPER ATI NG INSTRUCTIONS INSTRUCTION I INSTRUCTIONS INSTRUCTIONS LETTERS i i ! SITE RADIOLOGICAL OP RA G SURVEILLANCE TECNNICAL } EMERGENCY CONTROL INSTRUCTIONS INSTRUCTIONS I NS TRUCTIONS pung INS TRUCTIONS l l' ? L' [ K Figure 13.5-1 Plant Procedures 4
13.6 PLANT RECORDS 13.6.1 Plant History The CRBRP record program, under WA's responsibility as the plant operator, will observe all acts of Congress, Executive orders, and regulations of Federal agencies having jurisolction in records administration and comply with 10CFR50, Appendix B, Section XVil. TVA complies with Federal Power Commission regulations concerning the preservation and disposal of records of public utilities and licensees, f asofar as these regulations apply to TVA records relating to the generation, transmission, and sale of electric energy. The P1 ant Administrative Officer has responsibiiity for the general supervision and coordination of the plant master file. 13.6.2 coeratina Reccrds This will be included in the FSAR. 13.6.3 Event Records This will be included in the FSAR. O O Amend. 61 13.6-1 Sept. 1981 l
13.7 Radioloalcal Security N
~
The requirements of applicable provisions of 10 CFR 73.40, 73.55, 50.34 (c) and NRC Regulatory Guide 1.17 will be met for the CRBRP. This section discusses in general how the CRBRP will meet the requirements. The CRBRP Physical Security Plan, Safeguards Contingency Plan, and Security Personnel Training and Qualification Plan will be submitted as separate proprietary documents at the FSAR stage. Specific details will be provided by these 61 plans. 13.7.1 Oraanization and Personnel The Division of Property and Services and the Office of Power of the Tennessee Valley Authority shall share the security responsibilities for the CRBRP. The organization chart shown in Figure 13.7-1 delineates this responsibility which is explained in the following Section 13.7.1.1 and 13.7.1.2. 13.7.1.1 Division of Prooertv and Services The Division of Property and Services (P&SYS), with the assistance of other TVA organizations, develops guides and standards on property protection, reviews protection plans for compilance with these guides and standards, and advises in their development and application. P&SYS provides police and fire protection service on properties for which it is responsible and furnishes ti?ese services to other organizations in i accordance with their protection plans. The Public Safety Service (PSS) In the P&SVS PublIc Safety Service Branch furnishes this service. The Chief of the Public Safety Service Branch functions as overall TVA Emergency Coordinator in carrying out P&SYS's security responsibilities with other TVA organizations and providing liaison with feden e state, and local agencies on security and emergency preparedness matters. His organization provides supervision for the PSS. The supervisor of the Nuclear Operations Section, located in the Public Safety l Service Branch Office provides supervision of unit supervisor at each nuclear i 61 Plant. l The PublIc Safety (PS) Security unit at thr CRBRP is under direct supervision 61 of the PublIc Safety Service Branch, but f unctions as an onsite armed security force for the Plant Manager according to existing plans and the Manager's requirements. The Public Safety Service Branch is responsible for recruiting, training, and assigning security force personnel to the CRBRP unit as required. O l Amend. 61 13.7-1 Sept. 1981 l - . - - .. ,, ... . . . - - . . - . . - - .. - .--_.-. - - - , . , , .
13.7.1.2 Office of_ Power The Office of Power is responsible for prctection of power properties. It develops ootalled plans and applies specific measures, with the advice of Fa$VS. The Chief, Management Services Staff ? tm 'ower Manager's Of fice, represents the Power Manager on all security m- e, The Power Security Section which reports to the Chief of Management de.,1ces Staff cordinates planning and administration of industrial security measures with all concerned. Each Power division is responsible for security and fire protection and prevention of Power facilitles under its control In accordance with general policies and general Instructions from the Power Manager's Office. The 61 D1 vision of NucIear Power is responsIbie for security of nuclear plants. At the CRBRP, the Plant Manager and in his absence the Senior Plant Supervisor on duty is responsible for security of the plant in accordance with general poiicles, piens, and Instructions received through aministrative channels. 13.7.1.3 Emolovee Selection As discussed in Section 13.1, TVA will operate the plant, and accordingly, wI! provide alI ograting and security personnel who wIII be regular TVA employees. TVA appoints, promotes, transfers, and retains employees on the basis of merit and ef ficiency, as prescribed in the TVA Act and in accordance with other applicable Federal laws and regulations, it is the policy of TVA to promote
- present employees, whenever possible, who have demonstrated competence, l reliabilty, and stability to vacant positions in preference to hiring persons from outside the organization. This is often accomplished by upgrading employees through Internal training programs.
- Specific Instructions pertaining to personnel matters are contained in Section l Ill of the TVA Administrative Release Manual. These instructions are observed by all plant supervisors, especially as they apply to appointment, transfer, promotion, and retention of ernployees.
l Selection fo- a position is supportable by records of education, training, and experience, and by records of judgments which have been made regarding work l performance, ability, and condition of health. l l In selecting for placement or retention ir positions, covered by agreements negotiated between TVA and the employee organizations, the provisions of such agreements are observed. O Amend. 61 13.7-2 Sept. 1981
! i l
e Because of TVA's conformance to the Veteran's Preference Act, when employing i outside candidates for vacant positions, a large number of persons beginning employment have successfully completed tours of duty with the military forces of the USA. The availability for review of the military record of these candidates provides good control in the selection of. high-quality candidates. l i Each new annual TVA employee is given a physical examination and a netional agency check, and written inquiries are routinely made to references such as former employers, schools, and police. Before any employee is allowed , unescorted access to a nuclear plant protected area, there must be satisf actory results from his security check and emotional stability check. l PS of ficer selection procedures include a preemployment interview by the PSS area chief and one or more PSS unit supervisors in addition to the steps previously mentioned. Upon acceptance, the candidate's first six months of employment are probationary. Appointment as a PS officer Is dependent upon satisf actory service during this period and satisf actcry completion of
; training and qualification as provided for in the CRBRP Training and- ,
Qual:fication Plan to be submittd with the CRBRP Physical Security Plan. 13.7.1.4 Fmnlovee Evaluation Because of the general policy of promoting present emioyees rather than appointing candidates from outside TVA, most employees at the CRBRP will be known from their previous employment record with TVA. Although TVA employees are not given routine psychiatric examinations, they shall be given when an , employee's on-the-job performance Indicates that this is desirable. Observation of employee service is made as a regu!ar part of day-to-day > continuous supervisory function. When performing this function, supervisors ! shall be alert for any unusual behavioral patterns such as may result from
- mental stress, alcohol, or other drug abuse.
4 in addition to this kind of review, the performance of employees in management ' and salary policy positions are reviewed formally and the results reported in order (1) to f urtner aid in maintaining a high level of employee peformance and the maximum utilization of employee abilities; (2) to provide recorded evidence of employee performance for use in making judgments concerning transfc , demotion, promotion, and terminations; (3) to assure that anployees
- are adequately and systematically informed of the ef fectiveness of.their
! service; and (4) to f urther f acilitate the maintenance of a high standard of supervision in T/A. All employees' services are reviewed formally at the time [ of status changes and at such other times as may be required to achieve the l above purposes. A service review shall precede each recommendation for operator licensing or renewal of an operator license. s O Amend. 61 ! 13.7-3 Sept. 1981 i
13.7.1.5 industrial Security Trainina All employees shall receive training in security procedures with emphasis on being alert to the presence of unauthorized persons and evidence of forced entry. This training shall normally be conducted by a member of the Plant Security Force under the direct!on of the Plant Manager. 13.7.2 Plant Design The physical plant design has been developed so as to accommodate the necessary security provisions. TVA, along with the CRBRP Project Office and iis architect-engineer, Burns and Roe, will provide a continuing review of the plant design, as well as the detailed security provisions. Burns and Roe, as the architect-engineer for the Project, has been delegated the responsibility for detailing the security provi:, ions. The design criteria used at the CRBRP will assure that the physical security facilities and the plant layout are developed so as to thwart any attempted sabotage. The physical security design will: (1) Control entry to the plant site and portions of the plant; (2) Deter or discourago penetration by unauthorized persons; (3) Detect such penatrations in the event they occur; and (4) Apprehend in a timely manner unauthorized persons or authorized persons acting In a manner constituting a threat of sabotage. In the design and operation of the plant, care is taken to minimize the potential for Industrial sabotage by the use of access control measures to prevent unauthorized persons from entering the protected area. Should such persons succeed in entering the proteciad area, special access control measures will prevent them from entering vital equipment areas and the Special 61l Nuclear Material (SNM) material access area. 13.7.2.1 Design Features The design features and other physical security measures that will protect against or limit the effects of possible sabotage efforts include:
- a. A security barrier with dual inirusion detection system around the peri neter of the plant, with gates that are kept closed and locked except during tienes of authorized use.
O Amend. 61 13.7-4 Sept. 1981
- b. Employee and visitor parking located outside the security barrier.
- c. An Isolation zone extending from inside the security barrier to outside the barrier in which all activities will be controlled. This zone shall be vold of obtrus!ve structures and plant growth. In addition, a cleared zone will be maintained outside the isolation zone to f acil Itate observation of persons' approaching the isolation zone.
- d. A perimeter patrol road extending completely around the plant inside the security barrier.
61l e. Outdoor closed circuit television (CCTV) systems to permit observation . of the plant perimeter, Isolation zone, cleared zone, protected area, and approach roads.
- f. An outdoor iIghting system to provide 11lumination to the protected area and isolation zone at a level compatible for both visual and CCTV observation.
- g. A minimum number of exterior plant doors leading to vital areas, all of which shall be hardened against penetration and kepi locked or otherwise secured when not in use.
- h. A cardkey electronic access control system to control personnel access to vital areas in confcrmance with each employee's level of authoriza-tion.
I. An alarm system to indicate status of hatches, omergency exits and seldom used equipment or personnel access doors providing access to vital areas not cardreader equipped. J. An access control bullding (gatehouse) to contrci personnel access to the protected area and contain.ng,eaulpment to search personnel for 61 weapons and explosives. l k. A correunication system which will allow continuous communications ( between PS o.5ficers and the central alarm station. Also, redundant I cominunications links will be maintained beteen the plant and the local law enforcement agency. O !U ! Amend. 61 l 13.7-5 Sept. 1981
- 1. An electric power system to provide emergency power to the security and lighting loads during periods of " blackout" or loss of normal power.
- m. A force of trained, uniformed, and armed PS officers used on a three-shift basis to police the property, provide access control, respond to alarms, evaluate the situations, and neutralize the threats.
- n. Fire fighting and other emergency equipment located throughout the plant area to minimize the consequences of fires or explosions.
- o. Engineered safeguards and protective systems that are provided to minimize the consequences of fires or explosions or to minimize the ef fects of postulated major equipment f ailures, natural disasters, and operator errors which would also serve to minimize the ef fects of industrial sabotage.
13.7.2.2 Physical Arranaements The CRBRP site is in a remote location. It is uniIkely that major civil disorders would occur at or near the plant area. The plant is located on a peninsula formed by a meander of the Clinch River between river miles 14.5 and 18.6 near the center of a 1364-acre tract owned by and in the custody of the 61l United States Government (see Sectic 2.1). j 13.7.2.3 Owner-control led Area l Ultimately, a permanent access road to the plant will lead into the plant. O l During construction, a temporary construction road will lead into the I construction area. The perimeter of the reservation shall be marked prior to the completion of construction with signs to provide reasonable assurance that persons entering the area are aware they are on private property. Adequate roads shall be provided to patrol and control the entire reservation. Employee parking areas shali be located outside the security barrier so that only plant vehicles and trucks making deliveries will need to be admitted. A motor patrol of the reservation area shalI be made at least once each evening and night shift. While construction is in progress, the temporary construc-tion road will be the only route of access to the Projeci. A continuous access control guard post will be maintained on this road for the duration of 61l construction activities. Sectla 2.1 shows the reservation boundary of the owner-controfled area. O Amend. 61 13.7-6 Sept. 1981
13.7.2.4 Protected Area When all construction work is completed, there will be an 8-foot high perimeter security barrier enclosing the protected area. An isolation zone shalI be maintained both outside and inside the security barrier which meets or exceeds the NRC requirement as specified in 10CFR73.55 dated March 14, i 1980. A perimeter patrol road wilI be located Inside this barrier. A sectionalIzed intrusion detection system designed to be set f-checking and tamper-Indicating will be located along the barrier with sensors located on or between it and the patrol road. Closed-circuit television (CCTV) systems using lowilght level cameras including some with zoom lens and remote pan and tilt controls will be used to provide a means of promptly viewing the sector or general area involved. Proprietary Figures 13.7-2 through 13.7-5 Indicate 61 compl iance wIth ANSI N18.17-1973, Section '5-3. 13.7.2.5 Vital Eau toment and Vital Area @ - All vital equipment and material access areas shall be located within a vital 4 area or building which, !n turn, shall be located within the protected area. Doors and gates to vital areas and to other selected sensitive areas shall be kept closed and locked at all times when the areas are not occupied. 61l Proprietary Figures 13.7-6 through 13.7-11 Indicate compliance with ANSI N18.17-1973, Section 3.4, and other applicable guides and regulations. The material access area is located within the Reactor Service Building. No 6 11 activities other than those which require access to SNM or equipment employed in the process, use, or storage of SNM will be conducted in the material access area. J l O As construction nears completion and the equipment made operational, the doors and gates to vital areas shall be identified by signs which state that entry through them shall be with the permission of the shift engineer on a need basis. Upon completion of construction, these doors and gates plus others, including some exterior doors and the Power Storeroom shall be controlled by a cardkey access control system. The cardkey system shall be self-checking and tamper-Indicating and an emergency power source provided. The regular power supply and emergency supply will be supervised and the operation of each cardkey controlled door - tested no less frequently than once each seven days. All issues of cardkeys will be authorized by the Plant Manager according to individual needs of employees requiring access to areas controlled by the cardkey system. Each card in the system will be programmed into the computer 61 Individually and can be programmed out at any time if lost or cancelled. The cards will be issued and returned daily to insure that iMy do no leave the site. Also, since the card will be required in the performance of the employee's duties, this will serve as a continuing availability check of issued cards. l O Amend. 61 13.7-7 Sept. 1981 l
13.7.2.6 Alarm StatISD All security alanns will annunciato in a continuously manned central alarm station, located within the Plant Service Building, and in a secondary alarm
** "I " " '" * " ' "' "9' 61 Each sector of the outdoor Intrusion detection system and the operation of each cardkey controlled door will be tested no less frequently than once each seven days. Onsite and offsite communication facilities and the CCTV system will be tested at the 1,eginning of each PSS work shift.
Both alarw stations shall be considered vital areas. All intrusion alanns, 61 emergency exit alarms, and other alarms will be required, when purchased, to meet the level of performance and reliability specified by Interim Fed 6ral Specifications W-A-004508, GAS-FSS, dated February 6, 197.5 13.7.2.7 Security Barrier The security barrier shall consist of an 8-feet high No. 9 gauge chain link fence (7-feet fabric and 3 strands of barbed wire on angle brackets). Other fencing used for security and located within the protected area may be only 7 61 feet high without barbed wire. The alignment of the security barrier has a minimum nucber of angles and curves to facilitate effective observation and maximum length sectors of the intrusion detection system. 13.7.3 Security Plan The Plant Physical Security Plan shall describe security measures used to minimize the potential for radiological sabotage including access control, surveillance of vital equipment, and plans for responding to security threats in more detail than covered in the following paragraphs. 13.7.3.1 Access Control The CRBRP shall have a perimeter security barrier that sncloses all vital areas. The plant shall have two portals for normal access (1) a personnel portal and (2) a nearby vehicle gate. General public visitors shall not be permitted inside the security barrier. Employees and special visitor's parking arees shall be located outside the security barrier. Vehicle access shali be Iimited to those required for delivery of material, operations, maintenance, and security of the plant. Persons, packages, and vehicles shall be subject to search upon entering, leaving, and while within the plant area. There shall be a minimum number of outside accesses to the nuclear Island buildings. All of them shall have penetration resistant doors with frames, hinges, and locks or security devices designed to prevent forced entry and shall be alarmed or cardkey controlled. These O hend. 61 13.7-8 Sept. 1981
i doors shall be kept locked or secured when not * .ae. Also, a number of interior doors shall be cardkey controlled to . vent unauthorized accets to certain more Important areas. All persons authorized to enter the prote ,e area unescorted shall have had a , satisfactory security check and emotione stability check and shall have completed as a minimum a brief plant Indoctrination and radiation protection course which describes plant organization and layout, controlled zones, radia-tion and contamination hazards, exposure limits and controls, elementary 1 health physics, and pertinent sections of the site emergency plan. i Even those persons who are authorized unescorted access shall have their l movement limited by physical barriers, such as locked doors, to prevent them
- from entering areas containing vital equipment or areas of high radiation 1
levels. Only those who need access to these areas shall be provided means of entering. t When special visitors and other persons who have not completed this training enter the protected area, they shall be escorted by an employees trained in radiation protection and plant emergency procedures. The escort shall bs i responsible for the safety and-action of the people in his charge until he checks them out of the access portal. 13.7.3.2 Control of Personngt bv Catacarles Employees and visitors authorized unescorted access to the plant protected . areas by the Plant Manager will be issued a photo.-type identification (ID) badge with tamper-resistant features. These persons will be identified, Issued a radiation detection badge and dosimeters, and then be admitted to the
- protected area by the on-duty PS officer. Other persons not issued ID badges who require escorts may be identified by personal recggnization, WA identification card, or other available Identification media. They will be issued a white numbered visitor's badge to be worn on their outer garment while within the plant protected area.
l The only unescorted construction workers who may be inside the confines of the operating unit are those selected to perform maintenance or other work before final acceptance of equipment. Upon being granted unescorted access by the Plant Manager, these persons will be issued a photo ID badge and given a brief security Indoctrination course covering the evacuation procedure and relevant sections of the site emergency plan. Contractor personnel, manufacturers' representatives, and other special visitors who require unescorted access to the plant shall be logged in and i badged by the PS officer on duty. The officer shall then call the appropriate plant supervisor and arrange for an escort. 13.7.3.3 Access Control DurInc rearnegjag Upon nearing of an emergency, the FS officer on duty at the access portal , shall lock all doors to ensure controlled entry and exit. Special visitors who are onsite shalI be escorted to the access portal. ! Amend. 61 13.7-9 Sept. 1981 J
,e- 6 -y e e ,r*w ,a -,w w .we wrygm,r-,,w-w.v-+.,,,,-w- . r.v., -%-,-w-,up ry,ww w .ey, s w e s w s w v a,.ww.--, +- ce= w e w -= w -e = w =-ree-w e m
Plant employees shall report the predesignated stations from which they will be dispatched as needed to combat the oaergency. All access control procedures will be compatible with the CRBRP Radiological Emergency and Contingency Plans. 13.7.3.4 Survolliance of Vital Eautoment and Material Access Areas Unit operators shall continuously monitor the status of plant systems and equipment by means of annuaciators, Indicating iIghts, Indicators, and recorders. New equipment or material shall be inspected on delivery. Operating logs and computer printout data shall be periodically examined for changes in equipment performance. Most equipment will be in continuous operation and any change will Immediately be detected by the operator. Standby and emergency equipment shall be periodically tested on a routine basis as required by the technical specifications. Assistant unit operators shall Inspect equipment and spaces at least once each shift. In addition, the assistant shift engineers and other supervisory personnel knowledgeable in plant conditions shall make frequent unscheduled inspection tours through the plant. Procedures shall be employed to control access to the ylcinity of the j material access area. In addition, activities in the vicinity of the materisi
- access area will be monitored. The canbination of these ef forts should i
provide reasoneble assurance that unauthorized physical changes in the status of components of equipment will not be undetected for long periods. Key operating log sheets and selected recorder tracings sisall be reviewed daily except for weekends by the Plant Engineering Section. Abnormal changes observed shall be called to the attention cf the Plant Manager and the appropriate supervisors for investigation and corrective action, !f required. This operational audit shali serve to assure early detection of physical changes which would have a significaat bearing on plant performance. 13.7.3.5 Potential Security Threats The security system is designed to deter unauthorized persons from entering the protected and vital areas and to detect such attempts. Operating personnel are trained to be alert for unauthorized persons and to l appropriately notify the Pubile Safe +y Service. Detailed descriptions of decisions / actions regarding potential security threats shall be included in the CRBRP Contingency Plan. Planned actions in the event of civil disturbances, bomb threats, and other emergencies, will be included. Detailed procedures shall be provided plant employees so that they 61 may cope with these and other events in the optimum manner possible. 13.7.3.6 Administrative Procedures in the event of an incident of suspected sabotage or condition which threatens i the security of the plant, the Public Safety Service shall immediately notify the Plant Manager and initiate a thorough investigation. A report shall be prepared which includes as a minimum the cause of the O Amend. 61 13.7-10 Scpt. 1981
ever.t, extent of damage, if any, and action taken to prevent recurrer.cc of similar evont. Copies of the report shall be sent to the Plant Manager; O Assistant Director of Nuclear Power (Operations); Power Security Section; Chief, Public Safety Service Branch, P&SYS; and the Office of the General Counsel, When appropriate, the Plant Manager shall also report the situation to NRO. Representatives of the Power Security Section and Public Sefety Service Branch, P&SYS, shalI make an annual audit of the CRBRP Physical Security Plan for adequacy of content and performance. Based on their audit, they will make recommendations for revising and updating the plan and related plant pro-cedures. 13.7.3.7 Test and Ins 7settons This information will be supplied in the CRBRP Physical Security Plan. 4 O l i O Amend. 61 13.7-11 Sept. 1981
n . n 7- .. (u,( Board of Directors l l . Office of General Manager I 4 { Office of Power Office of Management 4 Operations Adm. Services i I Other Divisions I l Division of & Functions Management Other Division of Nuclear Power Services Staff Divisions Property &
; Services M&ES Operations Power Security Section i
Other Industrial Nuclear l I l ~ Staffs .. ' Other 4 F j ll l Branches u ,, I I j q Nuclear Power _____,-_____________j.l l Security Staff ' l '_ l- - - _ - - -Public l Safety Serv. l , l . j l l
!. l i CRBRP & i i Other l l Othler i Sections Nuc. P1ts ---------------l-----------------------------l--------------- Nuclear Operations
, .l -l
- l l Section v, > -----_------_--_i ; '
+
e2 P1 ant PSS -----------------------------------------------e
.c R Units +++++++++++++++++++++++++++++++++++
8$ Coordination Figure.13.7-1 CRBRP Security Organization
+++++ Administrative. Responsibility
***** . Functional Responsibility Overall Responsibility -
O l l l l Figures 13.7-2 through 13.7-11 O ! are Proprietary and .1ot 61 included in the PSAR. l l O Amend. 61 13.7-13 Sept. 1981 _ _ _ _ _ _ _ _ _ *T*"*Nrrey- ,_f ,
CHAPTER 14 INITIAL TESTS AND OPERATION TABLE OF CONTENTS fAER 14.0 INITI AL TESTS AM) OPERATION 14.1-1
14.1 DESCRIPTION
OF TEST PROGRAMS 14.1-1 14.1.1 Preoperational Test Program 14.1-2 14.1.2 Startup Test Program 14.1-2 14.1.3 Administration of Test Program 14.1-3 14.1.3.1 ResponsibliIties 14.1-3 14.1.3.2 Procedure Preparation 14.1-4 61 14.1.3.3 Conduct of Tests 14.1-5 14.1.3.4 Test Program Schedule 14.1-6 O 14.1.4 Test Objectives of First-of-a-Kind Principal Design Features 14.1-6 14.1.4.1 in-vessei Transfer Machine 14.1-6 l 14.1.4.2 Primary / Secondary Sodium Pump 14.1-9 14.1.4.3 Secondary Control Rod Drive 14.1-17 14.1.4.4 Upper Internals Structure and Upper internals 14.1-18 Structure Jacking Mechanism 14.1.4.5 CRBRP Intermediate Heat Exchange 14.1-19 61 14.1.4.6 Steam Generator Module 14.1-21 14.2 AUGNENTATION OF OPERATOR'S STAFF FOR INITIAL 14.2-1 TESTS Af0 OPERATION l O l 14-1 Amend. 61 Sept. 1981
LIST OF FIGURES Figure No. g O 61l 14.1-1 Schedule for Initial Test and Operation 14.1-23 1 O i l l l O l 14_gl Amend. 61 Sept. 1981
i i Q BPTER 14.0 INITIAL TESTS AND OPERATION O This chapter provides Information relating to the initial plant startup and operation program to show that the licensee plans to develop and conduct a comprehensive test program on this firsi-of-a-kind plant, and that necessary early planning has been done for successful achievement of these goals. The need is recognized for development of a comprehensive preoperational and initial startup test program for the CRBRP plant, the preparation of adequate test instructions for carrying out the programs, the proper conduct of the test programs, and assuring the validity of the test results. The test programs will provide additional assurance that the plant has been properly designed and constructed and is ready to operate lie a manner that will not endanger the health and safety of the public; that the operating instructions 61 for operating the plant safely have been evaluated and demonstrated; and that _l the plant operations personnel are knowledgeable about the plant procedures 6 11 and operating instructions and fully prepared to operate the plant in a safe manner. The test programs will al,o include testing for interactions such as the performance of interlot circuits in the reactor protection systems. It wilI be determined that proper permissive and prohibit functions are performed and that circuits normally active and supposedly unaf fected by the position of the mode switch perform their function in each mode. Care will be taken to ensure that redundant channels of equipment are tested independently.
/
14.1 DESCRIPTION
OF TEST PROGRAMS The initial test program for the plant is divided into two parts; preoperational testing, and the initial startup testing. 6d Phases 1 and 2 preoperational tests, are those conducted prior to fuel loading to demonstrate the capability of structures, systems, and components to meet performance requirements, including safety-related requirements. These tests are used to demonstrate that overall plant performance is acceptable and that the CRBRP is ready for initial Installation of fuel. For scheduling purposes, ! 61l the preoperational tests are divided into two phases. Phase 1 is defined as l testing following plant turnover from the constructor to initial introduction i of sodium into the Heat Transport System (HTS). Phase 2 testing is defined as the plant testing which requires sodium in tb HTS prior to initial core load. 61l Startup testing consists of such activities as fuel loading, precritical l tests, low power tests (including critical tests), and power ascension tests performed after fuel loading to completion of acceptance testing that confirm the design bases and demonstrate, where practical, 14.1-1 Amend. 61 Sept. 1981
-,-- ,-. -m . -.,,~,- - .,, -- , . . , - , - - . -----c- -.m.m, ,,-,,,~,.,,.----,,.e , ,--,,---,,,---.,---,--,------,,e.---.--.-e,,.-y -
, , - + - , , - - - , ,
that the plant is capable of withstanding the anticipated transients and postulated accidents. Startup testing is also divided into two phases for scheduling purposes. Phase 3 is defined as the testing period beginning with initial core load and extending to 5% powar. Phase 4 Is defined as the power ascension test porlod 61 and covers the power testing from 5% to 100% power. 14.1.1 Preooerational Test Proaram The objectives of the preoperational test program are to demonstrate the capabil ity of structures, systems, and components to meet performance requirements; to assure that, to the extent possible, procedures for operating the plant have been used and evaluated; and that the operating organization acquires suf ficient knowledge abaut the plant features and procedures to operate the plant in a safe manner. The preoperational test program will demonstrate not only that the design of systems, structures and components meet the objectives, but that construction of the plant has been done in a manner that assures that the plant can be operated safely. The preoporational teit program will begin only after a very significant portion of the plant construction is complete. Befor e a structure, system, or component is preoperationally tested, activities on it must be essentially compiete, wIth those Ineompiete portions clear 1y docunented In the test report. Before preoperational tests are started on a system, structure, or component, construction tests such as system flushing and cleaning, wiring checks, and leak tightness tests must be completed to the extent that meaningful test 61 results are obtained. Each system, subsystem, or component will have successfully passed the construction test and gone through a turnover l procedure pelor to commencement of preoperational testing. In addition, ! Initial calibration of instrumentation, and subsystem component functional 61 tests must be completed prior to subsystem preoperational testing. 14.1.2 Startuo Test Procram l The objectives of the startup test program are to assure orderly, safe fuel loading, low power testing, and approach to full power testing; and to confirm 61l the design bases and demonstrate, where practical, that the plant is capable of withstanding the anticipated transients and postulated accidents. The startup test program begins with fuel loading and ends with the Project Office accepting the CRBRP from the contractors based on satisfactory evaluation of test results and correction or acceptable provision for 61 corrections of all Identifled deficiencies and Incompiete items. Ttis program is started only after conclusion of all preoperational tests inat ;an be performed with no fuel in the reactor. l 14.1-2 Amend. 61 Sept. 1981
4 i ! l
. 61 l The startup test program is composed of the following activities:
31l Fuel Loading Tests (Scheduled in Phase 3) - provide assurance of a safe, orderly loading of the core, taking into consideration the first-of-a-kind core, the available nuclear instrumentation, and reactor control. 61] Precritical Testing (Scheduled in Phase 3) - includes thuss tests from initiation of core loading to criticality. These tests will assure that the startup proceeds in a slow and orderly manner, that charges in reactivity wilI be continuousIy monitored, that operations personneI are aware of core reactivity state, and all systems are aligned and in proper operation. 61l Low Power Testing (Scheduled in Phase 3) - includes those tests between i criticality and a 5 percent power level. These tests will confirm nuclear design parameters before nuclear heating and give confidence that the reactor power can be increased. Nuclear instrumentation will be confirmed at increased power levels. Power Ascensf or. Testing (Scheduled in Phase 4) - includes those tests at 6] various power levels between 5 and 100 percent power. Confirmation of reactor and plant design parameters are obtained at progressively higher power levels, each step giving confidence that the next higher power level can be safely accommodated. 14.1.3 Administration of Test Proaram This description of the administration of the test program applies to both the preoperational and startup phases. In planning and carrying out this program, j-j the guidelines of Regulatory Guide 1.68 will be used insof ar as they apply to i the LMFBR. This includes all of the Guide except those portions of Appendices I A&C that are unique to light water reactors. Other regulatory guides will be i reviewed at the time detailed test instructions are being develcped to i establish which guides have applicability to the program. t 14.1.3.1 RESPONSIBILITIES l 61l The Project Office (DOE) has overall responsibility for the plant initial test i program. Portions of the program have been assigned to others as follows.
' TVA is assigned responsibility for conduct of ihe Preoperational and Startup i
Testing along with its responsibility as plani operator. It also hc. responsibility for review and approval of tes- specifications for preparation of test Instructions from the test specifica' ions, for evaluation of adequacy l perating and emergency instructions during the test program, for on-site 61 approval of test results, and for recommending plant modifications as a result ! of deficiencies discovered during testing. As described in Section 14.2, the TVA normal oper ating organization will be. augmented during this test period. The responsibility ;Ur prformance of preoperational tests will be assigned by 14.1-3 I Amend. 61 Sept. 1981 I 4
,,,,,www,,,,m , won,w,em.,,n~,wwn-~m. . e,,w r y w w,- a , ,w -ee ev~,e e-~&n,--,-v w n , -s e-, ,,,-w--nw-aw,
the plant manager to a TVA Preoperational Test Section. The responsibility for performance of startup tests will be assigned to the plant technical -taff augmented by technical spectalists f rom TVA's central of fice. The Project Office has responsibility for review and approval of all aspects of the test program including scope, content, schedule, test specifications, test instructions, test results, and any plant modifications required as a 61 result of the test program. Westinchouse ARD as Lead Reactor Manuf acturer (W-LRM) is assigned responsibility for preparation of technical aspects of the initial plant test program. In carrying out this assignment they will utilize the services of the Reactor Manuf acturers, W-ARD as Reactor Manuf acturer (W-LRM), General Electric (GE-RM), and Atomics International (Al-RM) and coordinate with and uttiIze the input of Burns and Roe as Architect Engineer (A-E). This includes early planning of scope, schedule, and sequencing of the testing interfacing the construction schedule. W-LRM is responsible for preparation of test specifications, and for reviewing the test Instructions and test results for NSS systems under their cognizance. W-LRM is responsible or design of plant modifications required to their systems as a result of deficiencies discovered g during the test program. W-LRM will assign on-site personnel for technical direction during the test program. Burns and Roe as Ar chitect Engineer ( A-E) is assigned responsibility for preparing test spedfications, and for reviewing for technical adequacy the test instructions and test results of those BOP and NSS systems under their 61 congnizance. The A-E '= also responsible for design of any plant modifications required To their systems as a result of deficiencies discovered during the test proc, ram. The A-E will assign on-site personnel for technical direction of tests to their systems. 61 l W-LRM and the A-E will establish on-site staf f during the test and startup period for technical direction of the initial test and startup program. This technical direction will include supplying technical advice and Information to operations personnel to assist them in making decisions. This staff will have the capability to support the TVA plant staf f in both operations and testing. Technical direction does not include supervision of operations personnel. The responsibility for safe operation of the plant rests with TVA as the plant operator, t 61lStneandWebsterEngineeringCorporationastheConstructorisassignet responsibility to assist TVA as plant operator in assuring that all prerequisites are met before tests are started, for Insertion of detailed scheduling of the test program into the plant construction schedule, and for assisting as required in repair or modification of the plant as a result of deficiencies found in the test program. 61 l 14.1.3.2 Procedure Outifne Preoaration The Project Office has assigned the responsibility for preparation of test specifications, operating, maintenance, and surveillance procedure outlines to Westinghouse as Lead Reactor Manufacturer and Burns and Roe as Architect-Engineer. Using information 14.1-4 Amend. 61 Sept. 1981
from the A-E, GE, and AI, W-LRM will provide a current Test Sequence document s that describes the initial test program content and schedule. This document will be reviewed,and approved by The Project Office as a basis for further detailed planning of the initial test program. A Test Abstract document may be prep.3 red that provides a short description of each test and gives the objectives for the test. A Test Network will be developed that shows the sc:edule for each test. These documents will be reviewed and approved by the Project Office. Using this planning information as a basis, W-LRM and the A~E will prepare 61 test specifications that describe each test requirement in considerable detail. These specifications will describe prerequisites, test objectives, general test methods, and acceptance criteria. Test specifications will be 61l reviewed and approved by the Project Office. t Using test spec'.fications as a basis, TVA will prepare detailed test Instructions. Format and content of these test anstructions will conform to the guidance given in Appendix C of Regulatory Guide 1.68. Operating instructions will be ava!Iable at the time these test instructions are prepared, so that the cperating Instructions may be referenced in aligning systems for testing. These detailed test instructions will be reviewed and approved by the Project Office and TVA. Final review and approva; for use will be performed by TVA as the plant 67 operator. 61l Changes to test instructions that modify the objectives, intent, or significantly change the method of test performance will receive the same review as the initial Instructions. Minor changes that do not modify the
\ objectives, intent or significantly change the method of test performance may be made by the testing organization and will be documented and subsequently reviewed as previously described.
61l Each test instruction will contain a prerequisites section that will describe in detali ali prerequisites, including constrLction related, that must be satisfied before a test is performed. Sign-off sheets will be provided with each I.istruction to record verification that prerequis!tes are satisfied. Completion r,f this sign-off sheet will be mandatory before starting the test. 14.1.3.3 Conduct of Tests TVA as plant ope. ator has been assigned responsibility for conduct of Preoperational and Startup Tests. .This includes review and approval for use of all test specifications, preparation of test instructions from the test specifications, performing all operator manipulations required under the plant operating license, assuring that all test prerequisites are satisfied, performance of details of the instructions, and collection of data for approval of test results. TVA will be assisted in the performance of this 61 responsiollIty by W-LRM and the A-E and ,any necessary subcontractors acting as technical directors for tests under their cognizance. During the conduct of the tests, plant operating and emergency Instructions will be tes ted wherever possibic. This will help assure that 14.1-5 Amend. 61 Sept. 1981 I
the ins + ructions can be used to safely operate the plant, and provides further assuranc.a that the operator is f amillar with the Instructions and thoroughly trained to operate the plant. Test results will be compared to acceptance criteria by the on-site test group. DefIclencies wIlI be immediately reported to the Project Office, 61 W-LRM, and/or the A-E. Each deficiency will be evaluated by these participants, and the appropriate corrective action specified, such as retesting or instruction change. These corrective actions will be reviewed 61l and apprcved by the Project Of fice and TVA as plant operator. 61l A detailed review of test results will be made by W-LRM or the A-E for tests within their scope of responsibility. This detailed review will confirm the technical adequacy of the system, component, or structure to operate in accordance with design specifications. 14.1.3.4 Test Procram Schedule Figure 14.1-1 shows the schedule for each major phase of the initial test progrcrn. It also shows the schedule for preparation of plant instructions, key milestonas in staf fing for operation, and augmentation of ti.e plant staf f fcr Initial startup test assistance. As shown on this schedule, all plant instructions will be prepared befwe f uel loading. Operating, maintenance, and survelliance test instructions will be started at about the same time as initial test instruction preparation, and is 61 scheduled for completion before fuel loading. A smalI group of key TVA operations personnel wilI be on-site about three O years before start of the preoperational test program. From this nucleus the on-site operations staf f will increase in size at a sufficient rate to provide adequate support for the preoperational test program. This schedule allows suf ficient time for plant f amiliarization, and procedure review before testing starts. As described in Section 13.2 and shown in Figure 13.2-1, basic nuclear cocrses for operators and specialists training for technical personnel as well as assignment to a sodium cooled f ast reactor will have preceeded this period 14.1.4 TEST OBJECTIVES OF FIRST-OF-A-KIND PRINCIPAL DESIGN FEATURES The following Test Abstracts are provided per US-NRC NUREG-75/087 - Section 14.1, Review Responsibilities item 2 for special, unique or First-of-a-Kind principal design features ..cluded in the CRBRP. 14.1.4.1 IN-VESSEL TRANSFER MACHINE The only equipment of the reactor refueling system, which is mnsidered first-of-a-kind and unique to the CRBRP, is the in-vessel transfer machine (IVTM). The IVTM is installed in the reactor head during reacter refueling and is discussed in detail in Section 9.1.4.4. 14.1-6 O Amend. 61 Sept. 1981
in order to minimize preoperational testing of reactor refueling system p) ( equipment at the CRBRP, the IVTM will be tested and checked out extensively at the off-site test facility (currently planned at ETEC). The off-site tests are scheduled early in the program to ensure correc+ive actions can be taken to qualify the IVTM for CRBRP service without jeopardizing the overall plant construction schedule should any IVTM deficiencies be uncovered. The IVTM prototype will be tested extensively to demonstrate that the IVTM meets its specification performance and design requirements. The complete and Integrated IVTM assembly wilI be tested, including the control console with the minicomputer. 1 Af ter the IVTM has been assembled at the test site, and the assembly has been ! c:ecked out, the IVTM will first be subjected to individual and Integrated checkout tests. Following this, the IVTM will be performance tested simulating core assembly transfers. The tests will be performed in special test facilities containing a cluster of at least seven simulated core assemblies. The cluster will be capable of relative vert
- cal and horizontal displacements and side loads.
A. IPOlVIDUAL CHECKOUT TESTS The purpose of the individual checkout tests is to verify that the following IVTM functions can be performed: 61 l 1) Grapple and release of a Core Special Assenbly (CSA). 61 l
'" Raise and lower a CSA to positions corresponding to those encountered in the reactor vessel.
61 l 3) Identify and orient a CSA. 61 l 4) Provide adjacent CSA holddown when removing a CSA from the CSA cluster.
- 5) Provide cover gas containment and seal leakage detection capabiiIty.
Specific tests will include the following:
- 1) Calibration and checkout of all IVTM Interlocks, load cells, and the entire load control system.
- 2) Verification of all functions of the core assembly identification system.
- 3) Checkout of the grapple and holddown sleeve drive systems including removal of an artificially jammed core special assembly, p 14.1-7 Amend. 61 Sept. 1981
- 4) Calibration and checkout of the grapple and holddown sleeve position Indication systems.
- 5) Verification of the seal leakage monitoring and the seal pressurization control systems.
B. INTEGRATED CHECK UT TESTS The purpose of these tests is to prove that the IVTM meets the following objectives:
- 1) The IVTM can perform the sequence of functions iIsted in Section A yk' are required to transfer a core assembly in accordance with given operating profiles when asing computer and manual controls.
- 2) Insertion and removal of core assembly into and f rom the core can be accomplished under maximum misalignment in combination with maximum core assembly push and pulI loads.
61l 3) Release of coro assembly into an incorrect core position is prevented.
- 4) Release of a core assembly Into a transfer position in the absence of a core component pot cannot be accompiished.
- 5) Premature release of a core assembly during operation over the core is prevented when the core assembly is at a vertical position higher than a small tolerance above the fully seated position.
C. PERFORMANCE TESTS These tests are designed to simulate reactor refueling opc ations equivalent to at least fIvo refueling periods. The tests will be performed with e cluster of seven core special assemblies. 61 The core special assembly cluster will be of fset in relation to the IVTM to simulate core assembly insertion and removal under misaligned conditions. Integrated operations of the IVTM, control console, and computer will perform simulated actual refueling operation:. The major test objective is to demonstrate that all IVTM components, especially dynamic seals, will perform for a minimum of one ref ueling cycle. The test goal for all mechanical components of ihe IVTM (excluding clastomeric seals) is to demonstrate operation without f ailure. Post-test inspection of the mechanical components will establish the acceptability of component wear. The folIowing results w11l be ob+alned f rom these tests:
- 1) Weer data of dynamic seals.
- 2) Wear data of mechanical components.
14.1-8 Amend. 61 Sept. 1981
- 3) EstabiIsh transfer cycle speeds for automatic and manual operation.
- 4) Wear data of core assembly identification pawl.
- 5) Any operational limitations. l
- 6) Any defIclencies in the opnrating components and/or in the design.
- 7) Verify the computer control of the fuel transfer cycle.
- 8) Verification of'the core assembly identification system with respect to wear data obtained in item 4 above.
- 9) Verl'Ication of checkout and operational procedures.
D. PREOPERATIONAL IVTM TESTS AT CRBRP Those IVTM operations which are not simulated in the special test f acilities will be performed af ter IVTM installation, adjustments, and checkout at the CRBRP reactor small rotating plug prior to fuel loading. These tests will include:
- 1) Insertion and removal of core special assembiles into and out of a 61 l Core Component Pot (CCP), and transfer of those assembiles between selected core addresses.
- 2) Integrated operational tests of the IVTM with the reactor rotating plugs (RRP).
- 3) Integrated tests of the IVTM to demonstrate design protection against off-normal operations to confirm accident analysis assumptions.
- 4) integrated operational tests of the IVTM and Interfacing reactor refueling system equipment to assure joint operability.
Before the IVTM is installed on the small rotating plug for the first time, and af ter that, each year before ref uel irg, al l IVTM functions required for transfer wilI be checked out in the dry IVTM maintenance and storage f acilIty. 61 l 14.1.4.2 PRIMARY / SECONDARY SODIUM PUW Prototype Pump Tests A. Prototvoe Pumn Water Tests at Suonller's Facility The objectives of water testing is to make final trim to the Impeller and verify that the hydraulic performance of the pump meets the specification requirements regarding head and flow relationship, Net Positive Suction Head requirements,
)
14 1-9 Amend. 61
Sept. 1981 1 4
, - , - ,, y- -----s -------y - ,- - . , - ,. ,r.--- -
,+,.y,--- -, .--~,-m - - . - , . - -
and to verify coast down head of flow versus time, bubbler performance (level control), capability to operate for a sustained endurance period, and operation at a loop impedance ccmparable to 2 loop plant operations (Runout to 41,000 gpm). Functions which will be tested are:
- 1. Heat versos flow for plant loop impedance with speed as the variable.
- 2. Head versus flow for constant speed with variable Impedance for several different speeds. Check for hydraulfc Instabilities as reflected in slope of H-Q curves.
- 3. Net Positive Suction requirements will be checked by reducing cover gas pressure "hile operating at rated spee/. and head and flow until degradation or performance or excessive vibration is detected.
- 4. Coast down head and flow versus time will be checked.
- 5. Levol control of pump Internal fluid will be checked by varying the cover gas supply at the pump and monitoring pump fluid level.
- 6. Pump Auxiliary performance (Shaf t Seal Lubrication) will be evaluated by measuring seal leak rates during the sustained endurance run.
- 7. Pump hydraulic performance (head-flow) will be monitored for stability, and vibration levels of the , sump will be monitored during the steady state endurance runs.
- 8. Pump vibration will be monitored during startup an. aast down tests.
In addition to the water tests of the prototype pump, a scale model pump is being tested. B. Prototvoe Pumo Sodium Tests at the Sodlum Pumo Test Facilltv The overalI objective of Prototype Pump Testing is to prove capability of the pump to deliver 10000F sodium at the head and flow conditions specified for the 1est, and to verify that fluid borne temperature transients up to the Iimit of the test facility do not cause malfunction (bearing seizure). 14.1-10 l Amend. 61 Sept. 1981 l l
Specific objectives are:
- 1. Demonstrate that th, pump is mechanically & hydraulically stable when operated through its full Casign speed and flow range and to verify hydrostatic bearing performance in the sodium environment.
- 2. Determine pump hydraulic characteristics (head-flow map and effIclency) In sodium.
- 3. Demonstrate that high-temperature, and the associated structural temperature gradients do not degrade mechanical operation or hydraulic performance.
- 4. Demonstrate that the pump and pump auxillaries are capable of sustained operailon while pumpirg liquid s' alum at vr-lable flows and speeds.
- 5. Demonstrate ~ pump pony motor operation; verify hydrostatic Learing performance in sodium at pony motor speed, demonstrate pony motor developed head at near shut-off, measure head-flow characteristics at different pony motor speeds and different hydraulic loop impedances.
- 6. Determine any deleterious structural distortion caused by convection in the gas spaces.
- 7. Demonstrate ability of the pump to withstand sodium fluid temperature transients which simulate predicted plant operating and upset transients.
- 8. Demonstrate capability of the standpipe-bubbler to maintain adequate sodium level in the pump during steady-state and operating (speed and flow) transients.
- 9. Verify the pump drive response characteristics with the pump operating in sodium wIth Ioop impedance simulating the plant.
i 10. Demonstrate flow coastdown characteristics (head, flow, speed) from maximum facility flow and from pony motor speed and correlate to similar measurements made in water tests. Determine pump and motor compliance with rotating kine 11c energy requirements per E-Specif Ication 22A3444, Yabl ; 3.3.1.
- 11. Measure compliance with Net Positive Suction Head (NPSH) requirements.
- 12. Verify Instrument, Operation, and Maintenance (10M) Manual procedures 61 f r checkout of assembly, operation, disassembly, maintenance, and Inspection of pump and auxillaries.
14.1-11 Amend. 61 O Sept. 1981
- 13. Demonstrate the CRBRP prototype flow controller operatoin with the Drive System and the CRBRP Pertanent Magnet (PM) flowmeter.
- 14. Verify that established rate of dry pump preheat is satisfactory (es Indicated by tank and internal temperature gradients).
- 15. Determine hydraulic lapedance of the pump to low magnitude forward flow of sodium throtgh the pump rotor.
- 16. Confirm performance of the Shaft Seal regarding leak rates.
- 17. Verify suitability of the pump for subsequent use ' sodium after Cm ponent Handling and Cleaning Facility (CHCF) clianing operations.
- 18. Evaluate whether sodium migration upward or oli migration downward is a concern with the purge of gas feed, labyrinth, and shaf t seal arrangement.
- 19. Determine whether gas injection at the Intermealate Heat Exchanger (IHX) return nozzle causes adverse effects on pump sodium level stability or if sic 9 pumping occurs at the bubbler; and to measure 61 sodium carryover frm the bubbler to the gas system.
C. Primarv/ Secondary Sodium Pumas Construction and Preoperational Tests (No Sodium)
- 1. Pump Cover (Canopy) Seal Leakage Tests The test will be accompilshed by injecting helium into the canopy volume and monitoring the inside of the pump for the evidence of leakage by means of a helium detector. The test will prove the adequacy of the seal between the Upper Inner Structure and the tank, i.e., no cover gas leakage into the facility.
- 2. Electric Pcwer Phasing Check This test is a check of power phasing wiring frm the MG Set to the drive motor. Since high voltages are involved, it will be accomplished with substitute reduced voltages, and commercial phase meters, and *,erified later by observation of direction of rotation during star'up tests.
- 3. Pump Motor Runout Measurements The intent of the test is to verify that the pump and motor rotating el m ents are aligned. The objective is to 14.1-12 Amend. 61 Sept. 1981
verify that Installed runout is within specifications and to obtain cold (no sodium) shaft torque measurements as a reference value. O Commercial runout measurement Instrumentation is used.
- 4. Verify Operation of the Shaft Seal Lube System The intent of this test is to check out the automatic activation of oil transfer pumps, verify pressurization of tanks, to verify lube lines are fliled and flow is achieved, and to verify the leak tightness of the lubrication system. It is accomplished with the instrument panel which is a part of the Seal Lube System along with some activation procedures and operator action in draining and f11IIng tanks.
- 5. Preheat Monitoring The objective of tnis activity is to verify that thv pump heat-up rate does not exceed the specified rate (to prevent applying thermally induced over-stressing) and that temperature gradients throughout the pump do not exceed manufacturers Iimits.
Verify shaf t free rotation before and af ter preheat.
- 6. Level Monitoring During Plant Sodium Fill The Intent of this task is to verify that internal fluid level sensors detect rising sodium and that the pump tank / Upper Inner Structure s maintains leak tightness, and that pump gas flow is maintained.
D. Plant Preoperational Tests (Coolant inplace)
- 1. Pony Motor Sodium Circulation Run The objectives of this test are:
- a. Verify that the shaft seal lubrication system functions properly la automatic actuation of oil transfer pumps, oil leak rate, seal heat exchanger operation, and seal instrumentation.
- b. Verify that sodium level variations are detected by the inductive level probes.
- c. Verify that the sodium level control (purge gas and standpipe bubblor) system is performing properly.
- d. Verify that pump diagnostic Instrumentation through the Reactor Heat Transport Instrumentation System readout equipment is performing properly.
14.1-13 Amend. 61 Sept. 1981
- e. Verify that the pump rotating assombly operates satisf actorily with respect to vibration at pony motor speed.
- f. Verify pony motor operation (vibration, temperature, etc.).
- g. Verify shaf t seat oil leakage rates are within specification.
- h. Verify pump head and flow performance at pony motor speed.
I. Check pump pony motor performance wben electrical suppiy goes to emergency. J. Provide sodium circulation for sodium purification or other plant checkout. These tests will be accomplished with several sources of Information from the shaft seal lubrication system instrument panel, from the Reactor Heat Transport instrumentation System for sodium level Instruments, shaft position Indicators, and vibration measurements, and from plant Ioop *nstrumentation for developed head and fIow.
- 2. Pump Speed Control Run The intent of this test is to verify pump operations with the mrin drive motor, to verify pony motor clutch engagement, and to op'eate the loop at higher than pony motor flows for other loop activ! ties.
The objective is to:
- a. Verify head /fIow as function of speed. Verify shaft seal and lubrication system performance (automatic lube transfer, lube leakage rate, heat excaanger performance, Instrumentation) at all main motor speeds.
- b. Verify sodium flow rate changes realized from loop flow command.
- c. Verify sodium level behavior with speed changes and pump trips.
- d. Verify that pump rotating assembly operates satisf actorIIy regarding vibration.
- e. Verify coastdown (head, fIow, time) foflowing a trip.
- f. Verify proper engagement of the pony motor clutch.
14.1-14 l Amend. 61 Sept. 1981
4
- g. Verify motor and motor auxillaries are performing properly with respect to temperature and lining lubrication.
- h. Verify that shaft seal lubrication at 100% speed is satisfactory as Indicated by leak rate and automated tube oil transfer.
- 3. Multipump Tests (coolant Inplace, no fuel, but with a dummy core)
The Intent of these tests is to verify that multiple loop operation can be satisfactorIiy achleved by the controf system and pumps as i Indicated by response to speed command, flow command, and trip. 1 Verify that the individual loops are stable, (monitored flow corresponds to flow command) and that hunting does not exist. j Determine coastdown characteristics-(Head & Flow versus time) , following trip. Verify that pressure pulsations fran single and 2 combined pumps do not create any undesirable vibrations in the system. The objective is to:
- a. Test the pumps in all expected modes of plant operation.
- b. Verify loop flow stability (hunting).
- c. Verify pump coastdown (head, flow time) following a trip.
- 4. Seal Cartridge Replacement Validation This test consists of replacing a shaf t -aal . fran a shutriown pump which has been in the operating plant rodium loop. 'The Intent is to r verify procedures and equipment used fx seal replacement, and is a validation run where a plant operating environment will require-J personnel to adhere to the precautions necessary for eperating a hot (temperature) system and replacing a critical seal whose life is as 4 such as to require annual repiacement.
! The objective is to: Validate the procedures of replacing a shaft seal with loop sodium in place. Demonstrate:
- a. Purge.
. b. Sodium level control. l [
- c. Cartridge reptacement.
l i 14.1-15 Amend. 61 Cg Sept. 1981
- d. Post Installation checkout.
- e. Restoration of pump to service.
This procedure will be accomplished using special handling tools and by either &eining the sodium for the pump or by special loop procedures to prevent sodium level from reaching the vicinity of the shaft seal (pressurization, etc.).
- 5. Pony Motor Plant Operational Test (Fuel in Place)
Tha intent of this test is to verify that the pony motor operations yields the proper coolant flow for decay heat removal requirements. The objective is to:
- a. Verify that pony motor pumping flow rate is adequate for reactor decay heat removal.
- b. Verify leak tightness in pump pit as indicated by radiation sensors (primary only).
- c. Verify that temperature gradients in the tank structure resulting from pony motor ON or OFF (flow transients) or reactor temperature changes do not adversely of fect pump performance.
These tests will be accomplished using plant radlailon sensors, and the Reactor Heat Transport instrumentation system and the Plant Data Handling end Display System for diagnostic information, and Plant Control System for control.
- 6. Main Motor Plant Operaticaal Test (Single & Huiti Loop Operation)
The intent of this test is to varify that tre pump will perform satisfactorily for the plant over the full range of head-flow conditions. The objectives are to: l
- a. Verify pump performance at the several plant sodium flow / temperature, conditions.
- b. Verify leak tightness in pump pit as indicated by radiation sensors.
l c. Provide sodium flow in support of other systems. 1 l 14.1-16 Amend. 61 Sept. 1981
14.1.4.3 SECONDARY CONTROL ROD DRIVE The Secondary Control Rod satisfies the requirement to provide plant protection system shutdown capability which is both redundant and diverse from the Primary Control Rod System. Prior to installation of plant units there will be extensive tests conducted on components and prototypes. Tests to be conducted on components and prototypes are as follows. COMPONENT .IESI Daniper Water Test to 180 F Latch Sodium test to 1000 0F, to 1125 cycles on 1 unit Coll Cord Air Test 1000 cycles of 6 units Position Indication Accuracy tests plus long term stability Control Rod Flow Test Test in water to 180 F to determine flow spiits Latch Seal Test in water to 1800 F to determine pressure drops Nosepiece Flow Test in water to 180 F to determine pressure drops Prototype Tests Fuiiscaleprototgpetestsin 0 iIquid sodium 400 F to 1000 F to check scram time measurements Argon Control System Cycle over the range of temperature expected in plant REllABILITY TESTS l Latch Scram Accumulation of seram cycles In 10000F sodium Latch Real Time Accumulate long term holds (1 yr.) In 10000F sodium System Tests Test fulI scale prototypes in l lIquid sodium 4000F to 10000F accumulate scram to measure scram l times after short and long terms p holds , (/ 14.1-17 Amend. 61 Sept. 1981 l
RELIABILITY TESTS (CONT.) Pneumatic Valve /Cy1. Test the scram cyIInder under remperature and pressures expected in plant i Bellows Test the driveling bellows under sodlum vapor environments at l temperatures (4000F - 6000F) expected in plant service i Testing on the secondary control rod drive un;ts will be performed under plant start-up conditions to assure tnat the integrity of the plant units have not been viola ^ed during shipment, handling and Installation; to verify that proper Installation has been made; and that performance is not adversely af fecteo by fabrication tolerance build-up or by vessel expansion or' other thermal effects. t initially, the housing-to-vessel-nozzle seal will be checked for tightness and the position indicator system will be tested to assure proper functioning and to verify accuracy of rod position Indication, interface conditions will be measured to ensure that inputs to the secondary control rod drives are of a magnitude required to provide adequate secondary control rod response. Under plant operating conditions, a series of scram tests will be Initiated to verify scram time and repeatability, and successful withdrawal and latching functions. Failures, defined as deviation from specification performance values, or inability to perform the scram or latching functions on command, wilI require removal of the secondary control rod unit, disassembly, analysis of the fallere and defining of corrective action. The test series will be repeated using a modified secondary control rod plant unit. 14.1.4.4 DPPER INTERNALS STRUCTURE AND UPPER INTERNALS STRUCTURE JACKING MECHANISM Acceptance of the Upper Internals Structure (UlS) and Upper Internals Structure Jacking Mechanism (UISJM) as a first-of-a-kind design feature for CRBRP will be made on the basis of scale model testing, operational testing in the vendor's facility and verificntion of design characteristics daring the CRBRP Construction and Preoperat! anal and Startup Testing. Scale model water testing will be performed to confirm the thermal / hydraulic and vibration adequacy of the UlS and other outlet plenum structures. Flow distribution, pressure drop, and temperature distribution test data will be compared with thermal and hydraulic design criteria for the outlet plenum. Both steady state and transient tests will be performed. Typical outlet plenum structures will be dynam!cally tested to verify that there will not be adverse vibration during operation. Prior to operation during the CRBRP Preoperational and Startup Test Program, the UlSJM will be tested, as follows. Operation of the UlSJM 14.1-18 Amend. 61 Sept. 1981
control system will be verified independent of the Jacking mechcW sm. (' i Feedback signals which in normal operation come form strain gauge load sensors, position sensors and limit switches on Jacking me.chanisms will be electrically simulated into the control system. The Jacking mechanism vendor will verify satisf actory operation of the motor, gear and Jack for each Jacking mechanism independent of the UlSJM control system. Finally, testing will be peformed at the UlS vendor to verify that the upper internal structure in combination with the four Jacking mechanisms and the Jacking mechanism control system can operate in such a manner as to meet the overalI system functional requirements. During the installation at the Site, assembly will be checked and alignment verified as part of the CRBRP Construction Test Program. Overtravel limit switches will be set during i n. ital I ati on. The UlSJM sealing arrangement will be tested, as folic 4s. Prior to UlSJM fabrication, seat leak testing will be conducted on a prototypic seal which includes the buf fer 0-ring seals and piston rings. Fabrication seal tests will also be performed by the UlSJM fabricator. A final leak check of the seals will be performed as part of the CRBRP Construction Test Program af ter installation at the site. The operability of the four mechanical Jacking systems wilI be verifled in conjunction with the fuel handling operations which will be performed during the CRBRP Preoperation and Startup and Test Program. Test Phase 1 - Fuel Handling System Operation Check in Air Test Phase 2 - Fuel Handling System Operation Check in Sodium and installation of Two Special Core Asembiles Test Phase 3 - a. Removal of Special Core Assembiles for Inspection
- b. Initial Fuel Loading 61 The vibrational behavior of the UlS will be measured during Phase 2 System Operational Tests with sodium flowing through Core Special Assemblies at p.ny l motor flow to 100% and at temperatures of 4000F to 6000F. Significant flov induced lateral, vertical, and torsional structural vibrations will be measured by accelerometers located on the UlS.
Steady state temperature boundary conditions of the UlS will be measured using design vorifIcatf or. thermocouples during Phase 4 testing. The temperature I data obtained will be used to verify predicted temperature values used in the I structural analysis of the UlS. 14.1.4.5 CRBRP INTERMEDIATE HEAT EXCHANGER A. KEY FEATURE TESTING The design for the Intermediate Heat Exchanger (lHX) is based upon vendor tests confirmatory of design analyses. These vendor tests were key feature tests. Four specific tests were performed: (1) 30 Degree f} v Amend. 61 14.1-19 Sept. 1981
Model, (2) 360 Degree Model, (3) Intermediate Flow Test, and (4) Bellows Test.
- 1) The 30 Degree Model test was peformed to evaluate the shelI side flow characteristics and the possibility of tube vibration over the range of lHX operation. This test was an Isothermai test on a 30 degree, 20 ft. full scale segment of the tube bundle involving no heat transfer and using water as the testing medium. The results of the test showed the flow paths of the shell side fluid were as predicted analytically and that there was no tube vibration, therefore, confirming analysis of the tube bundle.
- 2) The 360 Degree Model test was performed to assure that there was uniform flow distribution in the Inlet region to the tube bundle.
This test was on a 0.3 scale model of the primary inlet plenum of the IHX, including the inlet piping configuration, using water as the testing medium. The results of this test provided the data for designing the inlet baffling to insure balanced flow distribution to the bundle region fran 7-1/25 to 100% flow.
- 3) The Intermediate Flow Test was performed to evaluate the flow distribution at tne intermediate inlet to the tube bundle. This test, using a 0.304 scale model and water as the testing medium, was designed spectfleally to evaluate the intermediate lower (Inlet) plenum flow dist.'Ibution. The results of this test provided the data for designing the baf fling to insure balanced flow distribution to the tubes at all flow rates.
- 4) The Bellows Test was performed to assure that the bellows on the downcomer would survive the design load cycling without failure. This test was a f atigue test which proved the adequacy of the bellows to at least 4 plant lifetimes without f ailure with a predicted capability to I withstand approximately 100 plant lifetimes. A squirm test was also i performed which assured the structural adequacy of the expansion members with respect to stabil!ty to approximately 2.7 times design pressure.
The results of the above testing confirms the IHX design adequacy. B. PREOPERATIONAL & STARTUP TESTING As an integral part of other plant testing and plant power operation, data will be gathered to verify the thermal per: rmance and Intermediate side pressure drop characteristics of the IHX. The thermal performance will be evaluated by measuring the power transferred tQ) and the log mean temperature dif ference (LMTD) and comparing that with the predicted heat transfer coof fIclent (UA) using UA = Q/LMTD. The intermediate side pressure drop characteristics can be 14.1-20 Amend. 61 Sept. 1981
inferred using the pressure taps at the Intermediato pump discharge and at the intermediate IXH outlet. This measurement, when calculated piping pressure O' losses between the pressure taps are deleteo, wilI be evaluated against vendor estimates of the Intermediate side pressure drop. The IXH leak tightness and Isolation of the primary system from the intermediate system will be demonstrated during the evacuation prior to initial sodium fill and in the sodium inventory observations during Phase 2 testing. 14.1.4.6 STEAM GENERATOR MODULE A. EEATURC TESTING The design of the Steam Generator vill be supported by several test programs designed to verify assumptions and provide quantitative data to confirm the adequacy of design analyses. These six tests are (1) the Hydraulic Test Model (HTM), (2) Large Leak Tests (LLT), (3) Few Tube Tests (FTT), (4) DNB tests (departure from nucleate boiling), (5) tube support wear tests and (6) material mechanical properties tests. See PSAR Section 5.5.3.1.5 for a description of these tests. B. PREOPERATIONAL AND.STARTUP TESTING A series of tests will be performed on the steam generator modules af ter they are Installed ai the site. These tests will be designed to show that s therraocouples are properly Installed, that they meet all the requirements for safe operation, and that they meet the expected performance requirments.
- 1. Preoperational Tests The position and alignment of each module will be checked af ter it is installed. The module will be checked for leak tightness on both the tube side and the shelI side before the sodium and water systems are filled.
The water side will be filled first and pressure testod in conjunction with the entire loop (the shell side of the steam ger.erator module will be pressure tested prior to installation). System tests of the water side will provide data on pressure loss vs. flow rate through the module at l temperatures up to 4000F Operability of the module isolation valves and l water dump and blowdown subsystem wilI be tested before the sodium side is filled.
. After alI of the IHTS and SGS components are heated to 4000F, the sodium side of the steam generator modules will be filled. System testing of the IHTS will provide data on pressure loss vs. flow rate through the shall side of the steam generator modules.
- 2. Startup Tests With the reactor operating, heat transfer and hydraulic performance date will be obtained at several powe- levels from zero power 14.1-21 Amend. 61 Sept. 1981 i
I -
to 1007 of rated power. These data wilI be used to verify the heat transfer capability and pressure loss calculations. System stability uncer transient conditions will be used to verify the heat transfer capability and pressure loss calculations. System stability under transient conditions will be demonstrated by changing power levels at the maximum planned rate. The objectives of these tests are to: a) Determine the overalI heat transfer coeffIclent and module pressure losses at rated power and operating conditions. b) Demonstrate stable operation at low power levels. c) Demonstrate stable operation at the maximum planned rate of change in power level. O l l Amend. 61 Sept. 1981
! O O O SCHEDULE FOR INITIAL TEST AND OPERATION ! (MONTHS + CRITICALITY) l l l l'
-105 -93 69 -57 -45 -33 -21 -9 !+l3 +15 +27 l
l Staffing ! l Key TVA Staff On-Site 7 l V ! ~ Bulk of Operating Staff On-Site i l
' Preop Test Crew Augmentation : I :
l l j Startup Test Augmentation : j : l % I RM/AE Technical Direction : : ! 6 l Test Program l l l Preoperational Tests : !: i I j Startup Tests : j : i i ! Procedure Preparation I 1 l j Test Plans and Specifications .: : i i Operating Procedure Outlines I
- Test Instructions : : l
- -@ m l .
FR Operating Instructions (Norm. & Emergency) :
. :l-
! O Surveillance Test Instructions : :l l a n enance ns medons : : 61 Figure 14.1-1 ,
14.2 AUGENTATION OF OPERATOR'S STAFF FOR INITI AL TESTS AM) OPERATION TVA's normal plant operating staf f, as described in Section 13.1.2, will be augmented during the initial tost and startup period. This augmentation will provide the operating staff with sufficient manpower to safely and effectively conduct the test program, as welI as perform those operations functions required during plant startup. The schedule for providing these additional personnel for augmentation Is shown in Figure 14.1-1. Regulatory Guide 1.58 and ANSI Standard N45-2.6-1978 wilI be used as a guide in develcping qualifications of augmenting personnel. The nucleus of the operational testing staf f will have had previous experience in testing TVA's light water nuclear plants. The augmenting personnel described below are in addition to those provided by 61l W-LRM and the A-E for technical direction as described previcusly in 14.1.3.1. During the preoperational testing phase the TVA plant operating staf f will be augrrented with a Preoperational Test Section. This section is an on-site group of TYA employees of the Division of Nuclear Power with the responsibit ity for the preoperational test program consisting of reviewing test specifications, writing test instructions, assuring that prerequisites 61 are satisfied, conducting the tests, evaluating the test results, and maintaining necessary records of those tests which demonstrate the functional performance and readiness of the various systems. This section is under the direct aupervision of the Preoperational Test Program Coordinator who reports to the Plant Manager on functional activities, and the assistant Director of O 61 Nuclear Power (Operations), of the Division of Nuclear Power, on administ.'ative activities. During the startup testing phase the plant technical staf f will be augmented 61l by technical personnel from IYA's Division of Nuclear Power. This includes technical support in nuclear, mechanical, chemical, instrumentation, computer, and general engineering. These technical support personnel wilI be under the functional supervision of the plant management, and administrative supervision 61l of TVA's Division of Nuclear Powe" central of fice. l l 14.2-1 Amend. 61 1 Sept. 1981 l I l
O V 15.1.2 Requirements and Criteria for Assessment of Fuel and Blanket Rod Transient Performance To assure that the CRBRP fuel and blanket rods will operate safely over their respective design lives the qualitative requirements of Table 15.1.2-1 have been implemented. 'pecifically, as discussed in Chapter 4, mechanical design limits (cumulative damage function and ductility limited strain) have been developed to assure cladding integrity is maintained through normal operation, all Anticipated Faults and the most severe Unlikely Fault. The complete details of cumulative mechanical damage function including both the theoretical derivat#on and experimental data base, are provided in Reference 53 of Section 4.2. For Extremely Unlikely Faults, limits have been establisned in terms of cladding and coolant temperatures tn conservatively ensure that core coolable geometry will be maintained. All these limits taken together ensure that the requirements of Table 15.1.2-1 are met. The acceptance criteria of Table 15.1.2-2 have been developed for the preliminary safety review (PSAR) to evaluate the acceptability of each transient analyzed relative to the requirements of Table 15.1.2-1. The use of the acceptance criteria allows preliminary assessment of each transient event without the need for detailed calculations of mechanical damage. Detailed calculations of mechanical damage will be performed for the final safety review (FSAR). The following subsections provide a brief description of the acceptance criteria of Table 15.1.2-2. 15.1.2.1 Acceptance Criteria for Anticipated and Unlikely Faults The preliminary acceptance criteria for the Anticipated and unlikely faults were established to assure that cladding integrity is maintained through-out all Anticipated Faults in the fuei (or blanket) lifetime and limiting-case Unlikely Fault. As indicated in Table 15.1.2-2, faults are considered acceptable if therg is no fuel melting and the maximgm cladding temperatures are less than 1500 F for Anticipated Faults and 1600 F for Unlikely Faults. These criteria not only assure that the 'fetime cladding perfnnance require-ments are met, they also assure a large margin to sodium boiling and to cladding melting, thereby assuring that core coolable geometry will be maintained. The bases for the preliminary acceptance criteria are the analyses of mechanical performance of fuel and bT . rods discussed in Chapter 4. Specifically, worst case umbrella transients corresponding to the thermal cord itions in the criteria have been combined at the required frequency at the worst time in life. In each case the mechanica' design limits, cumulative mechanical damage functicr. and ductility limited strain (see Section 4.2.1) have been satisfied. Therefore, if the Anticipated or Unlikely Fault produces lower fuel and cladding temperatures than those used in the corresponuing umbrella transients (and preliminary acceptance criteria), then the design performar,ce 61 and safety objectives of the fuel uo blanket rods are satisfied. p 15.1-50 Amend. 61 Sept. 1981
The actual mechanical damage to the cladding is a complicated function of temperature, stress (fuel and fission gas expansion), time at that stress, and accumulated irradiation (reduced strength and ductility). Therefore, violation of the preliminary acceptance criteria on fuel temperature, cladding temperature or both does not necessarily mean the transient has unacceptable damage. It does require the calculation of mechanical damage in accordance with the design procedures described in Chapter 4. The fault is unacceptable only if the correspond *,..g damage violates the mechanical design limits given in Section 4.1, or if the criteria for Extremely Unlikely Faults are not met. Should events occur which are more severe than the transient events used as the design envelope, aen the records of actual core environmental conditions will be utilized to determine the actual cumulative damage function; which would then oe compared to the design limits. 15.1.2.2 Acceptance Criteria for Extremely Unlikely Faults or Postelated-Accidents The allowable limit for an Extremely Unlikely Fault is defined as maintaining coolable geometry. As indicated in Table 15.1.2-2 events of this type are considered acceptable if the coolant temperature remains below boiling and the cladding temperature remains below melting. The basis for the acceptance criteria on Table 15.1.2-2 for these types of faults is that the geometry of the core must remain coolable following a faulted event to assure that damage will not progress. This limit is considered to be met when the cladding temperature is held below the melting point. If there is no cladding melting then no gross cladding relocation or gross channel blockage can eccur. Therefore, preventing
- ladding temperatures from exceeding the melting temperature will ensure maintaining a coolable core geometry.
Before the cladding melting temperature can be reached, it is necessary to first experience bulk sodium boiling and then dryout of the cladding. The prevention of sodium boiling is considered as a necessary and sufficient criterion for ensuring a core coolable geometry. 15.1.2.3 Acceptance Criteria Dependence on Shutdown Mode As noted in Table 4.2-35 in Chapter 4, the next higher level of damage is allowed for secondary shutdown system event termination. The rationale is that failure to actuate the primary shutdown system is a low probability event so that the combined probability of the event ocurring and secondary shutdown system activation being required is much lower than the probability of the event occurring. Therefore, application of the accep-tance criteria of Table 15.1.2-2 in the safety analyses reported in this chapter considers shutdown by the Primary and Secondary Shutdown System 61 actirn separately in a manner as described in Table 4.2-35. Amend. 61 15.1-51 Sept. 1981
TABLE 15.1.2-1 (m)61 w l EVENT CLASSIFICATION AND DAMAGE SEVERITY LIMITS Ivent Classification Severity level Mechanica. Oesign ROT Standard C-16-1 R3T Standard C-16-1 61I l (chapter 41 Nomal: Normal Operation: '40 Damage: Any condition of system startup. Nomal operation includes '.o damage is defined as 1) no design range operations, hot steady power operations and significant loss of effective standby. or shutdown other than those departures from steady fuel lifetime; 2) accomodations an upset, emergency. faulted or operation which are expected within the fuel and plant testing conditions. frequently or regularly in operating margins without the course of power operations, requiring automatic or manual refueling. maintenance, or prctective action; and 3) no maneuvering of the plant. planned release of radioactivity. Upset: Anticipated Faulted: Operational incident: Any abnormal incident not An of f-ncmal condition which An operational incident is causing a forced outage or individually may be espected defined as an occurrence which causing a forced outage for to occur once or more during m.ults in 1) nn redu tion of c which the corrective action the plant lifetime. effective fuel Iffetime below does not include any repair the design values; 2) accomo-of mechanical damage. dation with, at most, a reactor trip that assures the plant will be capable of returning to ,.,peration after corrective action to clear the trip cause; and/or 3) plant radioactivity releases that may approach the 10CFR20 guidelines. O Imergency: Wluely Faulted: Minor Incident: i I (/ Infrequent incident requiring shutdown for correction of An off-normal condition which individually is not espected A minor incident is defined as an occurrence which results in the condition or repair of to occur during the plant life- 1) a general reduction in the damage in the system. No tiv; however, when integrated fuel burnup capability and. at loss of structural integrity. over all plant components. most, a small fraction of fuel events in this category may be rod cladding failures; expected to occur a number of 2) safficient plant or fuel times. rod damage that could preclude resumption of operation for a considerable time and/or
- 3) plant radioactivity releases that may exceed 10CFR20 guide-lines. but does not result in interruption or restriction of public use of areas beyond the exclusion boundary.
Faulted: Extremely Unlikely Faulted: Major Incident: Postulated emt and conse- An off-normal condition of A major incident is defin( *s quences wher. ntegrity and such extremely low proba- an occurrence which results 4a operability muy be impaired bility that no events in this 1) substantial fuel and/or to the extent that censid- category are expected to occur cladding melting or distortien erations of pubile health during the plant lifetime, but in individual fuel rods,inut and safety are involved. which nevertheless represents the configuration remains extreme or liniting cases of coolable; 2) plant damage that 611 failur'$ "hich 8 *** Pr'ciud' " 5umptio" ' Pi*"t as design bases, ideatifi'd operations, but no loss of safety functions necessary to cope with the occurrence; and/or 3) radioactivity release that may exceed the 10CFR20 guidelines but are well within the 10CFR100 guidelines. s ( ~' ) 15.1-52 Amend. 61 Sept. 1981 L
TABLE 15.1.2-2 ACCEPTANCE CRITERIA FOR PRELIMINARY SAFETY EVALUATION Event Cladding Coolant SeveriV I4) Fuel Temperature Temperature Classification Level Temperature (*F) (*F) Anticipated Operational Solidus (1),(2) 1500 II) N/A Fault Incident Unlikely Minor Solidus (1),(2) 1600 II) N/A Fault Incident Extremely Major --- Solidus Saturation (3) . Unlikely Incident (2475) 7' Fault or $ Postulated Acc Hent NOTES: (1) For temperatures in excess of these values, transients shall be assessed using mechanical design procedures and design limits of Chapter 4.2. (2) No fuel melting at existing conditions. (3) No sodium boiling at existing pressure. (4) Applicable "E"ent Class"or " Severity Level" is based on Primary Shutdown System 61 action . For Secondary System Shutdown see Table 4.2-35. W h'
%8 = CL O O O
l a f 4 1
\
l I i ; 1 i i l CONTENTS OF
! PAGES 15.1-54 THROUGH 15.1-921 j !
l DELETED 61 i l i I i l l 15.1-54 Amend. 61 (next page is 15.1-93) Sept. 1981
15.2 REACTIVITY INSERTION DESIGN EVENTS - INTRODUCTION ( Q In the design approach to safety ducussed in Section 15.1.1 it was stated that the design in the second level emphasizes the need to insure and confirm the high reliability of the protection systems and of any component or system whose failure could lead to severe core damage. In keeping with this philosophy this section of the PSAR will examine the response charac-teristics of the reactor to 4. series of postulated reactivity insertion events. The reactor response to these events is identified through the resultant hot spot fuel pin cladding temperature. For these accident events either; 1) the resultant cladding temperature will be presented, or 2) it will be shown that the Plant Protection System will limit the reactivity insertion to a value less than a specified enveloping insertion. 51l Based on the discussion presented in Section 15.1.2, the severity of these events can only in-part be discerned by examining the resultant hot spot cladding temperatures. The overall severity of the event, as it effects the cladding integrity, is a function of the sum total of all the accumulated strains imposed on the cladding during its lifetime. There-fore, the severity of any event should be evaluated on a case by case basis using the cumulative damage function. In order to minimize the evaluation process and provide a ready determination of the relative severity of the event, the transients generated in this section can first 61 52 51 l be compared If the accidentto transient the umbrella fallstransient described within the time and Section 4.2.1.3.1. in temperature confines of the umbrella event, the conclusion can be ~ drawn that the design life and safety objectives of the fuel assemblies have been attained. If, however, the resultant clad temperature are beyond the time-temperature confines of the umbrella, then supplementary analysis is required to determine the severity of the event. The following conservative assumptions and conditions were used for the specific purpose of generating the worst case reactivity insertion tran-sients for this section.
- 1. All full power cases are for the reactor operating at thermal hydraulic design conditions with a power generation of 975 liWT at 3 loop operation. (Power uncertainties are discussed in Section 4.4.3.2).
- 2. Since the highest power fuel assembly and smallest Doppler coefficient occur at the beginning-of-equilibrium cycle (80EC) the transients are analyzed for th!s particular worst period in core life. .
- 3. With burnup, the power generation !ind steady state temperature decrease (flows are constant) in the fuel assemblies and conse-quently, the temperatures due to the transients would decrease. }
t 1 O V Amend. 61 15.2-1 Sept. 1981
- 4. The nominal Doppler coefficient for BOEC is - 0.0062 (see Sec-tien 4.3.2.3) however, for overpower transients it is nore con-servative to take the lower bouni value of the 20% uncertainty on this value. For studies in this section, -0.005 was used for the Doppler coefficient except where noted. (The exceptions being those cases where a larger Doppler yields more conservative results.) ,
- 5. Figure 4.2-93 of Section 4.2.3.1.3 of the PSAR includes 0.1 second unlatch time delay be'. ween start of CRDM stator current decay and start of the primary control rod motion. For preliminary Plant Protection System transient evaluation, a 0.2 second overall scram delay (see Section 15.2 and 7.2.1.2.3) has been assumed. This scram delay includes PPS logic, scram breaker and the unlatch time delay, leaving sufficient margin on overall PPS response time to assure conservative analysis.
- 6. The rod worths used to predict. Doct trip negative reactivity insertions are the design expected values for the primary control rods and the minimum expected values for the secondary control 61l rods. (See Section 15.1 for further details.) For both sets of control rods the single most reactive control rod is assumed to be stuck in the withdrawn position. At 80EC the primary con-trol rods negative reactivity insertion capability is less than any later time in the cycle. The purpose of these assumptions is to provide a realistic minimum prediction of shutdown reactivity and hence the slowest rate of power decrease. This provides a conservatively high prediction of reactor temperatures after shutdown.
5d 7. Three sigma (3a) hot channel factors were used for all the analyses and the temperatures shown are the inner surface of the hot pin cladding at the highest temperature position, both axially and circumferentially on the fuel rods. (Position is under the wire wrap). The possibility of additional fuel-cladding mechanical interaction during rapid reactivity insertion events is acknowledged as indicated in 61l Figure 4.2-22 and subsection 4.2.1.3.1.1. However, present models for transient fuel cladding mechanical interaction are admittedly lacking pheno-menologically at prototypic CRBRP design conditions, and therefore, are not used for PSAR analyses. The fuel models in codes used to calculate the effects of core disruptive accidents (e.g. SAS3A), are designed to give initial conditions for calculations of fuel motion from a ruptured rod, rather than to calculate detailed cladding responses to a terminated transient. These codes assume that cladding loads are simple functions fo fuel properties during the transient. Such assumptions are acceptable for detennining gross fuel rod behavior during a severe transient (e.g. , fail-no-fail), but are insufficient for calculating time varying cladding strain on cumulative damage function during a typical upset event. Amend. 61 15.2-2 Sept. 1981
i l O)
%vd '
D. ("L TABLE 15.2-1 REACTIVITY INSERTION DESIGN EVENTS l _ Max. Clad. Temp.* l Section , Primary Secondary No. Event Scram Scram Comments
~
15.2 Reactivity insert design events l .15.2.1 Anticipated Events 15.2.1.1 Control. assembly withdrawal 0 NA 1383 F Temp. shown for 1c/sec. withdrawal.
- i. startup (See 15.2.1.1) Resultant Temp. less than operating l
condition. (Full Power) 15.2.1.2 Control assembly withdrawal 0 f 1510 F 1610 F Based on extremely small withdrawal power rate - Results are within the guide-61l lines of Table 15.1.2-2. EN
'L 15.2.1.3 Seismic reactivity insertion 1440 F S1440 F Based on postulated 30c step reacti-
^
l d, (core, radial blanket and vity insertion - Results are within 61l control rod) - OBE guidelines of Table 15.1.2-2. 15.2.1.4 Small reactivity insertions 1500 F 1560 F For 2d/sec insertion case -- Results are within guidelines of i 61\ Table 15.1.2-2. 15.2.1.5 Inadvertc1t drop of single Less than Less than Results fall within guidelines of 61l control rod at full power init. cond. init. cond. Table 15.1.2-2. 15.2.2 Unlikely Events 15.2.2.1 Loss of hydraulic holddown 1415 F 1420 F Results are within guidelines of 61l . Table 15.1.2-2. pp mm 15.2.2.2 Core radial movement 1470 F 1510 F. For non-seismic conditions - Results fall within guidelines of
!* R 61l Table 15.1.2-2.- ,
O
- Fuel pin inside diameter cladding temperature (under wire wrap)
,- , - - ,- e
TABLE 15.2-1 Continued flax. Clad. Temp.
- Section Primary Secondary No. Event Scram Scram Coments 15.2.2.3 Mal-operation of reactor <l510 F <1610 F Less than limiting condition shown plant controllers in 15.2.1.2-1 15.2.3 Extremely Unlikely Events 15.2.3.1 Cold sodium insertion Less than Less than Results fall within the guidelines 61l init. cond. init. cond. of Table 15.1.2-2.
15.2.3.2 Gas bubbla .hrough core <1480 F <1480 F Results fall within the guidelines
^ * * ~2' 61l 15.2.3.3 Seismic reactivity insertion <l505 F NA Based on postulated 60c step reac-(core, radial blanket and tivity insertion - Results fall
% control red) - SSE within the guidelines of 7 61l aM e E l.2-2. a 15.2.3.4 Control assembly withdrawal NA 800 F For 20c/sec reactivity insertion - at startup-max. mech. speed (See 15.2.3.4) Results fall within the guidelines of 61l Ta bl e 15.1. 2-2. 15.2.3.5 Control assembly withdrawal 1420 F 1460 F For 20c/sec reactivity insertion - at power - max. mech. speed Results fall within the guidelines cf 611
- Fuel pin inside diameter cladding temperature (under wire wrap)
P&" Na
- F O O O
15.2.2.2 Sudden Core Radial Movement g (j 15.2.2.2.1 Identification of Causes and Accident Description The event to be c.o' dered here involves core radial motion which occurs rapidly and is diff ilt to accurately predict. This is in cr;.crast to normal core radial motion wMch occurs gradually and predictat<y in response to normal temperature changes and irradiation induced materiaF1 swelling and creep. TF latter type event is discussed in Section 4.2.2.4.1.8. 9 The type of sudden core radial motion to be evaluated has been termd
" stick-slip" motion. Stick-slip motion refers to a situation in which the reactor assemblies are restrained from moving radially by interassembly frictional forces at the assembly load planes (stick) and then suddenly move to a new position dictated by current temperature and irradiation environment as the interassembly frictional forces are suddenly removed or reduced (slip).
If it is postulated that sticking occurs while the reactor assemblies are bowed away from the core centerline, a sudden positive reactivity insertion can take place as the assemblies slip to an inwardly bowed shape (towards the core centerline). Such an event is unlikely since the buildup of inter-assembly frictional forces which would be required to cause sticking would occur only when the assenblies are in a compact inwardly bowed state. If the assemblies are bowed outward away from the core centerline, the interassembly gaps would be larger and then the probability of sticking would be miniv. On the other hand, if because of thermal and irradiation effects the ass m-blies tend to bow toward a compact state, theoretical compaction will not be
-m achieved due to manufacturing tolerances and frictional forces.
I \ U If the assemblies are prevented from achieving a compact state due to interassembly frictional effects, it is possible that a seismic event could overcome the frictional effects and allow the reactor assemblies to take on a more compact state. This is considered to be the only realistic initiating mechanism for a stick-slip type event. If the stick-slip event occurred, the reactivity insertion would cause temperature rises of the fuel, cladding, and coolant. The power rise would 611 trigger a primary control system scram if the limits of Section 15.1.3 were exceeded. 15.2.2.2.2 Event Evaluation: Model, Assumptions, and Conservatisms To determine the maximum possible reactivity insertion, the following analysis steps were followed:
- 1. Predict the difference in core assembly positions and bowing between refueling and full power.
- 2. Determine the reactivity worth factors associated with radial motion of each core assembly.
r v Amend. 61 15.2-43 Sept: 1981
- 3. From the predictions of maximum possible radial motion and worth factors, determine an upper limit for possible reactivity inser- .
tion from stick-slip. To predict the core assembly positions and bowing at refueling and full power conditions, a finite element model was constructed of a radial row of core assemblies. The reactor environmental conditions were then applied along with material characteristics to give bowing and position curves like those of Figures 4.2-88 through 4.2-92. Refer to Section 4.2.2.4.1.8 for further details 9 of the core assembly bowing analysis. Comparison of the bowing shapes for Fig-ures 4.2-88 through 4.2-92 shows an inward bowing at full power (100% power to flow ratio). The reactor assembld es were assumed to stick in the refueling position (at 07, power to flow raF o) and to then slip suddenly to the full power position. Conservative nominal compaction reactivity worth coefficients were determined by using the assumptions that all control rods would be parked above the core at the beg -ing of an equilibirum cycle. The worth coeffi-cients are shown in Table 4.3-14. The above procedure results in a prediction of approximately 60c for the maximum value of step reactivity insertion (see Section 4.2.2.4.1.8). 9 The above upper limi'. " considered to be conservative for the following reasons:
- 1. In the an'. lysis, all the gaps in the core were compressed com-pletely out whereas core compaction tests (1) indicate that not all gap <, will be compressed out in a real core. This is due to manufat.turing tolerances as well as frictional effects in the core.
- 2. The analysis assumptions were that sticking of the core asram-blies would occur where the assemblies are in their maxir outwardly bowed configuration. More realistically the s.ic' "
would not occur until substantial inwardly directed thermal bowing had already occurred and forces had begun t, bui' between assemblies. Thus, part of the bowing react ait, .- .> can be expected t, occur gradually which will be comp (ns .te_1 for by Doppler and taermal expansion effects. This woulo reduce the naximum possib'e step reactivity change.
- 3. The inherent vibrational motion of the core assemblies when flow is passing through would tend to prevent sticking. This should aid in allowing smooth translation of the core assemblies in response to thermal bowing.
- 1. W. C. Kinsel, "FTR Core Compaction and Withdrawal Tests," May 1973, HEDL-TME-73-58, UC-79 e, g, b.
Amend. 9 15.2-44 Dec. 1975
15.2.3.3 Seismic Reactivity Insertion (Feel, Radial Blanket and Control Rod (3 Assemblies) - SSE ( 15.2.3.3.1 Identification of Causes and Accident Description For the Safe Shutdown Earthquake (SSE) several conditions exist that compound the severity of the event. First, the earthquake can produce a loss of off-site electrical power causing a loss of power to the pumps and conse-quently a decay of the primary coolant flow. Second, the acceleration forces of the earthquake can cause compaction of the core due to closing of radial gaps between the assemblies at the above core load pad (ACLP) position. This can result in a net positive step reactivity insertion to the core. Third, when the control rods are scrammed, the rate of inward mction is decreased from the normal rate due to a retarding force resulting from seismic induced irpacts of the control rod assembly duct and driveline on surrounding guide structures. The first two subsystems initiate automatic reactor scram (the oper-ator can also initiate scram). Once power to the pumps is lost (it is assumed that che pumps remain operable following an SSE) a loss of electrical power (L0EP) trip system initiates scram after a 0.5 second delay. If a step reactivity insertion occurs the primary rods are scrammed when the reactor reaches 115% of full power. The worst combination of the above events with respect to core tem-peratures during the SSE would be to assume that a step roctivity insertion gS occurs 0.5 second after the power to the pumps is lost. If the step occurs ( ) earlier than this the : ore flow will be at a higher level when the control rod
"' insertion begins and the resultant temperatures for the event would be lower.
This is du! to the fact that for steps on the order of 30c the power rises very rapidly (in less than 0.1 second) and thus scram would be initiated by the 15% overpower condition almost instantaneously, If the step occurs later then 0.5 second the signal due to LOEP would have already started the plant protection system scram and the reactor power would be dropping below its initial value at the time of the reactivity insertion and again less severe temperature conditions would result. 15.2.3.3.2 Analysis of Effects and Consequences Since the highest power fuel assembly occurs at the beginning-of-equilibrium cycle (B0EC), the transients were analyzed for this particular worst period in core life. A minimum value Doppler coefficient of -0.005 was used for this core condition. This value is obtained by decreasing the nomi-nal Doppler coefficient by 20% for uncertainties as discussed in Section 4.3.2.3. Scram was taken at 15% overpower. [ Note: For the worst case chosen both the 15% overpower trip and the L0EP trip would cause the control rod insertion at nearly the same time as discussed in Section 15.2.3.3.1.] The maximum worth control assembly was assumed to be stuck and the shutdown worth was decreased by the appropriate amount. m kJ 15.2-77
Figure 15.2.3.3-1 shows the normal insertion rate and the seismic insertion rate for the primary rods. These insertion rates were calculated with the CRAB code (described in Section 4.4.3) usino the seismic force-time history curves in Section 3.9.3. As can be seen, at the 80EC the primary rods are banked such that some are 11", 23" and 36" withdrawn while at fu!1 power. Also, it can be noted that the earthquake causes only about a 0.32 second increase in scram time for an initially fully withdrawn control rod to be inserted its full 36". An analysis of various size step insertions occurring during the SSE has been performed with FORE-II (Ref. 1) (see Appendix A). The results of a parametric range of step intertions up to 90c are shown by Figure 15.2.3.3-2. Figures 15.2.3.3-3 to -6 show the variation in reactor oower and fuel assembly hot pin maximum fuel, cladding and coolant temperaturea (with 30 hot channel factors) . The size step insertion tnat would be expected to occur during the SSE has not been determined yet since it is a complicated study involving such effects as the thermal expansion of all the assemblies at their discrete temperatures, swelling of the assemblies, manufacturing tolerances and gap restrictions imposed by the core restraint system of the assemblies. This information goes into determining assembly bowing profiles and interassembly loads and gaps due to the earthquake forces. It can be seen from Fig-ure 15.2.3.3-2 that, even if a step insertion as large as 60c* should..cccur, the maximum cladding temperature would be less than 1505oF. The duration of the cladding temperature above its initial steady state value for this case would be about 1.0 second. Several runs for the SSE were also made for the highest power radial blanket assembly. For a 60f. step, the hot pin maximum cladding temperature was less than 1370 F. The radial blanket hot pin maximum temperatures for 60 and 90c steps are given by Figures 15.2.3.3-7, -8 and -9 for the fuel, cladding and coolant. The second temperature peak on Figure 15.2.3.3-8 is due to the slow release of internally stored heat in the radial blanket pins (i.e., radial blanket pins have a 0.52" 0.D. as compared to 0.23" for fuel pins) relative to the flow decay when the primary pumps are tripped. 61l As indicated in Section 15.1, parametric studies were performed to show the effect of using " minimum required" primary cont ol rod shutdown rate va'iues instead of the " expected" values (both having the highest worth rod assumed to be stuck). The temperatures described thus far in this section have been based on the expected rates of shutdown worth which, as described in this earlier section, are felt to give the more realistic evaluation of the transient. Figure 15.2.3.3-10 shows the fuel assembly hot spot cladding tem-perature for the two cases. As can be seen, for a 60c step the maximum tem-perature would increase from about 1505 F to 1570oF. As described in Sec-tion 15.1.2, the most limiting cladding temperature requirement for an extremely unlikely event is that no sodium boiling is allowed to occur. In order to achieve sodium saturation temperature during the SSE the cladding temperature must be in excess of 1700eF. oThe value calculated for FFTF under these conditions was 35c. O Amend. 61 15.2-78 Sept. 1981
15.3 UNDERC00 LING DESIGN EVENTS - INTRODUCTION Q Of particular importance to the safe operation of the CRBRP is the determination of the response characteristics of the reactor to a group of postulated undercooling events. The reactor response to these undercooling events is characterized, in this section of the PSAR, by the resulting fuel rod hot spot cladding temperature. For these accident events either,
- 1) the resultant fuel rod cladding temperature will be presented, or 2) it will be shown that the primary or secondary Plant Protection System trip will shut down the reactor before resulting plant temperature changes can be transported to the core. The impact of these Accident Events _on Plant Systems and components is less severe than the events presented in the Plant Duty Cycle List. olant components have been designed to provide 30 year life for the Plant Duty Cycles.
61l 51 l Based on the discussion presented in Section 15.1.2 a measure of the severity of these events can only in-part be ascerteined by the resultant cladding temperatures of any one event. The true severity of 4 ~ the event on the cladding integrity is a function of the sum total of all the accumulatea strains imposed on the cladding during its lifetime. Therefore, the severity of any event should be evaluated on a case by case basis using the cumulative damage function. In order to perform the evaluation process, the transients generated in this section are first compared to the guidelines 61l52151 l established in Section 15.1.2 and when necessary to the umbrella transients described in Section 4.2. If the accident transient falls within the time and temperature confines of the umbrella event, the. conclusion can be made that the design life and safety objectives of the fuel assemblies has been
' (m) v conserved. If however, the resultant cladding temperatures exceed the 61l52B1 I guidelines limits of Section 15.1.2 then supplementary analysis is required to determine the severity of the event.
The fc11owing is a list of the Thermal-Hydraulic initial conditions used for the accident events presented in this section; Themal Hydraulic Conditions Thermal Power (MWT) 975** Primary Flow (LB/Sec/ Loop) 3842 Primary Hot Leg Temperature ( F) 1015* Primary Cold Leg Temperature (*F) 750* Intermediate Flow (LB/Sec/ Loop) 3555 Intermediate Hot Leg Temperature ( F) 956* Intermediate Cold Leg Temperature ( F) 671* Hot Spot Clad Midwall Temperature ( F) 1365
*These values include an additional 20 F over their nomal value to allow for instrument error and control dead band allowance.
** Power uncertainties are discussed in Section 4.4 for 3 loop operation.
V . Amend. 61 15.3-1 Sept. 1981
Supplementing the above parameters, the following additional conservative assumptions and conditions were used for the analysis;
- 1. Maximum Decay Heat - The decay heat fo the end-of-cycle condition corresponding to long term power operating history at full power was used. This included an added 25% conservative 2a bias to cover uncertainties. The purpose was to provide maximum post-trip heat input to provide a conservatively high preaiction of core maximum temperature and a conservative evaluation of heat input to the decay heat removal system.
- 2. Most rapid flow coastdown - The minimum vendor specified sodium coolant pump inertia and maximum system pressure drop are combined to generate a conservatively fast rate of flow reduction following a coolant pump trip. The purpose of this assumption is to provih a minimum prediction of net reactor coolant flow during the period from pump trip to the time of reaching pony motor flowrate. This results in minimum heat removal from the reactor during this period and hence a conservative maximum prediction of core temperature.
- 3. Full power thennal hydraulic design condition operating points - The full power thermal-hydraulic rated condition is at 975 MW reactor power. The thermal-hydraulic design operating temperatures have been conservatively increased by 20 F to allow for instrument error and the control dead band.
The purpose of this assumption is to assure the nost conserva-tive prediction of severity for the events analyzed. The additional temperature bias for instrument error increases the conservatism of predicted reactor temperatures.
- 4. Shutdown Rod Worths with Maximum Worth Single Stuck Rod - The rod worth used to predict post trip negative reactivity insertions are the design expected values for the primary shutdown system control rods and the minimum expected values for the secondary shutdown system control rods (see Section 61l 15.1 for further details). For both sets of control rods, the single most reactive control rod is assumed to be stuck in the withdrawn position. The purpose of this assumption is to provide a realistic minimum prediction of shutdown reactivity and hence the slowest rate of power decrease. This provides a conservatively high prediction of reactor temperatures after shutJown.
- 5. A canservative 200 millisecond delay between the trip signal and the control rod insertion was used for these analyses.
In Section 4.2.3 of the PSAR the requirement for che scram speed is that this delay be less than 100 milliseconds. The additional 100 plus millisecond delay over the required value results in higher clad temperatures and thus a worse condition. Amend. 61 15.3-2 Sept. 1981
im h The effect of using " minimum required" primary control rod shutdown rate values instead of the " expected" values (both having the highest worth rod assumed to be stuck), is also shown in Figure 15.3.1.1-2. As indicated 61 l in Section 15.1 the core temperatures described in this chapter for the primary system have been based on the expected rates of shutdown worth which give the more realistic evaluation of the transient. The secondary rod insertion rates used are the minimum rates. Figure 15.3.1.1-2 shows the hot spot cladding temperatures for the two cases. As can be seen, there would be about a 10 F increase from using the minimum rates. Thus, using minimum instead of expected primary rod insertion rates does not significantly change the nature of or effects of the transient. I 15.3.1.1.3 Conclusions The loss of off-site electrical power results in a simultaneoas loss of sodium pump power and the consequent reduction in core flow. The primgry shutdown system limits the clad midwall hotspot temperature to 1410 F. In the unlikely event that the primary shutdown system does not operate, the secondgry shutdown system limits the hot spot midwall- clad temperature to 1630 F. This is an acceptable result because analysis of 52 the transient has shown that the cladding damage (cumulative damage function) 51l does not exceed the limit for an emergency event. O O Amend. 61 15.3-7a Sept. 1981
Thermal Analysis - Fuel Redistribution Outside Duct 59l 3. The accident sequence has been further extended to investigate the consequences of a loss of fuel assembly integrity. It was hypothe-(]/
- q. sized that fuel particles might leave .the fuel assembly duct, fall down in the annular space between hexagonal housing and circular CCP, and accu-mulate at the CCP bottom. Due to considerable geometrical distortion of the fuel assembly near the fueled region (from overtemperature during this event) and the presence of solidified, previously molten, material from the fuel assembly duct and cladding near the (colder) CCP wall, only a restricted passage for fuel particles will exist. Only a small amount of fuel material would therefore be expected to fall to the bottom of the CCP. 25% of the fuel material was judged to be the upper limit of this amount. However, the value was varied up to 100%, to show the effect of this parameter.
The calculated peak t',usient temperatures in the fuel, CCP, and nearest seal are plotted in Figure 15.7.3.1-8 for these amounts of fuel present at the CCP bottom. After an initial drop of the fuel temperature due to the fuel relocation in a cold area, the fuel temperature rises slowly. The peak CCP temperature at the CCP side and bottom, and the temperature of the nearest seal (lower cold wall) also rise slowly. The calculations show that at about 1.5 hours later initiation of the event, i.e., after loss of sodium from the CCP, the transient temperatures in the fuel and CCP reach steady-state conditions if 25% of the fuel fragments are accumulated at the CCP bottom. The steady-state temperatures are as follows: Center of Fuel 3140 F CCP, Bottom 1890 F CCP, Side 18650F Lower Cold Wall Seal 260 F The movement of fuel particles from the original fuel region within the fuel assembly to the CCP bottom has the beneficial effect of
- lowering the energy density of the heat source and thereby lowering the temperatures of the fuel and its surrounding. This explains the lower temperatures when 25% of the fuel has accumulated in the CCP bottom. A stress analysis indicated that the stresses in the CCP, due to support of its own weight and that of the fuel assembly, are very low. The tensile stress in the tubular part of the CCP is 230 psi, the compressive stress at the CCP bottom is 700 psi. This compares to an uftimate strength of about 6000 psi for the CCP material (SS 304) at 1900 F.
Fission Product Release Analysis-The analysis and the supporting temperature data presented above show that the postulated accident will not lead to any fuel melting, but 29 Amend. 59 15.7-20d Dec. 1980
could lead to extensive clad melting. It can be conservatively estimated that most of the figsion progucts which are volatile in the temperature range of about 2800 to 3500 F are released into the EVTM. This tempera-ture range corresponds to the maximum axial steady-state temperature which the fuel rods in the assemblies reach, dependent on their radial location (see Figure 15.7.3.1-5). Table 15.7.3.1-3 lists those fission product elements contained in a fuel assembly of the equilibrium cgre at the end of cycle which are in the molten or vapor phase below 3500 F. The entire isotopic content of fission products is given in Table 12.1-35. The fission products of Table 15.7.3.1-3 are assumed to be released into the EVTM either partially or completely, depending on their melting points and partial pressures. The maximum cold wall temperature of the EVTM was calculated to be 435U F (see Figure 15.2.3.1-5). This " hot spot" is at an axial location coresponding to the midplane of the fueled region in the fuel assembly. The nearest seals are 6.3 ft downwards at the lower end of the cold wall near the air inlet module. These seals will not reach temperatures higher than 260 F during this accident. The elastomer seals will contain the radioactive fission products in the EVTM. Permeabigities of elastomeric seals have been experimentally determined up to 300 F (see Reference 1 of Section 15.5.2.3). Itwasthereforeconcludedtgatallfissionproductswhicharein the liquid or gaseous phase above 260 F are plated out on the cold surfaces e in the EVTM, specifically at the cold wall and/or near the seals. Only fisgion products which are in the liquid or gaseous phase at or below W 260 F were considered to leave the double seals by diffusion. The diffusion rates of fission products from the EVTM to the RSB/RCB are given in Table 15.5.2.3-3. In determining these diffusion rates. git was assumed tgat about 15% of all EVTM seals are at a temperature of 300 F and 85% at 150 F. This assumption is conservative with respect to the postulated accident, since only one set of seals representing about 1% of all EVTM seals, could exceed a temperature of 150oF. The diffusion rates of Table 15.5.2.3-3 are therefore higher than those which would be expected as a result of the accident discussed here. Fission products other than those listed in Table 15.5.2.3-3, but which are volatile at EVTM seal temperatures, were discussed in the response to Question 001.212. As shown there, only Cs and Rb need to be considered, yet the radioactivity contribution of all Cs and Rb isotopes combined, passing through the hottest EVTM seals, is smaller than that of all other volatile fission product isotopes (i.e. mainly of Xe 133,1131, and 1132) by a factor of approximately 105 at 36 hr. after reactor shutdown, and by a factor of more 61l 59 than 103 at 80 days af ter reactor shutdown. Based on the above considerations, the radioactivity leakage from the EVTM to the RSB/RCB due to the postulated accident will be less 29 than, or is enveloped by the leakage presented in Section 15.5.2.3. Amend. 61 15.7-20e Sept. 1981
16.6 ADMINISTRATIVE CCNfRCLS 16.6.1 QIranization
- 1. We plant manager has onsite responsibility for the safe operation of the facility and shall report to the Assistant Director of Nuclear Power (Operations). In the absence of 61 the plant manager, the assistant plant manager will asstane his responsibilities.
- 2. %e portion of W'A management line of responsibility which relates to the operation of the plant is shown in Figure 16.6-1.
- 3. %e fenctional organization for the operation of the plant shall be shown in Figure 16.6-2.
- 4. Shift manning requirenents shall, as a mininun, be as delineated in Section 16.6.8.
- 5. Qualifications of the OBRP management and operating staff shall meet the mininun acceptable levels as described in ANSI /ANS-3.1-1978, Selection and Training of Nuclear Power 61 Plant Personnel, dated January 17, 1978.
- 6. Retraining and replacenant training of plant personnel shall be in accordance with ANSI /ANS-3.1-1978, Selection and 1
Training of Nuclear Power Plant Personnel, dated January 17, 1978. %e mininun frequency of the retraining progra shall be every two years. 16.6.2 Review aM AvHt
%e Manager of Power has delegated responsibility to the l Nuclear Safety Review Board (NSBB) to monitor the Plant Operations Review
! Cmunittee activities and to ensure the proper operational safety review by off-site personnel who have no direct responsibility for plant operaticos.
%e Office of Power Quality Assurance and Audit Staff has the responsibility for the plant audit functions.
16.6.2.1 melaar Rafety Review anard he Nuclear Safety Review Board (NSRB) advises the Manager of Power on the adequacy and implenentation of WA nuclear safety policies and progras and assures that them policies and progres are in compliance with NRC regulatory requirenents. In general, the review and investigation 61 functions are perfonned irdeperdently of NSRB. %e NSRB is responsible for evaluating the results of such activities to determine that ell nuclear safety-related aspects are being adequately considered. In addition, the NSRB may conduct reviews or investigations of any nuclear safety-related activity in order to evaluate the WA nuclear safety progrm. O Amend. 61 16.6-1 Sept. 1981
%e NSRB is cmprised of a chairman and at least rive other mmbers appointed or approved by the Manager of Power. Mmbers of the NSFL may be frm the Office of Power, or other WA organization or external to WA. %e NSRB meets on a periodic and as-required basis to perform those 61 functions identified above.
Mmbership requirments and responsibilities of the NSRB will be defined in the Nuclear Safety Review Board Charter and the Nuclear Safety Review Board Guidelines, both of which are formally approved by the Manager of Power. 16.6.2.2 _ Office of Power G mlity Assurance and Audit Sh ff 61 he Office of Power Quality Assurance and Audit (OA&A) Staff 61' Organization will be responsible for assuring the imp 1 mentation and maintenance of an effective quality assurance program, including the auditing of all safety-related activities of the CRBRP. % rough the audit progrm , existing and potential deficiencies are identified and appropriate corrective actions are assigned. n rough formal audit reports, the Nuclear Safety Review Board and Manager of Power are advised of any identified deviations from procedural requirments and licensing comnitments.
%e Quality Assurance and Audit Staff Organization is comprised l
al of two sections plus a number of Quality Assurance Representatives who are resident in the operating nuclear plants and report directly to the QA&A Manager on the status of in-plant quality assurance. Its functional arrangment is shown in Figure 16.6-4. l 16.6.2.3 Plant Ooerations Review Cerrnittee (PORC) O l 1. Mathershin-The PORC shall consist of the plant manager, assistant plant manager, maintenance supervisors, health physics supervisor, operations super"!sor, plant engineering supervisor, and Supervisor, Quality Assurance Staff. An assistant plant supervisor may serve as an alternate cannittee mmber when his supervisor is absent; however, no more than two alternates shall participate as voting mmbers in PORC 61 activities at any one time.
%e plant manager will serve as chairman of the PORC. We e.ssistant plant manager will serve as chairman in the absence of the plant manager.
- 2. Meetino Frecuency
%e PORC shall meet at regular monthly intervals and for special meetings as called by the chairman or as requested by individual msbers.
O 16.6-2 Amend. 61 Sept. 1981
- 3. morum O h chairman or his designated alternate, plus four of the other members, or their alternates, will constitute a quorum. A member will be considered present if he is in 61 telephone communication with the committee. The msnber who is absent and whose alternate has not been provided must not be that menber having principal responsibility or expertise in the area being reviewed.
- 4. Duties and Rosmannihilities
'Ihe PORC serves in an advisory capacity to the plant managet and as an investigation and reporting body to the Nucleu Safety Review Board in matters related to safety in plani.
operations. h plant manager has the final responsibility in determining the matters to be implemented ancVor referred to the Ibclear Safety Review Board. The responsibilities of the ccanittee will include:
- a. Review all standard and energency operating and maintenance instructions and any proposed revisions thereto, with principal attention to provisions for safe operation.
- b. Review proposed changes to the license and Technical 61 Specifications.
- c. Review proposed changes to equignent or systems having safety significance, or which may constitute "an unreviewed safety question," pursuant to 10CFR50.59. i
- d. Investigate reported or suspected incidents involving safety questions, violations of the Technical Specifications, and violations of plant instructions pertinent to nuclear safety.
- e. ew repoMle occuneces, esM ets, operadng 61 anomalies, and abnormal performance of plant equignent.
- f. Maintain a general surveillance of plant activities to identify possible safety hazards.
- g. Review plans for special fuel handling, plant maintenance, operations, and tests or experiments which may involve special safety considerations, and the results thereof, where applicable,
- b. Review adequacy of quality assurance progran and reconnend any appropriate changes.
- i. Review adequacy of Technical Spec:ifications and recumaid any appropriate changes.
16.6-3 Amend. 61 Sept. 1981
61 l j. Review unit operations to detect potential nuclear safety hazards.
- k. Review all proposed tests and experiments that affect 61 I nuclear safety.
- 1. Review the site Radiological Dnergency Plan and the Plant 61 Physical Security Plan.
- m. Review adequacy of sployee training progres and reconnended changes.
- n. Review every unplanned onsite release of radioactive material to the environs.
- o. Review changes to the Radwaste Treatment systm.
t
- p. Review meeting minutes of the Radiological Assessment Review Ccenittee (RARC) .
61
- q. Performance of reviews as requested by plant manager.
- 5. Authority The PORC shall be advisory to the plant manager.
- 6. Records Minutes shall be kept for all PORC meetings with copies sent to Director, Division of Nuclear Power, Assistant Director of Nuclear Power (Operations), and Chainnan of the Nuclear 61 Safety Review Board.
- 7. Procedyres Written administrative instructions including applicable check-off lists prepared and maintained describing the method for sutinission and content of presentations to the ccrmittee, review and approval by menbers of connittee actions, dissmination of minutes, agenda and scheduling of meetings.
16.6.3 Instructions A. Detailed written instructions, includirg a@licable check-off lists covering itms listed below shall be prepared, a @ roved and adhered to.
- 1. Nennal startap, operation, and shutdown of the reactor and of all systes and cmponents involving nuclear safety of the facility.
O 16.6-4 Amend. 61 Sept. 1981
- 2. Refueling operations.
d
.3. Actions to be taken to currect specific and foreseen potential malfunctions .sf systens or components, including responses to alanns and abnormal reactivity changes.
- 4. Bnergency conditions involving potential or actual releam of radioactivity.
- 5. Preventive or corrective maintenance operations which could have an effect on the safety of the reactor.
- 6. Surveillance and testing requirerents.
- 7. Radiation control procedures.
- 8. Radiological Dnergemy Plan inplenenting procedures.
- 9. Plant security program implenenting procedures.
B. Written procedures pertaining to those itens listed above shall be reviewed by PORC and aIproved by the plant manager prior to implenentation except that ter.porary changes to procedures which do not change the intent of the original procedure may be made with the concurrence of two persons holding senior reactor operator licenses. Such tenporary O changes shall be documented and subsequently reviewed by PORC and approved by the plant manager. C. Drills on actions to be taken under energency conditions involving release of radioactivity are specified in the radiological energency plan and shall be conducted annually. Annual drills shall also be conducted on the actions to be taken following failures of safety-related systens or components. D. Radiation Control Prendures l l Radiation control procedures shall be maintained and made available to all plant personM.. %ese procedures shall show permissible radiation exp>sure and shall be consistent with the requirenents of 10GR20. his radiation protection progran shall be organized to meet the requirenents of 10CFR20. %e project is proposing, hcwever, that the provisions of Section 20.203(c) subparagraphs (2), (3), and (4) apply only to areas where the radiation levels are continuously 1,000 mReg ht or greater. Amend. 61 16.6-5 Sept. 1981
. _ . ~. _ . _
61 l 16.6.4 Actions to be Taken in the Event of ReDorhhle Occurrence in Plant Operation A. Any reportable occurence shall be prmptly reporced to the Director, Division of Nuclear Power, and the NSRB, and shall be prmptly reviewed by PORC. Wis comittee shall prepare a separate report for each reportable occurrence. mis report shall include an evaluation of the cause of the occurrence and reccanendations for appropriate action to prevent or reduce the probability of a repetition of the 61 occurrence. B. Copies of all such reports shall be sutnitted to the Assistant Director of Nuclear Power (Operations); Manager, Nuclear Regulations & Safety; Chief, Radiological Hygiene Branch; Supervisor, Nuclear Safety Review Staff; and the 61 01ainnan of the NSRB for their review. C. We plant manager shall notify the NRC within 24 hours, as specified 11. Specification 16.6.7 of the circumstances of any abnonnal occurrence. A written report shall follow within 10 days. 16.6.5 Action to be Taken in the Event a Safety T.imit is Exceeded If a safety limit as defined in 10CFR50.36(c) (1) (i) is exceeded, the reactor shall be shut down and reactor cperation shall not be resumed until authorized by the NRC. A prmpt report shall be made to the Director, Divisicn of Nuclear Power and the Chairman of the NSRB. A complete analysis of the circumstances leading up to and resulting frm the situation, together with reccmnendations to prevent a recurrence, shall be prepared by the IORC. Wis report shall be sutaitted to the Assistant Director of Nuclear Power (Operations); Chief, Radiological Hygiene Branch; Manager, Nuclear Regulation & Safety; Supervisor, Nuclear Safety Review Staff; and the NSRB. Notification of such occurrences will be made to the NRC by the plant manager within 24 hours as specified in Specification 16.6.7 followed by a written report within 10 days to the Director, Office 61 of Management Information and Progra Analysis US NRC. 16.6.6 Station Operating Records A. Records ancVor logs shall be kept in a manner convenient for i review as indicatv. below:
- 1. All normal plant operations including such itm s as power level, fuel exposure, and shutdowns.
- 2. Principal maintenance activities.
- 3. Abnormal occurrences.
O 16.6-6 Amend. 61 Sept. 1981 e
- 4. Checks, inspections, tests, and calibrations of components and systens, including such diverse itens as
; source leakage. ;
)'
- 5. Reviews of changes made to the procedures or equipnent or reviews of tests and experiments, to comply with 10CFR50.59.
i 6. Radioactive shipnents. l 7. Record of annual physical inventory verifying accountability of sources or record.
- 8. Gaseous and liquid radioactive waste release to the environs.
- 9. Off-site enviror1 mental monitoring surveys.
I
- 10. Fuel inventories and transfers.
- 11. Plant radiation and contamination surveys. ,
- 12. Radiation exposures for all plant personnel.
! 13. Updated, corrected, tnd as-built drawings of the plant.
- 14. Minutes of meetings of the Nuclear Safety Review Board.
B. Except where covered by applicable regulations, items 1 , through 6 above shall be retained for a period of at least 5 years and itens 7 through 14 shall be retained for the life of the plant. A complete inventory of radioactive materials in possession shall be maintained current at all times. 16.6.7 3mprtina Damir-tm i A. Routine and Reportable Occurrence Reports to NRC i Information under this category to be reported to the NIC
- l. includes the following:
- 1. Reports required by Title 10, Code of Federal Regulations.
I
- 2. Reports of radioactive discharges and radiological monitoring which have been transferred, per directive
- from NRC, fran Appendix B to Appendix A of Regulatory Guide 1.16.
l
- 3. Reports required by the current revision of Regulatory Guide 1.16, " Reporting of Operating Information -
l Appendix A, Technical Specifications", as applicable to liquid metal fast breeder reactors with exceptions to be determined. 61 16.6-7 Amend. 61 Sept. 1981
B. Special Reports to NBC Special reports to the NRC are required covering inspections, tests, and malatenance activities, and nonroutine activities which ar6 specified in parts of Title 10, Chapter 1, Oxle of Federal Regulations. Specific requirments will be included in the FSAR. C. Environmental Monitoring Reports to Environmental Protection igency (EPA) Reporting information to the EPA concerning non-radiological environmental surveillance and enviromental impact will 61 follow the guidelines and requirments of the NPDS permits issued to the CRBRP. 16.6.8 Minim m S w fina A. Table 16.6-1 shows the nunber of shift personnel whenever the plant is not at shutdown or reteling shutdown conditions. B. A licensed senior operator shall be present at the site at all times when there is fuel in the reactor. C. A licensed operator shall be in the control rom when the reactor contains a potential critical mass. D. A licensed senior operator shall be in direct charge of a reactor refueling operation; i.e., able to devote full time to the refueling operation. E. A Shift Technical Advisor shall be onsite at all times except when the reactor is at refueling t m perature. F. A health physics technician shall be present at the facility at all times when there is fuel in the reactor. G. 'IWo licensed operators shall be in the control room during startups, while shutting down the reactor, and during recovery frm any plant trip. H. Either the plant manager or the assistant plant manager shall have acquired the experience and traihing normally required for examination by the NPC for a Senior Reactor Operator's License, whether or not the examination is taken. In addition, the operations supervisor or assistant 61 operations supervisor shall have an SFO license. Amend. 61 O 16.6-8 Sept. 1981
i 61 TABLE 16.6-1 MINIMJM SHIFT CREW REQUIREENTS Shift Position Members Type of License Shift Engineers (SE) 1 SR0(a) 61 l Shift Technical hdvisor(STA) i None Assistant Shift Engineers (ASE) 1 SR0(6) q Unit Operators (UO) 2 R0(b) Assistant Unit Operators (AUO) 2 None Health Physics Technician 1 None 61
"**
- 0" l
(a) SRO - Senior Reactor Operator (b) RO - Reactor Operator O Amend. 61 ! 16.6-9 Sept. 1981 l
O MAN 7GER OF P W ER DIVISICN CF NUGEAR PCWER I, ASSISTANT DIRECIOR OF NUGEAR PCHER (OPERATimS) O NUGEAR PLANIS (INCLUDING CRBRP) Figure 16.6-1, 'IVA Office of Power Organization Line of Responsibi.lity. for the CRBRP 6738-1 16,6-10 Amend. 61 O Sept. 1981
i PLANT NutEIR (I I Qualf ty Assurance Ass't Plant Supervisor (1) Manager (1) I QA tafety 5taff (4) ll Engineering (?) lI I Plant Service g Sta f f (3) s, NUC PR '- Central Office I t fin p C fi*l ; tun'hrtCt ( EINttRING Operation I"
- roPf-mlP, t.m :
r----------a,..-....._. i sm vi= (I) s.pereiso. q f to t, tn ( litst N T ! Ctter (Adm. ser, ices) (- As sis t .,nt
- - - - - - - - - - 8 7 Ytsi tw' es j ,
Operkt IV s Superv isor g I
- Chemical _ aeactor Mechanical Engineer (t) Engineer (1) Engineer (1) { Shif t tagineers Chemical *leuclear flechanical -
~
(ngineers (2 Engineers (7) Engineers (2) Shif t Cheutcal Engineerin9 tagineers tagineer {-tn91neerin9 Alde Alde - Trainees (2) ($MI (II ($tatistics) (g) Unit Chemical Engineer ~ Operators Engineer Associate (g) Associate (1) Ass t s t ent Cheetcal Enginee rin9 Uni t Labora tory a.4,5 Analysts (6) I8I ,', opera'orsJ I
I i J' I L l
^^" nt - s a t-
_ (Ae tces Store _s - . . sal se etces
<5a I Plant Security Wt leestth Physits Services (1)
(1) i C t (1) ser a fit t i Office Staff ggy
- i. d4J) I list I ful I perchanical Electrical lastruerat
, Itaintenance fes ta tenante ne g at en.ac,
- 1) Supervisor (t) Surervisor (1) Superviser (1)
~' CIIrt i Clert - (Adm. Ses.) i (Asn. Sus.) l l Assistant Electrical Ass Maat lastrument Assistant ~ ' g) Supervisor Supervisor fagineer Supervise + s I"9' " gag
-., c.) e n,, nee,, n, I
, i H Enginaer i Aides I I Crafts Electrica8 -(ng*r. Instrument tiert
( Ak Sus ) 1I [ l - Toremen su)T(5: y,,,,,, (I) aide (II foremen (3) ~ (4) Administrative Supervisten g g ~~ (
- Shif t Technical Engineer Clert $r. Instr.
,_ Craf ts"* Personnel are included in I#d*' $' "*"'
l[g) (M) the thsclear [ngiacer Group Electricians (in (37]
** Isot included in the I Wrers Plant Penusacnt 5taff. Jr, g,s tr, Mrs hanic s
[g3) (8) (s. ni i ( f l FIGURE 16.6-2 CRBRP Organization Chart 1 (1yearafterpower l operations) i 6717-1 'i i 16.6-11 Amend. 61 Sept. 1981 r . L f t I a
Tl Manager of Deput Power ops 09 Ass't M'gr Poder ops l Division of Division of Division of Division of PWR Sys ops Fossil & Hydro Nuclear Power Energy Cons & RT l l Nuclear l Plants Including CRBRP l Asg Di d
T i i J WESSEE VALLEY AUTHORITY Manager of Power l l 1 l ) Manager Ass't Manager Ass't M'gr Manager of
-Powar of Power of Power Power Engin'g
! Finance, Regulatory, Budget, Control, Personnel Power Plannin! Safety Review Staff Division of Division of TR Division of Division of lanning & Fuels u Power Power Constr. Engineering Utilization I. Manager Division of Energy ? Power Demonstrations & Tech. )tricts FIGURE 16.6-3 Office of Power Organization (TVA) i 16 6-12 Sept. 1981 1 1 I y o
.j
i i 4 l @ i w 4 W 1
) oo 4 1 MANAGER OF POWER l
l QUALITY ASSURANCE AND AUDIT i STAFF i r i 4 1 1 1 w m J - m l L f QUALITY AU0lTS QUALITY PLANNING l S ECTION AND ENGINEERING' 1 . S ECTI ON 1 l
- CENTRAL
--- OFFI_CE _ _ __ __ _._. _. _. _
j CRBRP SITE i l 4 IN-PLANT QA & A 4 COORDINATORS 1 i
$N l
j 3$
- CL Figure 16.6-4. Quality Assurance and Audit Staff Organization
{ @m CO H b W i
i i l l AMENDMENT 61 LIST OF RESPONSES TO NRC QUESTIONS There are no new NRC Questions in Amendment 61. O G Q-i
1 4 1 Question 001.268 (DS.2.2.2) Provide an explanation of the apparent inconsistency between using a fuel cracking model in SAS which forces fuel against cladding (see Page 05-3) and the assumption that 50% of the maximum axial fuel expansion ' reactivity effect can be used in the LOF analysis. Include in your explanation fluence and irradiation effects, in particular, those that might affect relative motion at the cladding-fuel interface.
Response
The CRBRP Project has consolidated all considerations given Hypothetical Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b,
, PSAR Section 1.6) and its associated references; consequently, PSAR 1
Appendices D id F have been withdrawn in Amendments 24 and 60 respectively . The response ., this question is now found in Section 2.2 of Reference Q001.268-1. 60h1
Reference:
Q001.268-1: W. R. Bohl, J. E.' Cahalan and D. R. Ferguson, "An Analysis of the Unprotected Loss-of-Flow Accident in the Clinch River Breeder Reactor with an End-of-Equilibrium-Cycle j Core", ANL/ RAS 77-15, May 1977 (Availability: U. S. DOE Technical Information Center). 61 l l l l l l l [ Amend. 61 l Q001.268-1 Sept. 1981 i l l i
, . , , . . . , - ~ . . , , , - , ..w. . , . . . , , . , , - , ...,,.-,,n, ,,_.,,,,,,n.an.....-,-- .,.,,,_,..,,,.,,-n...,- ..,,,-,-n,n, , - , .
Question 001.451 (F6.2.3.4.3) Based on the statements that calculated fuel temperatures are unrealis-tically high and that "This condition would be conservative...", justify this conclusion considering the following: a) Early failures might be beneficial to the outcome of the accident; b) Assuming use of the burst failure criterion, the more restructuring, the higher the pin is likely to fail; c) The more restructuri~ng, the less transient fission gas is available for pressurization; d) The unrealistic stimulation of early ' failures may introduce inco-herence which would make the outcome non-conservative; e) Fuel-pellet clad contact (i.e., zero gap) may enhance local boiling and/or local secondary loading failures. Response:
- l60 In order to clarify the statement that for the CRBR E0EC core represen-tation in SAS3A, the calculated steady-state fuel temperatures are high and this would yield conservative predictions concerning transient behavior an E0EC TOP case was run where the minimum and maximum no-O pressure watts cm"gonductjon V e 0 gontribution sec to,,the can conductance was set at 2 (3522 Btu ft-' hr -lDF -I). The radiation and contact gap conductance contributions emDloyed the same algorithm as the E0EC TOP base case. Figure Q001.451-1 shows the highest-power channel steady-state axial temperature profiles. As can be seen from comparison with Fig. 5-18 of Ref.15, PSAR Section 1.6, the temperature drop across the l60 fuel-cladding gap has been reduced by more than a factor of four.
The transient for this parametric was similar to the base case, i.e., fuel pin failure resulted in hydraulic sweepout of fuel and a benign accident ( termination. Figure ~Q001.451-2 gives the overall power and reactivity traces. l These may be compared to Fig.6-1 of Ref.15, PSAR Section 1.6 for the base case. l31l60 The hiah cap conductance case was more incoherent than the base case and the
~
final net reactivity was not as low since fewer pins failed and less cavity fuel was available for sweepout following each pin failure. In general, these results do strongly suggest that it is indeed conservative to use a low l steady-state gap conductance which increases as a TOP accident progresses ! (particularly, when in the SAS model trans'ent fuel cracking is modeled as l resulting in fuel-cladding contact). The specific points raised in the question can be addressed as follows: i
- Note that Appendix F has been withdrawn. The text, upon which the question l was based, can now be found in Section 5 of Reference 15, PSAR Section 1.6. 60
'O Amend. 60 Q001.451-1 Feb. 1981 l
(a) It is true that (1) early pin failures would be beneficial to the accident outcome, and (2) the first pin failure is at 12.65 sec in h the high gap conductance case versus 12.00 sec in the base case. However, these observations miss the important point regarding the molten fuel present at pin failure. Figure Q001.451-3 and Figure 6-5 of Reference 15, PSAR Section 1.6, compre the fuel conditions versus time for the two 61 cases. The base case has a centerline molten feel fraction of over 40% at fail ure. The high gap conductance has a centerline molten fuel fraction of less than 30%. These pins fail with less molten fuel because of the greater transient fission gas availability due to the colder steady-state temperatures. In a slow transient, a reduction in the amount of molten fuel that is available for rapid mixing with sodium will reduce the extent of any fuel-coolant heat transfer. The fuel-coolant interaction pressures are thus lower in the high gap conductance case. In short, the absolute time into the transient is not relevant if the steady-state conditions are different. (b) Figure Q001.451-4 and Figure 5-21 of Ref. 15, PSAR Section 1.6, compare ]60 the restructuring boundaries for the two cases. The differences are appreciable,.but the high gap conductance case still has a central void extending over 55 cm., and pin failure is predicted to be only one SAS node lower, 47 cm. This is not significant due to the reduced FCI. , Figure Q001.451-5, and.0001.451-6, and Fig. 6-2 o' Ref. 15, PSAR Section 131 1.6, compare the voiding reactivity. The high gap conductance parametric 60 reduces the peak voiding reactivity in Channel 8 f om s9 cents to 16.5 cents. (Channel 5 no longer even fails due to the increased incoherence). Figure Q001.451-7, and Figure 6-4 of Ref.15, PSAR Sectio.1 1.6, compare the 31 voiding profiles in Channel 8. The lower FCI zone interface was pushed l60 downward almost 10 cm more in the base case. In the context of the burst failure criterion, there is not a strong sendcivity of the pin failure location to the degree of restructuring. (c) Figure Q001.451-8 and Figure 5-22 of Ref.15, PSAR Section 1.6, compare thel60 steady-state fission gas retention. There is obviously more fission gas available with the lower steady-state fuel temperatures. As pointed out above, this results in pin failure with a smaller amount of molten fuel. i With more restructuring, more fuel must melt to get sufficient pressuri-l zation for pin failure. The base case had 45.4 gm of molten fuel in the pin cavity at failure; the high gap conductance parametric had 23.6 gm. In arslow TOP accident, less transient fission gas availa-bility is a conservative assumption. (d) As discussed in point (a), it is the time from which fuel first starts
- to melt that is important. In the base case, initial fuel melting l produced little pressurization in the peak pins; the lower power pins were allowed to get closer to their failure thresholds. In the high gap conductance case, initial fuel melting failed the peak pins ter-minating the transient. There will be more incoherence if the gap conductance is increased.
l Q001.451-2 Amend. 61 Sep 1981 h l l
1 -s Question 001.541 (F6.3.2.3) Discuss in a quantitative manner the effect of the shield plates, the vortex suppressor plate and crush tubes on the sodium slug impact forces transmitted to the head. Discuss how the CDA analysis presented accounts for the presence of these members.
Response
The CRBRP Project has consolidated all considerations given Hypothetical Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b, PSAR Section 1.6) and its associated references; consequently, PSAR Appendices D and F have been withdrawn in Amendments 24 and 60 respectively. The response to this question is now found in Section 5.4 of Reference 10a and 14, PSAR Section 1.6. 60 61 1 O i ( ! () Q001.541-1 Amend. 61 Sept. 1981 l .- . - - _ _ - - - ~ , . - , _ _ . , . _ _ . _ , _ _ . . , _ . _ . -
n
! )
Question 001.550 (F6.3.4.2) V Provide an assessment of the maximum stresses in the transition joints betweeq the Inconel 600, the Stainless Steel ress il and Vessel Flange. Discus; whether there is a potential problem with these welds being able to withstand the CDA near the end of the plant life due to material degradation effects in the weld or heat affected zones. Response:
- The service er.vironment and temperature of the transition region of the reactor vessel will not degrade its performance. PSAR Section 5.2.6 has been expanded to discuss the transition region.
While concerns have been expressed and failures noted in a recent survey (Ref. Q001.550-1) of dissimilar metal transition joints, they have been associated with substantially more severe service conditions than those associated with the joints in question. Thermal cycling between room temperature and 1000 to 1200 F has typically been utilized to evaluate the performance of dissimilar n.etd joints. This cycle is extremely severe in comparison to the temperature cycle between 400 and 465*F for less thanl60 1000 cycles to which the vessel transit:nn joint (ferritic to inconel 600) will be subjected. It has also been suggested by several studies that in the presence of an oxidizing environment a mechanical notch will form at the interface between weld metal (nickel alloy) and ferritig g) V base metal because of differences in oxidation resistance at high (<1000 F) temperatures. The vessel transition weld will not be subjected to l60 high temperatures (1000*F and up) nor to a highly oxidizing environment. The pressure vessel nozzle safe end transition joint utilized in PWR plants consists of an overlay of Inconel 182 (coated stick version of Inconel 82) on SA 508 class 2 nozzle forgings for subsequent joining to Type 304 stainless pipe. This joint is about 30" in diameter, 3" thick and operates at approximately 6000F. Performance experience with this joint has not indicated the occurence of any problems related'to material degradation in service. Such factors as microfissuring of the weld filler metal and the presence of oxide inclusions cannot be overlooked in performing these transition weld;. These concern; have been overcome through the introduction of imoroved weld filler metal compositions, recognition and control of base metal dilution effects on the filler metal, welding process qualifications with related inspections, and monitoring of welding and inspection techniques. The generally accepted method of makino transition welds between austenitic stainless and ferritic steels for service under substantially more severe conditions than those imposed on the transition joints in this vessel is to usc nic' el-base filler metals. Large numbers of such welds are in service in p .rochemical plants and fossil as well as conver.tional nuclear power plants with apparently satisfactory results. (~% b
- Note that Appendix F has been withdrawn. The text, upon which the question was based, can now be found in Section 5 of Reference 10a, PSAR Section 1.6. 60 Amend. 60 0001.550-1 Feb. 1981
O Based on these considerations, it is expected that there will be no appreciable material degradation over the plant life in the transition joint region. All preliminary analyses show that both stresses and strains in the transition area satisfy the SMBDB criteria (stress and strain) defined in reference 10a, PSAR Section 1.6. Therefore, the SMBDB l61 criteria will be satisfied in the transition joint region at the end of 60 plant life.
References:
Q001.550-1: J. F. King, " Behavior and Properties of Welded Transition Jo'nts Between Austenite Stainless and Ferrite Steels - A .iterature Review", ORNL-TM-5163, November 1975. O O Q001.550-2 Amend. 61 Sept. 1981 s
. . - - - -w- - . - .,,
Question 001.554 (F7.2.5) Describe and discuss the interaction of the shear rings, risers, riser bolting and support bearing during a CDA. Provide details of the arrangement of shear rings, shear ring keepers and their material properties. Discuss the response of the bearing and retention in its seat during a CDA. Indicate at what load the shear ring keepers can be displaced. Discuss the possibility and the effect of non-uniform loading in the shear rings and non-uniform deformation which may result in tilting of the plug and, the subsequent potential for the plug not to reseat and raal.
Response
The CRBRP Project has consolidated all considerations given Hypotheticai Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b, l.32 PSAR Section 1.6) and its associated references; consequently, PSAR Appendices D and F have been withdrawn in Amendments 24 and 60 respectively. The response to this question is now found in Reference 11, PSAR Section 1.6. 60 l61 ! /~} .- i Amend. 61 0001.554-1 Sept. 1981
O Question 001.555 (F1.5.2.9 Part II) Considering the differences between the FFT' and the CRBR designs, as well as the differences in the mechanical source term, describe how you plan to verify the response of the upper portion of the CRBRP reactor structure experimentally. Describe and discuss any planned experimental program and its schedule to address this matter. ,
Response
The CRBRP Project has consolidated all considerations given Hypothetical l Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b, PSAR Section 1.6) and its associated references; consequently, PSAR Appendices D and F have been withdrawn in Amendments 24 and 60 respectively. The response to this question is now found in Section 5.4.2.4 of Reference 10a and 14, PSAR Section 1.6. 60 61 . O 4 Amend. 61 Q001.555-1 Sept. 1981
..- ~ . . . _ _ _ .-. , _ _ .._ _ . _ _ _ . _ ._ _ _ ...__.. _ . . . _ . _ . . _ , . _ _ . . _ . _ _ . . _ . . - . _ . . .
Question 001.608 The experimental program at Stanford Research Institute, planned to verify the response of the upper portion of the CRBRP reactor structure, i.e., ro-tating plugs, reactor head, shear rings, hold down system, upper internal structure, etc., is outlined in Section 1.5.2.9 of the PSAR. Provide in the PSAR the details (sketches or drawings) of the designs to be tested, the loads, the matrix and sequence of tests, the parameters to be studied, the expected accomplishments, and the way the experimental output data will be used to justify the current head design, assuming a 661 MW-sec structural design basis load and a 1200 MW-sec maximum load.
Response
The CRBRP Project has consolic'ated all considerations given Hypothetical Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b, PSAR Section 1.6) and its associated references; consequently, PSAR Appendices D and F have been withdrawn in Amendments 24 and 60 respectively. The response to this question is now found in Reference 14, PSAR Section 1.6. 60 l 61 O 1 i Q001.608-1 Amend. 61 Sept. 1981
Question 001.610 Additional descriptive information on the SRI test program is required as indicated: (a) Provide drcwings of the test models to be used for all phases of testing. (b) Provide a technical discussion of the scaling laws and parameters being scaled, and indicate what criteria for acceptability are being used to judge success. -Include in this discussion not only the vessel head and its details but the in-vessel components and structures, and nozzle simulation. (c) Provide the detailed test plan for all planned tests including a 1,11 discussion of the source simulation. Discuss how it relates to the fuel vapor expansion and the NRC specified source term (1200 MJ). (d) Frovide detailed locations of test instrumentation, descriptions of
, the response characteristics and data recording. Describe how these locations and recorded data will allow an independent study to be made of the energy distribution.
(e) Describe the measurements to be taken at the vessel support to O' verify the loads transmitted to the supports. (f) For each test plan discuss all material selections, basis and dynamic properties. (g) Indicate if high speed photography will be taken.of the experiments and whether the displaced water will be measured,
Response
The CRBRP Project has consolidated all considerations given Hypothetical i Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b, PSAR Section 1.6) and its associated references; consequently, PEAR Appendices D and F have been withdrawn in Amendments 24 and 60 respectively. The response to this question is now found in Section 5.4.2.4 of Reference 10a and 14, PSAR Section 1.6. 61 i l lO 60 Amend. 61 Q001.610-1 Sept. 1981 t
O Question 001.612
' Clarify the purpose of the phase II SRI tests (Amendment 34, page 1.5-461, SM-1), and elaborate on what these tests will demonstrate in the absence of a dynamic loading.
Response
The CRBRP Project has consolidated all considerations given Hypothetical Core Disruptive Accidents into report CRBRP-3 (References 10a and 10b, r PSAR Section 1.6) and its associated references; consequently, PSAR Appendices D and F have been withdrawn in knendments 24 and 60 respectively. The response to this question is now found in Section 5.4.2.4 of Reference 10a and 14, PSAR Section 1.6. 60 61 f i ( . I i l Amend. 61 q001.612-1 Sept. 1981 l
Question 001.690 h) 's Your response to Item 001.502 is not entirely acceptable. number of concerns with the response which you have provided. We have a Reassess your response to Item 001.502 in light of the concerns delineated below; your reassessment should include an evaluation for each of items provided below: (1) The statement that the particles entrained in fluid above the Upper Internal Structure (UIS) exoerience little radial movement is mis-leading. On the contrary, according to calculations performed with the MIX code (
Reference:
Fiqure E3, ANL-CT-75-41, " MIX-A Computer Code for Transient Thermal-Hydraulic Analysis of LMFBR Outlet Plenums," May 1975) after 35 sec of flow coastdown, the flow is essentially a stratified recirculation flow, with fluid immediately above the UIS moving horizontally. Therefore, the debris particles initially situated above the UIS would be carried away by the recirculation flow, resulting in little settlement of the debris on ton of the UIS. As a result, the assumption that all debris particles directly above the UIS settle on its surface may lead to serious error and the settled fraction could be well below the 27% claimed in the response. (2) As is evident from Figure 23 of the aforementioned reference, only the fluid near the outlet nozzle moves essentially radially. In the recirculation flow region, both the vertical and horizontal flow velocities are equally imoortant. Therefore, the assumption that the sodium in the settling zones moves only in the radial direction may lead to substantial error in calculating the fraction settled in the (V] upoer sodium olenum. (3) The assumption that the fraction settled in the reactor vessel is the ratio of the vertical distance traveled to the initial height of the column of particles is difficult to conceive physically. This assump-tion needs farther elaboration and justification. Since this assump-tion plays an important role in the whole theory, provide your plans for exoerimental verification. (4) Equation (2) in your response to Item 001.502 appears to be incorrect because the actual particle settling velocity (relative to the vessel wall) is not the terminal velocity. Instead, it should be equal to the vector sum of the teminal velocity and the fluid verticle velocity.. If this equation is only an approximation, this is not a good approx-imation since in most of the flow regions the verticle f ow velocity is equally important. . (5) The significance of the upper plate area on the UIS is not understood, as well as why this area establishes the fraction of material to be swept toward the upper plenum outlets. (6) In your analyses of settling rate and particle concentrations per linear foot of piping, you have not accounted for oarticle interactions. Provide us with your assessment of this effect. O Q001.690-1 Amend. 60 Feb. 1981 J
~ - _ _ _ _ _ _ _ _ _ _ _
(7) The equation at the bottom of page Q001.502-4 is not valid for the verticle piping segments. By streaming through the verticle pipe, the debris may disperse again, resulting in redistribution of the debris particles in the verticle pipe and at its outlet. Resnonse: The assessment of in-vessel debris behavior has been updated and is provided in CRBRP-3, Volume 2 (Reference 10b of PSAR Section 1.6), Section 3.1. l61 Specific comments contained in the question are discussed below. (1) The analysis of the settling of fuel particles in the reactor vessel plenum and their transport to the PHTS piping is currently based on two dimensional sodium flow that is predicted with the VARR II computer code .(see Appendix A of the PSAR) and the particle mechanics discussed in Section 3.1.1.1 of CRBRP-3, Volume 2. With this model only a small fraction of the fuel particles would oe predicted to settle on the top of the UlS. tam e 3-7 i of CRBRP-3, Volume 2, contains the predicted fuel fraction remaining in the reactor vessel and the fraction exiting through the outlet piping. (2) As indicated in (1) above, a new analytical model has been deve'oped which removes the approximation previously followed that only radial flow occurs in the settling zone. The particles move due to the forces of gravity and drag forces from the two dimension fluid flow streamlines. (3) This assumption is not used in the current model. Approximations were made in the previous model to conserva tively predict the amount of fuel material 9 transported to the PHTS piping. The revised model has reduced this amount to 18-25 percent (see Table 3.7 of CRBRP-3, Volume 2) of the core fuel as comoared to the previous prediction of 35% reported in the response to Question 001.502. (4) In the revised model, the fluid drag forces on the particles are based on the relative velocity of the particle and the flowing sodium. The particle terminal velocity relative to the reactor vessel wall is properly calculated. (5) The geometry of the UIS in the present model is shown on Figure 3-1 of CRBRP-3, Volume 2. The total fuel debris is introduced into the flowing stream in the region between the top of the core and the bottom of the UIS. Particle deposition is calculated on the basis of the forces exerted upon them. It is no longer assumed that deposition on the top of the UIS is proportional to the solid area of the top of the UIS. (6) Particle interaction was consider ed in Section 3.1.1.1 of CRBRP-3, Volume 2. It was concluded that because of the low particle concentration (0.7 v/o), settling would occur without hinderance. Since agglomeration of particles would enhance in-vessel settling and reduce fuel carried into the piping, it is conservative to exclude this potential interaction. O Amend. 61 Q001.690-2 Sept. 1981 l +. _ . _ _ _
Comparison of CRBRP Containment Cleanup System to Regulatory Guide 1.52, Rev. 2 Table Q011.25-1 (Sheet 1) Regulatory Applicable to CRBRP Containment 4 Position Cleanup System Remarks Yes No 1.a X 1.b X System does not contain absorbers. l 61 See CRBRP-3 Vol. 2 Report 1.c X System does not contain adsorber 1.d X CRBRP has no Containment Spray System 1.e X 2.a X Redundancy is provided for active components only (,) 2. b X Redundant components will be physically separated 2.c X 2.d X System will be designed to withstand maximum expected pressure 2.e X
~
2.f X System does not contain HEPA i Filters 2.g X l 2.h X System manually activated, see l Sect. 2.3 of CRBRP-3 Volume 2 l 2.1 X 2.j X
- 2. k X No outside air intakes are l required 2.1 X 3.a X System does not contain a demister 3.b X System does not contain a demister O- 3.c X System does not contain a demister Amend. 61 Q011.25-3 Sept.-1981
Revised Response Table Q011.25-1 (Sheet 2) Regulatory Applicable to CRBRP Containment Position C1eanup System Remarks Yes No 3.d X System does not contain HEPA Filters 3.e X System does not contain adsorber 3.f X System does not contain adsorber or HEPA filters 3.g X System is all welded leaktight 3.h X System does not contain water drains 3.1 X System does not contain adsorber 3.j X System does not contain adsorber 3.k X System does not contain adsorber 3.1 X 3.m X 3.n X 3.o X 3.p A 4.a X 4.b X System does not have filter banks 4.c X System does not have adsorber and HEPA filters 4.d X System does not have heaters 4.e X System does not have adsorber or HEPA filters 5.a X No 00P or activated carbon test are required 5.b X System does not have HEPA filter or iodine adsorber 5.c X System does not have HEPA filters 5.d X System does not have activated carbon adsorber g Amend. 60 QO11.25 4 Feb. 1981
O Question 421.1 (13.7) U Section 13.7 of the PSAR is deficient with respect to addressing the re-quirements of 10 CFR Part 73. Provide the following: (1) A statement indicating that the physical protection of the Clinch River Breeder Reactor Plant will raet the requirements of Sections 73.40, 73.50, 73.60, and 73.70. (2) Revise or delete the statements in Section 13.7 of the PSAR which are not in accordance with the requirements of 10 CFR Part 73 (e.g., search requirements). (3) Address the Subsections of 73.50 and 73.60 which' relate to design considerations, and potentially long lead time items requiring pre-liminary recognition at the PSAR stage of the licensing process. The following subsections contain subject matter in the above cate-gories: 73.50(a)(4);(b)(1),(2),(3),(4),(5);(c)(1),(4),(6); (d)(1), (2); (e)(1), (2), (3), (4); (g)(1); 73.60(a)(1), (2) (3), (7); (b); (c). Your response should also discuss your degree of conformance to the applicable Regulatory Guides (e.g., 5.7, 5.12, 5.14,5.20,5.43,5.44) which provide criteria acceptable to the Staff for meeting the requirements of various subsections identi-fied above. Any deviation from the Guides should be fully ex-plained, including the basis for substantiating the equivalency of n your alternate position to that set forth in the Guide. Informa-V tion submitted in response to item 421.2 need not be repeated here. Your response should be withheld from public disclosure pursuant to 10 CFR 2.790 if it contains information which could compromise your safeguards program.
Response
Part (1): Requirements of 10 CFR 73.40, 73.50, 73.60, and 73.70 shall be met in the generic sense at the Clinch River Breeder Reactor Plant in that some subsections of the regulations are not appli-cable, i.e., handling of uranium scrap, limiting activities 4 within material access areas to those associated only with the use, storage and handling of SNM. Part (2): See revised Section 13.7. l61 Part (3): Applicable sections of Regulation 73.60 shall be met for the Clinch River Breeder Reactor Plant when SNM is onsite to provide assurance that the license will not present undue risks to the health and-safety of the public. Regulation 73.50 is not strictly applicable to the CRBRP, however, the. Project will . comply with spirit and intent of 11 those requirements as appropriate. For further information on how the spirit and intent of Regulation 73.50 will be met, see the. proprietary response to Question 421.3. The regulatory positions ,35 Amend. 61 Q421.1-1 Sept. 1981
as stated in Regulatory Guides 5.7, 5.12, 5.14, 5.20, 5.43, and 5.44 will be congruously applied to the procedural mea-sures detailed for the physical protection of the Clinch River Breeder Reactor Plant. (In several of the:;e guides, regulatory positions do not apply to the physical protection and safeguards being established for the protection of the CRBRP nuclear fuel against acts of sabotage and theft in a power plant, i.e., RegulatoryGuide5.1.4C.2.A(4),(5).) Implementation of the referenced Regulatory Guides will be detailed in the Industrial Security Section of the FSAR. The subsections referenced are addressed as follows: 73.50(a)(4) - See new Subsections 13.7.1.1, 13.7.1.3, 13.7.2.1 and 13.7.3.5: 73.50(b)(1)(2),(3),(4),(5) - See new Subsections 13.7.2.la),b),e),f), 13.7.2.5 and 13.7.2.7 73.50(c)(1),(4),(6) - 5ee new Subsections 13.7.2.5, 13.7.3.1 and 13.7.3.2. 73.50(d)(1),(2) ,See new Subsections 13.7.2.11), 13.7.2.5 and 13.7.2.6 73.50(e)(1),(2),(3),(4) - See new Subsections 13.7.2.1k),1). 73.50(g)(1) - See new Subsection 13.7.3.5. 61 73.60(a)(1),(2),(3),(7) - See new subsections 13.7.2.5 and 13.7.3.4. Reactor Fuel will be received at the Clinch River Breeder Reactor Plant in sealed con-tainers and off loaded within the RSB. The re-fueling operation for the reactor with the ex-
~
vessel transfer machine. The main control center for the refueling process is located at the south end within the RSB. All fuel assemblies are handled remotely; there are no hands on operations from shipping cask to reactor core. (See Section 9.1). 73.60(b)- Exiting personnel need not be searched since there is no access to SNii. 61 73.60(c)- See new Subsection 13.7.2.5. Amend. 61 Q421.1-2 Sept. 1981
Question 421.2 (13.7)
/ 'In order to evaluate the acceptability of the facility design with respect \ to protection against industrial sabotage and theft of special nuclear material, the folicwing additional infonnation is required to determine conformance with the design requirements of 10 CFR 73, and the design criteria set forth in Regulatory Guide 1.17 and ANSI N18.17-1973. Your response should be withheld from public disclosure pursuant to 10 CFR 2.790.
(1) Provide figures and/or drawings which identify the following: (a) Owner-controlled area, including private property markers, park-ing lot (s), and roads to be used for surveillance. (b) Protected areas, including the associated isolation zones, clear area, physical barriers, access control points, lighting, intrusion monitoring and/or perimeter alarm systems, and roads or pathways to be used for surveillance. (c) Vital equipment and vital areas, including all access points. (d) Material access areas, including all access points. (e) Alarm station locations. (2) Provide sufficient infor.ution which shows that the physical barrier p construction for the protected areas, vital areas, and material g access areas satisfies the requirements of 10 CFR 73.2. (3) Describe the des.ign features to be used for protecting all potential access points into the vital areas and material access areas against unauthroized intrusion. Such features should include locking de-vices and intrusion detection devices. (4) Indicate that all intrusion alarms, emergency exit alarms, alarm systems, and lina supervisory systems will meet the level of per-formance and reliability specified by the Interim Federal Specifica-tions W-A-00450B (GSA-FSS), dated February 6,1973. , (5) Describe the physical security protection to be utilized in the design for the protection of security system service panels and wiring for protectiv- devices, security communications systems, and door lock actuators. ,
Response
Part (1): (a) See Section 2.1. (b) See proprietary Figures 13.7-2 through 13.7-5 of revised Section 13.7. (c) See proprietary Figures 13.7-6 through 13.7-11 of revised Section 13.7.
) (d) See proprietary Figures 13.7-6 through 13.7-11 of revised Section 13.7.
(e) See proprietary Figures 13.7-6 and 13.7-11 of revised Section 13.7. 61 1 421.2-1 Amend. 61 i . . .. . - - _ -
]
Part (2): See new Subsection 13.7.2.7 for protected area physical barrier. All buildings housing vital areas and material access areas are located in the protected area and consequently encircisd by the protected area security barrier. Each of these buildings has a limited number of access doors controlled by a key-card. All emergency exits shall be continuously alarmed. All ex-terior doors and interior doors to vital areas and materials access areas shall be designed to resist forced entry. The ex-terior doors shall have missile protection. Also, each of these buildings is seismically designed and constructed of re-inforced concrete. However, the reactor containment building above elevation 816'-0" is the exception, having a steel con-tainment shell of approximately li"s thickness. None of these buildings have windows. Other exterior openings shall be secured by grates or covers of sufficient strength such that the integrity of the wall is not lessened. Part(3): See subsections 13.7.2 and 13.7.2.1. Part (4): See subsection 13.7.2.6. Part (5): See subsections 13.7.2.4, 13.7.2.5 and 13.7.2.6. 61 O f l l i h Q421.2-2 Amend. 61 Sep t. 1931 L
~
)s
-( ) Question 421.6 (13.7) 4 Section 13.7.3.1 of the PSAR states that all outside access points to' plant
' buildings will be locked, and alarmed or card-key controlled. Resolve the inconsistency between this statement and Figure 13.7-3 which shows five access points on the periphery of the plant service building which are neither alarmed nor card-key controlled.
Response
i f Proprietary Figure 13.7-6 has been revised to resolve the inconsistency noted in the question. ! Since Figure 13.7-6 is proprietary, it will be forwarded to NRC separately. 61 i 1. O i I l ) i
- O Q421.6-1 Amend. 61 Sept. 1981
---w -v - + - w ,w_ h , v. E r ,,,..w,,,,--wmew,mw-wwy, ,,-e,+cpy,-,,www4-.--,,--~~4.,-www-pw.-,w.,,-=,_w,y-eyw.,-ww.
Question 421.10 (13.7) Identify the location of the two alarm stations mentioned in Section 13.7.2.6 of the PSAR. We are unable to locate these stations on Figure l 13.7-4 as stated.
Response
The CAS and SAS are identified on Figures 13.7-6 and 13.7-11 respectively which are proprietary. 61 O Q421.10-1 O- Amend. 61 Sept. 1981
. . - .. -}}