ML19224C348
| ML19224C348 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 06/29/1979 |
| From: | Copeland R ENERGY, DEPT. OF |
| To: | |
| Shared Package | |
| ML19224C347 | List: |
| References | |
| NUDOCS 7907020195 | |
| Download: ML19224C348 (230) | |
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0834 Department of Energy Clinch River E$reeder Reactor Plant Project Office PO. Box U Oak Ridge, Tennessee 37830 DocKct No. 50-537 File: 05.10 lune 29,1979 Mr. Domenic B. Dassallo Acting Director Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Dassallo:
AMENDMENT NO. 50 TO THE PRELIMINARY SAFETY ANALYSIS REPORT FOR CLINCH RIVER BREEDER REACTOR PLANT The application for a Construction Permit and Class 104(b) Operating 9 License for the Clinch River Breeder Rcactor Plant, docketed April 10, 1975, in NRC Docket No. 50-537, is hereby amended by the submission of Amendment No. 50 to the Preliminary Safety Analysis Report pursuant to 50.34(a) of 10 CFR Part 50. This Amendment No. 50 includes: an update to Section 9.9.2, " Emergency Plant Service Water System"; an update to Chapter 13, " Conduct of Operations"; and other updates and revisions, as well as a response to NRC's request for additional information contained in a letter dated August 17, 1976.
A Certificate of Service, confirming ser vice of Amendment No. 50 to the PSAR upon designated local public officials and representatives of the EPA, will be filed with your office after service has been mde. inree signed originals of this letter and 97 copies of this amendment, each with a copy of the submittal letter, are hereby submitted.
Since ly, a~
,ond Cop ar 4
l PS:79:161 A ing Assis Directo for Public afety Enclosure '){g ]g cc: Service List SUBSCRIBgandE SWORN to before me Standard Distribution day of June, 1979.
9 Licensing Distribution this if W&&
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!, k 0Vlf/-L Not yfPQblic Ccmmission Emp;res May 21,1980 7
13
STAfiDARD DISTRIBUTI0il Mr. R. Balent (2) Mr. W. W. Dewald, Project Manager (2)
Vice President and General Manager CRBRP Reactor Plant Atomics International Division Westinghouse Electric Corporation Rockwell International Advanced Reactors Division P. O. Box 309 P. O. Box 158 Canoga Park, CA 91304 Madison, PA 15663 Mr. Michael C. Ascher (2)
Project Manager, CRBRP Burns and Roe, Inc. Mr. C. R. Adams (1)
Resident Manager, CRBRP 700 Kinderkamack Road Burns and Roe, Inc.
Oradell, flJ 07649 P. O. Box T f!r. Lochlin W. Caffey (2)
Director fir. George G. Glenn, Manager (2)
Clinch River Breeder Reactor Plant Clinch River Project P. O. Box U General Electric Company Oak Ridge, Til 3783) P. O. Box 5020 Sunnyvale, CA 94086 Mr. Dean Armstrong (2)
Acting Project Manager, CRBRP Mr. Hardy B. Adams, Jr. (2)
Stone & Webster Engineering Corp. Projects Manager, LMFBR Programs 9 P. O. Box 811 Oak Ridge, Til 37830 Tennessee Valley Authority 1300 Commerce Union Bank Building Mr. Don E. Erb (1)
Acting Assistant Director for Reactor Projects Division of Reactor Research and Technology U. S. Department of Energy Washington, DC 20545 Mr. Harold H. Hoffman (1)
Site Representative U. S. Department of Energy Westinghouse Electric Corporation Advanced Reactors Division P. O. Box 158 Madison, PA 15663 Mr. J. E. fiolan (2)
Project ifanager, CRBRP s, Westinghouse Electric Corporation /68 iUu,,
Advanced Reactors Division P. O. Box W Oak Ridge, Til 37830 4/19/79
LICEriSIllG DISTRIBUTI0ft fir. Itugh Parris fianager of Power Tennessee Valley Authority 830 Power Building Chattanooga, Til 37401 Mr. R. M. Li ttle Electric Power Research Institute 3412 ilillview Avenue Palo Al to, CA 94303 Dr. Jeffrey 11. Beoido, Manager Analysis and Safety Department Gas Cooled Fast Reactor Program P. O. Box 81608 San Diego, CA 92138 9
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, l \J )
SERVICE LIST Atomic Safety & Licensing Board Anthony Z. Roisman, Esq.
U. S. thclear Regulatory Commisgion flatural Resources Defense Council Washington, D. C. 20555 91715th Street, IN Washington, DC 20036 Atomic Safety & Licensing Board Panel U. S. f'uclear Regulatory Commission Dr. Cadet H. Hand, Jr. , Director Washington, D. C. 20555 Bodega Marine Laboratory University of California S. Wallace Brewer, Judge P. O. Box 247 Office of County Judge Bodega Bay, CA 94923 Roane County Court House Kingston, Yft 37763 Lewis E. Hallace, Esq.
Division of Law Dr. Thomas Cochran Tennessee Valley Authority flatural Resources Defense Council, Inc. Knoxville, T!i 37902 917 15th Street,fiW 8th Floor Washington, DC 20005 Docketing & Service Station Office of the Secretary U. S. fluclear Regulatory Commi.ssion
@ Washington, DC 20555 Counsel for flRC Staf f U. S. fiuclear Regulatory Commission
!!ashington, DC 20555 William B. Hubbard, Esq.
Assistant Attorney General State of Tennessee Office of the Attorney General 422 Supreme Court Building f!ashville, Tf1 37219 Mr. Gustave A. Linenberger Atomic Safety & Licensing Board U. S. f uclear Regulatory Commission Washington, DC 20555 Marshall E. Miller, Esq, Chairman Atomic Safety & Licensing Board U. S. fluclear Regulatory Commission Washington, DC 20555 Luther M. Reed, Esq.
Attorney for the City of Oak Ridge 253 Main Street, East ')
" 6 p' ]'G" Oak Ridge, Tri 37830
Amendment 50 Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report (Docket No. 50-537)
This fiftieth amendment to the Clinch River Breeder Reactor Plant Preliminary Saiety Analysis Report includes updates to sections describing the Emergency Plant Service Water System and Conduct of Oper-ations, other updates and revisions as well as responses to .*4RC's request for additional information. Vertical margin lines on the left hand side of the page are used to identify new design information while lines on the right hand side of the page identify question / response information.
A page replacement guide appears following the list of responses to NRC questions.
NRC Ques. No.
011.23 200 }!
PAGE REPLACEMENT GUIDE FOR AMENDMENT 50 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT (DOCKET NO. 50-537)
Transmitted herein is Amendment 50 to the Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket No. 50-537.
Amendment 50 cot.sists of new and replacement pages for the PSAR text and question / response supplement pages.
The following attached sheets list Amendment 50 pages and in-structions for their incorporation into the Preliminary Safety Analysis Report.
e na i .a-
Amer.dment 50 Page Replacement Guide RemcVe These Pages In_ sert These Pages Chapter 3 3.8-1, la 3.P-1, la 3.8-3, 3a 3.8-3, 3a 3.8-42 thru 3.8-47a 3.8-42 thru 3.8-47a Chapter 4 4-iib, iic 4-iib, iiba, iic 4-xv, xva 4-xv, xva 4-xvii, xviia, xviii 4-xvii, xviia, xviii 4-xxi, xxia, xxii 4-xxi, xxia, xxii 4.2-95, 95a 4.2-95, 95a 4.2-ll5d, llSe 4.2-ll5d, ll5e 4.2-3280 4.2-328d Chapter 5
( 5.4-12 5.4-12 5.6-6a, 7 5.6-6a, /
5.6-20, 21 5.6-20, ?i 5.6-34, 34a 5.6-34, 34a Chapter 6 6.2-27d, 27e 6.2-27d, 27e 6.3-lb, 2, 2a, 2b 6.3-2, 2a Chapter 7 7-ix, x 7-ix, x 7.1-17, 18 7.1-17, 18 1.2-77, 78 7.2-77, 78 7.5-7, 8 7.5-7, 8 7.5-15, 16 7.5-15, 16 7.5-26, 27, 27a, 28, 29, 30 7.5-26 thru 7.5-30 7.5-33b, 33c 7.5-33b, 33c, 33d 7.5-34, 35 7.5-34, 35 7.5-42 7.5-42 7.6-3, 3a, 3b, 3c 7.6-3, 3a, 3b, 3c A 1
2bf3 I IJ
Remove These Pages Insert These Pages Chapter 9 9.1-13, 14 9.1-13, 14 9.2-3, 3a 9.2-3, 3a 9.2-7, 8, 8a, 8b, 9 9.2-7, 7c, 8, 9, 9a 9.3-12, 13 9.3-12, 13 9.5-3, 4 9.5-3, 4 9.5-7, 8 9.5-7, 8 9.6-65, 66 9.6-65, 66 9.6-84 9.6-84 9.6-87 9.6-87 9.6-92 9.6-92 9.8-1 thru 9.8-6 9.8-1 thru 9.8-6 9.8-7, 7a, 8, 9, 10, 11 9. 8-7, 7a , 8, 9, 10, 11 9.8-14, 15 9.8-14, 15 9.9-1 thru 9.9-16 0 91 thru 9.9-16 9.9-18, 19 9.9-18, 19 9.13-11, 12 9.13-11, 12
- 9.13-46a Chapter 11 ll-i thru ll-x ll-i thru ll-x 11.1-24, 25 11.1-24, 25 O.
1- 11.2-1, 2 11.2-12, 13 11.2-1, 2 11.2-12, 13
- 11. 2- 31 11.2-31
- 11. 3-1 thru 11.3-6 11.3-1 thru 11.3-6 11.3-9 thru 11.3-12 11.3-9, 10, ll, lla, 12 11.3-15, 16 11.3-15, 16 ll . 3-18a ,18b ,18c ,19, 20, 21 ll.3-18a, 18b, 18c, 19, 20, 21 11.3-24 thru 11.3-33 11.3-24 thru 11.3-33 11.3-36 thru 11.3-39 11.3-36 thru 11.3-39 ll.3-39c, 40, 41, 42 ll.3-39c, 40, 41, 42 11.3-49, 50, 51, 52 11.3-49 11.4-3, 4 11.3-3, 3a, 4 11.4-12, 13 11.4-12, 13 ll.5-6b, 7 ll.5-6b, 7 Chapter 12 12-i thru 12-ix 12-i, ii, iia, iii, thru ix 12.1-37 12.1-37 12.1-65a, 65b, 65c 12.1-65a, 65b, 65c 12.1-78, 79 12.1-78, 79 12.2-16, 17, 18 -
12.3-11 thru 12.3-16 12.3-11 thru 12.3-16 268 ii/;
B
R_emove lhese Pages Insert These Pages Chapter 13 13-i thru 13-vii 13-i, ii, iii, iiia, iv thru vii 13.1-1 thru 13.1-22 13.1-1 thru 13.1-21 13.2-1 thru 13.2-7 13.2-1 thru 13.2-7 13.3-1 thru 13.3-8 13.3-1 thru 13.3-21 13.5-3, 4, 5 13.5-3, 4, 5 13.7-1 thru 5, Sa, 6 thru 10 13. 7-1 thru 10,10a ,10b Chapter 15 15.5-20, 21 15.5-20, 21
- 15. 7-11, 12, 13, 13a 15.7-11, 12, 13 Chapter 16 16.2-2 16.2-2 268 1i5 C
O Amendment 50 Question / Response Supplement The Question / Response Supplement contains an Amendment 50 tab to be inserted following page Q-i (Amendment 49, April 1979). Page Q-i (Amendment 50, June 1979) is to follow the Amendment 50 tab.
The Questions / Response Supplement page listed below should be inserted in the proper numerical order following the correct section tabs.
The parentheses beside the question indicates the number of pages contained in the Question / Response.
NRC Ques. No.
011.23 (1)
Remove These Pages Insert These Pages Q001.301-1 Q001. 301 -1 Q001.581-8 thru 11 Q001. 581 -8, 9, 10, 10a , 11 Q001. 581 -14, 15 0001.581-14, 14a, 15 Q222.77-1 Q222.77-1 Q310.3-1, 2 Q310.3-1, 2 Q331.20-1 Q331.20-1 Q422.1-1 thru 8 Q422.1-1, 2 Q422.2-1 thru 8 Q422.2-1, 2 Q422.3-1 Q422.3-1 2006 D
3.8 DESIGN OF CATEGORY I STRUCTURES 3.8.1 Concrete Containment (Not Applicable) 3.8.2 Steel Containment System 3.8.2.1 Description 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 case to the spring line of the ellipsoidal-45 spherical dome. The cylindrical shell is embedded in concrete up to the 47 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 733 feet cp tc the 45l elevation of the underside of the operating floor, there is another rein-forced concrete wall of sufficient thickness designed to prevent bucklinc 45lof the steel shell. Neither of the two concrete walls are considered part of the containment vessel. Alumina-silica insulation is attached 33 to the inside surface of the Containment Vessel from elevation 816 feet to elevation 823 feet. The insulation is 3 inches thick and has a value of 0.0267 Btu /hr - ft OF. Its purpose is to limit the shell temperagure 48 at elevation 816 feet during Design Basis Accidents to less than 130 F. The vessel includes: its shell, a (" bottom liner plate, one 45 access airlock, one emergency egress airlock, vacuum relief system, one equipment hatch, penetrations, inspection ladders, miscellaneous appur-tenances and attachments. The configuration of the Containment Building is shown in figures in Section 1.z. ~ ' design lifetime of the contai-39 ment vesc^1 shall be 30 years. 3.8.2.2 Applicable Codes, Standards and Specifications 3.8.2.2.1 Codes The Containment Vessel will be designed, material procured, fabricated, installed and tested in accordance with the requirements of the ASME B&PV Code, Section III, Division 1, 1974 Edition with Addenda through Winter 1974 and Code cases 1713,1714,1809,1682 and 1785 and 50 43 ASME-III, 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". 2, 0 0 l ,l _,, 3.8-1 Amend. 50 June 1979
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 an d 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, Fabrication and Erection of Structural Steel Buildings (AISC, February 12,1969). 3.8.2.2.2 Design Specification Sunnary 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. 3.8-la Amend. 48 Feb. 1979
Tolerances The Containment Vessel as constructed shall not exceed the tol-erance requirements of NE-4000 of ASME-III for fabrication or erection. The dimensional control procedures shall meet the require-roents of RDT STD F3-15T. The out-of-plumb tolerances shall not exceed 1/500. The out-of-roundness tolerance shall not exceed h of one percent of the nominal inside diameter. 3.8.2.2.3 Applicable NRC Regulations and Regulatory Guides 15 NRC Regulatory 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). 1.11- Instrument Lines Penetrating Primary Reactor Cor,tainment 45 (March 10,1971) 1.12: Instrumentation for Earthquakes (Revision 1, April,1974) 1.15: Testing of Reinforcing Bars for Category 1, Concrete Struc-tures (Revisionl, December 28,1972) 1.19: Nondestructive Examination of Primary Containment Liner Welds (Revision 1, August 11,1972) Et 1.29: Seismic Design Classification (Revision 2, August 197 6) 1.55: Concrete Placement in Category 1, Structures (June 1973) 45 1.57: Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components ( June, 1973) 1.60: Design Response Spectra for Seismic Design of Nuclear Power Plants (Revision I . December,1973) 1.61: Damping Values for Seismic Design of Nuclear Power Plants 45 (Oct.1973) 1.63: Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plant: (Oct. 1973) SC 1.85. Materials Code Case Acceptability - ASME Section Ill m , , Division 1, 1976 /D0 l ,l b, 45 1.92: Combining Modal Responses and Spatial Components in Seismic Response Analysis (Revision 1, Feb., 1976) Amend. 50 June 197f 3.8-3
O Of the above, Regulatory Guide 1.63 is applicable after the following changes:
- 1. Deleting " Water-cooled" wherever it appears.
- 2. Replacing "Aprendix 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".
S. Substituting "(Sunme.1972 Addenda)" following". ..ASME Boiler and Pressure Vessel Code" with ",1974 Edition". Construction No special construction techniques are anticipated for this containment vessel. O Amend. 45 July 1978 3.8-3a
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NO. OF PAGES 7 O DUPL!CATE: ALREADY ENTERED INTO SYSTEM UNDER ANO.
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O'lLLEGIBLE: HARD COPY AT: O PDR O CF 0 OTHER 268 119
TABLE OF C0flTENTS Continued Page No. 34\ 4. 2. 2.1. 2. 9 Removable Radial Shielding (RRS) 4.2-95a 4.2.2.1.3 Design Loading 4.2-97 4.2.2.1.3.1. Core Support Structure 4.2-98 44f34 4.2.2.1.3.2 opper Internals Structure (UIS) 4.2-101 4 . . 2 . 2 Design Description 4.2-105 4.2.2.2.2.1 Reactor Internals Structures 4.2-105 4.2.2.2.2.1.1 Core Support Structure 4. 2-10S a 44l 4.2.2.2.1.2 Inlet Module 4.2-107a 4.2.2.2.1.3 Bypass Flow Module 4.2-108 44 ! 4.2.2.2.2.1.4 Fixed Radial Shield 4. 2-108 a 4.2.2.2.1.5 Fuel Transfer and Storage Positions 4.2-109 4.2.2.2.1.6 Horizontal Baffle 4.2-109 4.2.2.2.1.7 Upper Internal Structure 4.2-111 44l 4.2.2.2.1.8 Core Restraint System 4.2-113 44l 4.2.2.2.1.9 Removable Radial Shield 4.2-114 4.2.2.2.1.10 Maintainability 4.2-114 34< 4.2.2.2.1.11 Surveillance and In-Service Inspection Surveillance 4.2-114 4.2.2.3 Design Criteria 4.2-114 4.2.2.3.1 Lower Internals Structures (LIS) 4.2-115 4.2.2.3.1.1 Core Support Structure 4.2-115 4.2.2.3.1.2 Lower Inlet Module (LIM),B pass Flow Module (BPFM), 4.2-115 and Core Former Structure CFS) 4.2.2.3.1.3 Horizontal Baffle (HB), Fuel Trans'ar & Storage 4.2-115 Assembly (FT&SA), and Fixed Radial Shield (FRS) 50 4.2 -l l 5a 4.2.2.3.2 Upper Internals Structure (UIS) 268 120 Amend. 50 4-iib June 1979
TABLE OF CONTENTS Continued Page No. 4.2.2.3.2.1 Class 1 Appurtenances 4.2-ll5a 4.2.2.3.2.2 Internal Structure 4.2 -ll 5a 4.2.2.3.3 Additional Material Properties 4.2-llSe 4.2.2.3.3.1 Inconel 718 Fatigue Properties 4.2-ll5e 4.2.2.3.3.2 Environmental Effects on Material Properties 4.2-ll5f 50 4.2.2.3.3.2.1 Sodium Effects 4.2-ll5f O
} h'b \
Amend. 50 June 1979 4-iiba
TABLE OF C0flTENTS Continued Page No. 4.2.2.3.3.2.2 Irradiation Effects 4.2-Il7a 4.2.2.3.3.3 Friction, Ucar, and Self Welding Considerations 4.2-Il7a 4.2-ll7c 4.2.2.4 Evaluation 4.2.2.4.1 Lower Internals Structure Summary 4.2-ll7c 4.2.2.4.1.1 Analysis of Core Support Structure (CSS) 4.2-118 4.2.2.4.1.2 Analysis of Lower Inlet Module (LIM) 4.2-130 4.2.2.4.1.3 Analysis of Bypass Flow Modules 4.2-131 4.2.2.4.1.4 Analysis of the Fixed Radial Shielding 4.2-132 4.2.2.4.1.5 Analysis of the Fuel Transfer and Storage Assembly 4.2-132 4.2.2.4.1.6 Analysis of the Horizontal Baffle 4.2-132 4.2.2.4.1.7 Upper Internal Structure 4.2-132 4.2.2.4.1.7.1 Summary of Results 4.2-132 4.2.2.4.1.7.2 Material Properties 4.2-133 4.2.2.4.1.8 Core Restraint System Analysis 4.2-146 4.2.2.4.1.8.1 Summary of Resul ts 4.2-146 4.2.7.4.1.8.2 Material Properties 4.2-146 4.2.2.4.1.8.3 Analysis of Assembly Bowing and Duct Dilation 4.2-146a 4i. l 34 4.2.2.4.1.9 Remo.able Radial Shielding (RRS) 4.2-150 4.2.2.5 Welding and Seizing of Reactor Internal Parts 4 .2 -1 51 4.2.3 Reactivity Control Systems 4.4-151 4.2.3.1 Design Casis J.4-152
. . n ra ety Design CrH_eria L 2-152 34 4.2.3.1.2 Control Rod System Clearances 4.2-153 4.2.3.1.3 Mechanical Insertion Requirements d.2-153 4.2.3.1.4 Positioning Requirements '; 22 4 ?-159b rg 4.2.3.1.5 Structural Requirements 4.2-160 4.2.3.1.6 Envi ronmen tal Requi remen ts 4.2-165 4-iic Amend. 44 April 1978
LIST OF FIGURES Continued Figure No. Page No. 4.2-46 Core Fonner Structure 4. 2-328 d 50l 4.2-47 Limited Free Bow Core Restraint Conc e '. 4.2-329 4.2-48 Design Fatigue Curve for Nickel-Chrcaium 4.2-330 yl Alloy 713 4.2-48A Corrosion Rate of Types 304 arid 316 /:ainless 4.2-330a Steel in Flowing Sodium 4.2-483 Corrosion Rate of Inconel 718 in Sodium 4.2-330b 4.2-48C Design Factor for Estimating the Decrease in 4.2-330c Rupture Strength of Type 300 and 316 Stainiess Steel from Surface Interactions with Sodium 4.2-48D Carbon Equilibrium Values for Type 304 and Type 4.2-330d 316 Stainless Steel as a Function of Temperature 4.2-48E Curve for Determining Averace Type 304 and 316 4.2-330e Stainless Steel Interstitial Concentration as a Function of Sodium Exposure Time and Component Thickness 4.?-48F Diffusion Coefficients for Carbon in Type 304 and 4.2-330f Type 316 Stainless Steel 4.2-48G Calculation of Stress-Strain Curves for Type 304 4.2-3300 and 3i6 Stainless Steel 4.2-48" Calculation of Interstitial Gradients in Sodium- 4.2-330h Exposed Austenetic St,inless Steel 4.2-48I Effect of Carbon and Ni vogen Concentration on the 4.2-330i Fatigue Strength of Type '16 Stainless Steel at Room Temperature 4.2-48J Rupture Strength Design Allowances for Interstitial 4.2-330j Transfer Effects in Type 304 and Type 316 Stainless Steel 4.2-48K Effect af Carbon Conceneration on the Fatigue 4.2-330k Strenr,th of Type 316 Stainless Steel at 1200 F 44 1 4.2-49 CSS / Reactor Vessel Stress Evaluation Locations 4.2-331 4.2-50 Core Support Structure (CSS) Concept 4.2-332 4.2-51 CSS Geometry Used in Analytical Model 4.2-333
')
L- hJ y' 13~ t L_ J 4-xv Amend. ';0 June 1979
LIST OF FIGURES Continued O Figure No. Page No. Minimum Ligament Section in the Core Support Plate 4.2-334 4.2-52 CSS Ligament Models 4.2-335 4.2-53 CSS / Reactor Vessel Axisymmetric Model with Heat 4.2-336 4.2-54 Transfer Boundaries Identified Comparison of RV-lE and RV-4E Thernal Transients 4.2-337 4.2-55 Displacement Plot for Unit Design t,P's 4.2-338 4.2-56 4.2-57 Displacement Plot for Dead Weight ,oads 4.2-339 Displacement Plot Due to Lateral SSE Loads 4.2-340 4.2-58 Design Fatigue Curve for 304 Stainless Steel 4.2-341 4.2-59 Cone Geometry used in Buckling Analysis 4.2-342 4.2-60 Lower Inlet flodule Wall Structural flodel 4.2-343 4.2-61 Coolant Screen and Header Model 4.2-344 4.2-62 Receptacle Model 4.2-345 4.2-63 4.2-63A One-Eighth BMFM 3-D Finite Element Thermal flolded Geometry 4.2-345a 4.2-638 BPFM Critical Sections in the One-Eighth B?Fri 3-D 44 Finite Element Stress flodule 4.2-345b 4.2-346 4.2-64 Duty Cycle-535 Trips 4.2-347 4.2-65 Creep / Fatigue Damage in Ir.conel 718 (Thickness = 6 inches) RV-IU Transient 4.2-348 4.2-66 Striping Evaluation (200 F Stream-to-Stream t.T) 2bO i i 4-xva Amend. 44 April 1978
LIST OF FIGURES (Continued) Figure No. Pa_ge No. 4.2-87 Equilibirium Cycle Assembly Two-Dimensional 4.2-369 Temperature and Flux Distributions at Core Midplane 4.2-87a Illustration of Effects of Core Thennal 4.2-369a and Flux Gradients in Radial Blanket Rea (not to scale) 4.2-88 Assembly Bows and Interface Loads at Start of 4.2-370 First Cycle at Operating Temperatures 4.2-89 Assembly Bows and Interface Loads at End of 4.2-371 First Cycle 4.2-90 Assembly Bow and Interface Loads at End of 4.2-372 First Cycle at Refueling Temperature 4.2-91 Assembly Bow and Interface Loads at End of 4.2-373 Second Cycle Operating Temperature 4.2-92 Assembly Bow and Interface Loacs at End of 4.2-374 Second Cycle Refueling Temperature 4.2-92a Bowing Displacements and Reactivity. Change versus 4.2-374a Power to Flow Ratio 4.2-92b Bowing Reactivity versus Power to Flow Ratio for 4.2-374b 44' Various Analysis Assumptions 4.2-93 Control Rod Systems Scram Insertion Pequirements 4.2-375 4.2-93a Calculated Axial Profile of B C Captures in the 4 4.2-375a 34 Secondary Control Assembly 4.2-94 Secondary Control Rod System Scram Insertion 4.2-376 Requi remen t. 4.2-95 Deleteo 4.2-377 4.2-95a Control Rod System Maximum Misalignment Sources 4.2-377a 4.2.95b Control Rod System Maximum Sources for the Reactor 4.2-377b 50 Refueling Condition for the Reactor Operating Condition 4.2.96 Cor. trol Rod System Histogram (5 year) 4.2-373 @ ') I 4-xvil
'69J
/- 'I L J Amend. 50 June 1979
l.IST OF FIGURES (Continued) Figure No. Page No. 4.2-97 Reactor Steady State Thermal Environment 4.2-379 1 4.2-97A Exposure Time Required to Cause Room Temperature Toughness to Fall Below 15 Ft. Lbs. 4.2-379a 4.2-978 Exposure Time Required to Deplete Room Temperature Toughness by One Half 4.2-379b 25 4.2-98 Control Rod Systems - Reactor Elevation 4.2-380 4.2-99 Control Assenhly Positions in Core Layout 4.2-381 44 4.2-100 Primary !!echanical Sub-System 4.2-382 4.2-101 Primary Control Rod System Layout 4.2-383 4.2-102 Primary Control Rod Drive Mechanism 4.2-389 4.2-103 Lower primary CRDM and Driveline Assembly 4.2-390 4.2-104 Primary Control Assembly 4.2-391 268 126 Amend. 50 June 1979 4-xviia
LIST OF FIGURES Continuea Figure flo. Page fio. 4.2-105 Secondary Control Rod System Schematic 4.2-392 4.2-106 Secondary Control Assembly (SCA) 4.2-393 4.2-107 Secondary Control Rod Driveline (SCRD) 4.2-394 4.2-108 Secondary Control Rod Drive Mechanism (SCRDM) 4.2-395 44l 4.2-109 Diagram of the Control Rnd System flodel Used for 4 . 2- 3 i'8 Evaluation of Retardation Forces due to Lateral flisalienments 4.2-110 Control Assembly Outer Duct Bowing 4.2-399 4.2-111 Primary Control Assembly Duct Bowing 4.2-400 Configurations I 4 2 -Illa Pellet Sweliing and Clad Gap 4.2-400a 34 4.2-Illb Primary Control Rod Axial Burnup Profile t.2-4n0b
^ 2-112
. Primary Rod Control System Scram Insertion 4.2-401 Distance vs. Tine 4.2-113 Comparison Analysis Tests for FFTF Control 4.2-402 Assembly (flisaligned Conditions) 4.2-114 First Cycle Full Power, Preliminary Primary 4.2-403 Control Rod System Scram Insertion Rates 4.2-115 Equilibrium Cycle, Full Power, Preliminary 4.2-404 Primary Control Systen Scram Insertion Rates 4.2-116 Deleted 4.2-405 4.2-117 Deleted 4.2-406 50 4.2-118 Deleted 4.2-407 4.2-119 Scra"1 of Typical Primary Control Rods-SSE Seismic 4.2-408 Based vs. Normal Conditions 4.2-120 Absorber Assembly Scram Impact Dynamic Model 4.2-409 4.2-121 SCRS Rod Displacement vs. Time 4.2-410 4.2-122 SCRS Scram Reactivity Insertion vs. Time 4.2-411 2bh / cC' ,
Amend. 50 4-xviii June 1979
LIST OF FIGURES _ Continued Figure No. Page No. 4.3-30 Criticality of Fuel Assemblies 4.3-119 4 . 3- 31 Schematic of Feedback Effects 4.3-120 4.3-32 Transfer Function for Various Feedback 4.3 ~121 Reactivities 4.3-33 Transfer Functions for Various Feedback 4.3-122 Reactivities 4.3-33a Bowing Reactivity Function for Stability Analysis 4.3-122a 4.3-122b Reactor Power Response to Inherent Reactivity Feedback Following a 2c Step Insertion 4.3-122b 4.3-33c Fuel Assembly Maximum Fuel Temperature Response to Inherent Reactivity Feedbacks Following a 2c Step Insertion 4.3-122c 4.3-33d Fuel Assembly Maximum Cladding Temperature Response to Inherent Reactivity Feedbacks Following a 2c Step Insertion 4.3-122d 4.3-122e Fuel Assembly Maximum Coolant Temperature Response to Inherent Reactivity Feedbacks 44 Following a 2c Step Insertion 4.3-122e 4.3-34 Radial Neutron Flux Distributions 4.3-123 4.3-33 Typical Axial Flux Shape in the Core and Blankets 4.3-124 4.3-36 Calculational Scheme for Analysis of CRBRP 4.3-125 4.3-37 Calculational Scheme for Evaluation of ZPPR 4.3-126 Critical Experiments 4.3-38 ZPPR-3 Calculational Flow Diagram 4.3-127 4.3-39 I in Zones and Central Plate and Pin Cells for 4.3-128 ZPPR-2 4.3-40 Core Layout for ZPPR-3 Phases 1B, 2 and 7 4.3-129 4.3-41 Reference 632 Drawer Sodium Void Confie. ration 4.3-130 for the Modified Phase 3 Core of ZPPR f.s./mbly 3 4.3-42 ZPPR-3 Phase 3, Modi #ied Core, Two-Dimensional 4.3-131 (R,7) Calculation Model 4-x)xP V;p 1icv
~, n Amend. 44 April 1978
/ LIST OF FIGURES Continued Figure flo. Page No. 4.3-43 ZPPR Assembly 4 Phase 1 Reference Configuration 4.3-132 4.4-1 Typical Fuel Rod Axial Temperature Profile 4.4-54 4.4-2 CRBRP Flow, Pressure Drop, Hydraulic Loads and 4.4-55 Coolant Temperature Distribution for Plant Thermal / Hydraulic 4.4-3 Inlet Module Core Map, Showing Orificing 4.4-56 Requirements of Reactor Assemblies 50l 4.4-3A Lower Core Support Plate Pressure Manifold System 4.4-56a 4.4-4 Modular CSS Assembly: Flow Paths 4.4-57 4.4-5 Elevation of Typical Lower Inlet Module 4.4-58 50l (_ t)oO IiLw/ Amend. 50 June 1979 4-xxia
LIST OF FIGURES Continued Figure flo. Page No. 4.4-6 Inlet Module For Bypass Flow 4.4-59 l20 50 4.4-7 Radial Shield Orificing Flow Path 4.4-60 4.4-8 Upper Internal s Structure Concept Fea <ure 4.4-61 4.4-9 Orificing Scheme and Nominal Flow Rates at 4.4-62 Flant T&H Design Conditions 4.4-10 Equilibrium Fuel Assembly and Radial Blanket 4.4-63 Assembly Management 4.4-11 Flow Maldistribution in the Five Fuel Assembly 4.4-64 Orificing Zones at a Power-to-Flow Ra tio of 1.0 4.4-12 Flow fialdistribution Factors in the Four Radial 4.4-65 Blanket Orificing Zones and the Central Assemblies at a Power to Flow Ratio of 1.00 4.4-13 Flow fialdistribution Factors in the Hot Assembly, 4.4 6C Average Fuel Assemblies, and Average Radial Blanket Assemblies at a Power to Flow Ratio of 1.000 4.4-14 Flow Maldistribution Factors in the Hot Assembly, 4.4-67 Average Fuel Assemblies, and Average Radial Blanket Assemblies at a Power to Flow Ratio of 2.000 4'4~CO 4.4-15 Effect of Buoyancy on Hot T.hannel Enthalpy Rise in the Hot Assembly a: Low Reactor Flow Rates 4.4-69 4.4-16 Graphical Illustration of Seni-Statistical Method 4.4-17 Fuel and Radial Blailket Assemblies Linear 4.4-70 Power Rating at Equilibrium BOL- Conditions 4.4-18 Envelope of Fuel anc' Radial Blanket Assenblies 4.4-71 Linear Power Rating at Equilibriun E0L Conditions 4.4-19 Envelope of Fuel and Radia. Cla. set Assemblies 4.4-72 Mixed Mean Outlet Temperatures for P' ant Thermal Hydraulic Contitions at Equilibrium BOL Amend. 50 4-xxii 26g June 1979 e]ur
Bases - The prime objective of this requirement is to assure the functional capability of all core components is preserved. It is possible to postulate a number of potential failure mechanisms associated with a general condition of duct to duct contact in the active core region. One such failure mechanism is duct brittle fracture under seismic loading. Anc ther would be excessive Afor-mation which could occur due to buildup of large interduct co.itact loads. By not allowinq duct contact to become a general condition, the functional capability of all core components is preserved. The consequences of local duct to duct contact wil' be evaluated. i) Requirement - The design life for the core formers is 30 years for the 75% plant capacity factor. Bases - This requirement assures that the core fonners will not need to be replaced during the design life of the plant. The primary incentive for this requirement is economic and operational benefits. j) Requirement - Design of the core restraint system shall be consistent with annual refueling. Bases - This requirement provides consistency with established plant operating guidelines, k) Requirenent - The dynamic radial deflection of the inner profile of the Upper Core Former (UCF) shall be less than + 0.100 inch, relative to the LCF centerline during an SSE or OBE event. Bases - Proper alignment of the control assembly handling socket is necessary to meet protection system (SCRAJ1) insertion rate requirements. The insertion rates required during a seismic event are slightly lower than those required during non-seismic events (see Figures 4.2-93 and 4.2-94). For non-seismic events, control rod system alignment requirements 44l are defined by Figures 4.2-95A and B. The contribution from the core re-straint system and core former clearances and tolerances is limited to less than 0.530 for the control assembly handling socket to reactor vessel centerline. However, for seismic events, a similar alignment limit was considered not to be applicable, since dynamic analysis is being performad to adequately account for deflections of each component interfacing with the control rod system. Qualita tively, the objective for the core former is to limit seismic deflections to as small as practical, within the various design restrictions. The 0.100 inch dynamic deflection limit is thus established as a reasonable design objective for such a iarge (150 inch 0.D.) ring. It should be pointed out 14 Amend. 44 4.2-95 268 l}l April 1978
O that the UCF time vs. deflection history in relation to that for the control rod system and other interfacing components is also an important parameter in the scram analysis, and is not easily factored into a simple deflection requirement.
- 1) Requirement - The Lower Core Former (LCF) radial deflection relative to the LCF centerline during an SSE or OBE shall be limited to less than
.100 inches.
Bases - The Lower Core Former, interfacing with the above core load pads on the reactor assemblies, performs a geometry control function. To limit the seismic loads on the assemblies and to limit the magnitude of and reactivity insertion during ' seismic event, the LCF deflection must be limited. If the distance across the core former were significantly reduced due to an elliptical deflection of the ring, the reactor assemblies would be compacted, causing a reactivity insertion and higher loads on the load pads. By limiting the LCF deflection to less that 0.100, an increase in load pads is avoided since adequate clearances exist between load pads and between the outermost load pad and the Lower Core Former. This is shown in Figures 4.2-84 and 4.2-92. Preliminary ctre restraint models indicate that cumulative clearance is slightly more than 0.150 in;h2s between the first row blanket assembly load pad and the LCF segment for on-power operation. Therefore, if the LCF deflects up to 0.100 inches, the compaction would primarily affect the shield and outer blanket assemblies and would not be expected to result in a significantly greater seismic reactivity insertion. During low power operation, the gap existing in the outer core region is less than for on-power operation due to the outward assembly bow. This could result in a slightly larger seismic reactivity insertion; however, the allowable reactivity insertion is also larger for low power operation. Further evaluations are planned to better determine the effect of core former deflection on seismic reactivity insertions. These evaluations will lead to an improved definition of LCF deflection requirements. There will be a seismic induced core compaction in an orthogonal direction which will occur even with zero LCF deflection. This is discussed in Section 15.2.1.3. The intent of the LCF deflection limit is to prevent an addition-al source of core compaction during a seismic compaction event. 14
-n 3 kJ" 50 Amend. 50 4.2-95a U""
i A
" t D- +
ti T d D To meet criterion: D e
# 0 (7) A cycle limiting criterion is required to verify the applicability of the modified rule. The effective number of allowable design cycles is:
I n = e nlD\ qD, g Where n is the total number of significant strain cycles between hold periods. Low amplitude high cycle strain fluctuations (such as normal power fluctuations) need not be considered in n if they are elastic excursions that result in negligible fatigue damage. For the modified rule to be applicable, n shall not exceed 3000 for type 316 stainless steel nor 6000 for typ8 304 stainless steel. Modification of Creep Damage Rules In cases where a local stress concentration exists, the creep-fatigue damage evaluation may be modified as described herein. (1) The material is austenitic stainless steel Type 304 or 316 solution treated. (2) The structure does not require a Code Stamp under existing Code rules. (3) Simplified or rigorous inelastic analysis is used. (4) Strest rupture test data of the same type of stress concentration with similar geometric proportions tested at prototypic temperatures are used as a basis for modification of the Code Strength. The test temperature may be higher than the service temperature in order to more closely simulate the actual component lifetime and the stress level. (5) The notched stress rupture data shall be from specimens which are comparably or more severely loaded than the component, i.e., membrane loading of a notched specimen should be more severe than a gradient loading. (6) The stress rupture test data include data up to 1/60 of the component lifetime at prototypic temperatures or the equivalent when a short-time high temperature combination is used to simulate the desired long-time 49 service environment.
') , _ n -
'dd 4.2-ll5d Amend. 50 June 1979
(7) Subject to the above limitations, the creep damage may be calculated in accordance with F9-4T and Code Case 1592 as modified. The modi- g fication is to use a peak stress to rupture design curve based upon W the stress to rupture design curve in Code Case 1592 adjusted for the influence of a non-linear stress state caused by the presence of a geometric stress concentration as with the following: Step 1 - Determine the smooth specimen stress rupture strength curve by tests of the same material at the temperature of interest. Step 2 - Determine the stress rupture strength curve with the presence of the geometric stress concentrations under the same conditions in (1) with specimens of the same heat of material with the same histories. Analytically determine the peak stress relative to the net stress thus defining the stress rupture strength in terms of " peak stress" vs. time to rupture. Step 3 - Ratio Code Case 1592 stress to rupture design curve by the ratio of Step 1 divided by Step 2. This must be done for at least 3 points in time with a separation in time of at least two orders of magnitude. In cases where the strength ratio varies with lifetime, the lesser of the value extrapolated to the component lifetime or the experimental value for the longest duration tests shall be used. (8) The total creep-fatigue damage is determined by adding to the creep damage and fatigue damage calculated in accordance with T-1411, -1412,
-1413, and -1414 of Code Case 1592.
(9) The allowable creep-fatigue damage (D) is determined from the lesser of the values from Figure T-1420-2 of Code Case 1592 (See Figure 4.2-and an average of test values from creep-fatigue interaction tests of notched specimens. (10) The greater of the damage using the modified rule and the damage using the stress unaltered by the stress concentrations and the Code Case 1592 49 stress to rupture design curve shall be used. 4.2.2.3.3 Additional Material Properties 4.2.2.3.3.1 Inconel 718 Fatigue Properties The Alloy 718 design fatigue curve, Figure 4.2-48, proposed for inclusion in the NSM Handbook as interim data, shall be used until super-ceded. The effects of the fabrication processes and service environment 41 on the structural integrity of the UIS shall be considered. The effect 4.2-ll5e 26[] )) emend.50 dune 1979
VIS ':EYUAY m R R 1 KEY. #
~ }/ _
t (
/
, - 1 NTAL (J y [ G 'T [!ril F q i i UPPER CORE FORMER y f' SUPPORT RING LOWER CORE l' N FIXED h CORE BARREL RADIAL SHIELD Figure 4.2-46. Core Former Structure Amend. 50 4.2-328d June 1979 268 135
traps are air ccoled, and process the sodium at a flow of 60 gpm. The single trap is sufficient to limit oxygen and hydrogen to a maximum of 2 and 0.2 ppm respectively, and to limit the tritium concentration in the sodium to 0.062 pCi/gm (corresponds to a transport of tritium to the steam generator water of 0.016 Ci/ day). A more detailed description of the Intermediate SC Sodium Processing System is given in Section 9.3.2.4. 5.4.3 Design Evaluation. 5.4.3.1 Analytical Methods and Data The design of the IHTS is based upon technology attained during the development, design, construction and operation of sodium systems of similar type IHTS hydraulic and fluid volume changes are calculated and based upon sodium thermodynamic data from " Standard FFTF Values for Physical and Thermophysical Properties of Sodium" (Ref. 2) The IHTS sodium pressure losses are calculated using Darcy's formula, the general equation for pressure drop of fluids. The factors for equivalent lengths of fittings and pipe friction are based upon " Tube Turns," Bulletin TT725 (Ref. 3) and " Friction Factors for Pipe Flow," by L. F. Moody (Ref. 4), respectively. 5.4. 3.1.1 Structural Evaluation Plan (SEP) The procedure for developing SEP's for the IHTS is the same as described for the PHTS in Section 5.3.3.1.1. 5 . 4 . 3.1. 2 Stress Analysis Verification The IHTS stress analysis verification procedures and methods are the same as those described for the PHTS in Section 5.3.3.1.2.
- 5. 4. 3.1. 3 Compliance with Code Requirements _
The design of the IHTS will be in compliance with code requirements as identified for the PHTS in Section 5.3.3.1.3. 5.4-12 Amend. 50 June 1979 @ "1 ! r .. b )9
AFP Motor Drives These motor drives will be synchronous speed squirrel cage induc-tion motors of 980 horsepower. These motors will be selected from a 17 vendor's standc.rd line and no special requirements are anticipated. AFP Turbine Drive This component will be ebtained from an experienced vendor and will be sized to produce 1960 horsepower. The turbine will be constructed with 17 sufficient quality assurance coverage to assure its reliability during ser-Vice. The auxiliary feedpump turbine is not keot hot for quick start operation. The drive turbine concept sclected for the Auxiliary Feed eump is based on the capability of thic turbine to withstand severe service con-ditions. This is accomplished by constructing the turbine wheel from a single forging with buckets milled into the forging. The start-up procedure is similar to thct for the RCIC turbine in a BWR in that it will occur without pre-warming. 25 Pump Integrity The auxiliary feed pumps will be designed to the requirements of ASME B&PV Code, Section III, Class 3. In addition, the pumps and their supports will be designed to Seismic Category 1 requirements. Allowable stress limits are specified in Table 3.9-3 arid pressure limits are speci-i
)7 fied in Table 3.9-4.
5.6.1.2.3.3 Protected!!aterStoragTan_kJPWSTJ The PW5T holds the protected water to.be supplied to the steam drums in the event of loss of normal f eedwater er normal heat sink. The size is deteroined by detailed analysis of the heat removal conditions during the first several hours af ter shutdown and by anticipated component leakage rates. The ' nk will be constructed to the requirements for en ASf1E Section III/ Class 2 vessel and it will operate at low temperature (< 200 F) and low pressure (<l5 17 psig), 268 1R Amend. 47 5.6-6 a Nov. 1973
5.6.1.2.3.4 SGAHRS Piping and Support The SGAHRS riping is described below and is shown in Figure 5.1-5. The SGAHRS piping will be designed in acsordance with the ASME
' Code Section III as specified in Section 5.f. l .l .2. The material 49l specifications are discussed in Section 5.6.1.1.4.
The SGAHRS piping runs can be categorized as follows:
- a. PWST Fill l_ine This 3 inch low pressure, low temperature, class 3 carbon steel line runs from the 10 inch alternate water supply line through the motor-driven, normally closed PWST fill valve to the PWST inlet.
- b. Protected Water Storage Tank (PWST) to Auxil'ary Feedwater Pump (AFP) Inlet There are three low pressure, low temperature, uninsulated carbon steel lines from the PWST to the three auxillary feedwater pump inlets. Two of the lines, each of which leads to a half size, motor-driven pump are 6 inches in diameter and the third line to the full size turbine-driven pump is 8 inches. All three lines contain a manually ooerated, locked open valve and an electrical operated, normally open isolation valve. These lines are Class 2 from the PWST to the electrically-operated isolation valve 49 l and then Class 3 to the pump inlet.
- c. Alternate Supply Line to AFWP Inlet The alternate supply line provides the capability for the AFW pumps tc take suction from the condensate storage tank.
A 10 inch carbon steel line runs from the condensate tank junction to the first branch line. An 8 inch branch line 49l passes through an electrically-operated, ncrmally closed isolaticr valve and tees into the 8 inch turbie pump inlet piping. Two 6-inch branch lines each pass through c.;ctrically-49 operateJ, normally closed isolation valves and then tee into the 6 inch motor-driven pump inlet piping. The total run of niping is Class 3.
- d. Auxiliary Feedwater Pump Discharge to Discharge Header (Inclusive)
The 6 inch carbon steel turbine pump discharge litie leads to a 6 inch discharge header. This header in turn has three di - charge points, one to each steam drum feedwater supply loop a 6 inch carbon steel line from each motor dri'len pump feeds into a 6 inch header which also has three discharge points. ona 43 17 to each drum. Amer.d. 50 June 1979 5.6-7 n 268 1w
5.6.2 Direct Heat Removal Service (DHRS) 5.6.2.1 Design Bases 5.6.2.1.1 Performance Objectives The Reliability Program, discussed in Appendix C, will provide 41 verification that SGAHRS removes residual heat following a reactor shut-down with a high level of reliability. Hence it is judged that only stcam and feedwater trains backed by SGAHRS are necessary to safely and reliably remove residual heat following shutdown from three loop, full power operation. To enhance the reliability of decay heat removal, the DHRS provides a fourth redundant heat removal path and heat sink. The impact on overall shutdown heat removal reliability by inclusion of the 41 DHRS is being determined by the Reliability Program described in Appendix C. The DHRS provides this supplementary capability by satisfying the following objectives:
- a. Function to remove reactor decay heat following reactor shutcown from three loop rated power operation, assuming loss of all hea*
transfer through the IHX's at the time of reactor trip. Oncra-tion of three primary pumo pony motors and maximum reactor decay power is assumed. Under these conditions, the DHRS is desianed to provide sufficient cooling to ensure primary coolant boundary integrity and prevent loss of in-place coolable geometry of the core. To meet this obiective, DHRS components will be sized such that, under these conditions, the average bull p5imary sodium tennera-ture will be limited to aoproximately 1140 F. Caoability will be provided to pernit remote manual initiation of DHRS from the Control Room. The overflow and makeup circuit and the snent fuel cooling system will be able to be cross connected in a manner which permits both EVST NaK cooling trains to be used 20 when DHRS is removing full capacity heat load.
- b. Accommodate the thermal transients resulting from normal, upset, emergency and faulted plant events in which continued performance of its function is not impaired.
- c. Accommodate floods (Section 3.4), tornadoes (Section 2.3), missiles (Section 3.5), and earthquakes (Section 3.7), in which continued performance of its function is not impaired.
- d. Function in a manner which will not signiricantly reduce the reliability and availability of the EVST heat removal chain.
This objective requires the EVST NaK cooling circuits to be designed to remove concurrently the heat generated by the spent fuel and the reactor decay heat. 263 13? Amend. 43 5.6-20 Oct. 1977
5.6.2.1.2 A_pplicable Code Criteria and Cases The components of the DHRS shall be designed, f abricated, erected, constructed, tested and inspected to the standards of Section III of the ASME Code,1974 edition through the summer 1974 Addenda, Class 1 or 2 as 26 listed in Table 5.6-12. Applicable code cases will be used to supplement the design analysis required by the ASME Code. 5.6.2.1.3 Surveillance Requirements The need for surveillance of the DHRS piping and components will 26I be determined as the system design progresses and as the need to monitor austenitic stainless steel is determined by ongoing programs. If a require-ment is identified, a surveillance program will be designed in accordance with the philosophy of 10 CFR 50, Appendix H. E.6.2.1.4 Material Considerations High Temperature Design Criteria High temperature components in the DHRS will be analyzed in 26l accordance with the requirements specified in ASME Boiler and Pressure Vessel Code, Section III, as supplemented by the applicable code cases and RDT standards. Material Specifications Stainless steel materials which satisfy the requirements of the ASME Code will be specified for use in the DHRS system, as noted in Table 26 5.6-13. 49I 5 .6.2.1.5 Leak Detection Requirements Th< DHRS will be monitored for sodium and NaK leaks and leak 26 indication will be provided in the Control Room by the leak detection system described in Section 7.5.5. 5.6.2.1.6 Instrumentation Requirements DHRS is remote manually activated and controlled from the Control Room. Instrumentation required to monitored the condition of the DHRS consists of thermocouples on the EVST sodium outlet lines (3 loops) and level indicators in the EVST and the Reactor Vessel (RV). These instruments confirm that the sodium levels in the RV remain above the loop outlet nozzles 50 46 and that temperatures remain below design limits. Other DHRS diagnostic instrumentation is not essential for DHRS operation as the pumps and air blast heat exchanger are being operated at maximum design rates. When the reactor decay heat load has dropped sufficiently, the cooling capacity of the system may manually be reduced by lowering flow-rates or fan speed, or by - shutting down one o ne of the EVST cooling trains. . 26 Amend. 50 5.6-21 June 1979 268 140
TABLE 5.6-5 SGAHRS VALVE CLASSIFICATION NORMAL OPERATING VALVE ACTIVE / INACTIVE POSITION MODE Alternate AFW Supply Active Open Isolation 44l PWST Fill Inactive Closed Isolation PWST Drain Inactive Closed Isolation PWST Level Indicator Inactive Open Isola tion AFW Pump Inlet (Manual) Inactive Open Isolation 50 AFW Pumo Inlet (Electrical) Active Open Isolation Alternate AFW Pump Inlet Active Closed Isolation 26 Isolation Pump Recirculation Active Closed Pump Discharge C/V Active Check Pump Discharge Isolation (Manual) Inactive Open Isolation AFW Supply Isolation (Manual) Inactive Open Isolation AFW Supply Control Active Onen Flow Control 50 AFW Supply Isolation (Electrical) Active Closed Isolation AFW Supply Isolation (Manual) Inactive Open Isolation AFW Supply C/V Active Check 50 AFW Supply C/V Active Check Steam Drum Vent Control Active Closed Vent, Pressure Control 26 Superheater Vent Control Active Closed Vent, Pressure Control PACC Steam Supply Inactive Open Isolation PACC Steam Supply Bypass Inactive Open Isolation PACC Condensate Return Inactive Open Isolation PACC Noncondensible Vent Active Closed Vent PACC Noncondensible Vent Isolation Inactive Open Isolation Drive Turbine Steam Supply Isolation
- 50) (Electrical) Active Closed Isolation 26 Drive Turbine Steam Supply C/V Active Check Drive Turbine Steam Supply Isolation Inactive Open Isolation (Manual)
Amend; 50 5.6-34 26,3 1j, une 1979 e , i
TABLE 5.6-5 (Cont'd) NORMAL OPERATING VALVE ACTIVE / INACTIVE POSITION f t0DE __ Drive Turbine Steam Supply Pressure Centrol Active Closed Pressure Control Pressure Instrument (Pump In.et) Inactive Open Isolation Pressure Instrument (Pump Discharge) Inactive Open Isolation Pressure Instrument (Turbine Inlet) Inactive Open Isolation 50 26 0 k ' t '-
'2 thi))
O 5.6-34a Amend. 50 June 1979
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- 2. A Heating, Ventilation and Air Conitioning (HVAC) System to keep the main control room slightly pressurized at all times and at temperatures, humidities, and air purity levels adequate for conducting safe, efficient plant control operations.
- 3. A low leakage enclosure for the main Control Room and its adjoining rooms to provide the capability for keeping a positive air pressure level with~n the enclosure.
- 4. Two alternate and widely separated air intakes and redundant filter units, to limit the amount of contamination entering the Control Room.
- 5. Airborne hazard monitors that detect unsafe concent .tions of smoke, toxic chemicals and unsafe radiation levels, annunciate the presence of the hazard and automatically 22 49
- ransfer the HVAC system to its accident mou_ of operation.
- 6. Appropriate fire suppression equipment.
- 7. Of fice and living accommodations appropriate for long term occupancy.
- 8. An amply stocked inventory of emergency equipment and supplies.
- 9. Installation of two door vestibules to prevent unfiltered air entering the main Control Room.
49l The habitability system is also required to permit continuous control room occupancy under accident conditions. These provisions enable the operators to remain on duty without relief for as long as required. It is, therefore, not necessary to plan on routine shift changes by the control room personnel. However, if conditions necessitate a change in personnel, this can be accomplished without undue radiation exposure. The calculated radiation exposure for either ingress or egress 49 at 24 hours af ter the accident is less than 1 mrem. This estimate is based on direct exposure from fission product gases, halogens, volatile solids, fission products, and activated sodium evenly distributed through-out the reactor containment building free volume. These sources consti-49 tute direct shine dose only. 25 The ingress or egress exposure resulting from the radioactivity (annulus leakage previoualy stated) is approximately 3.50 mrem external exposure (whole body dose). The internal exposure to the lungs, thyroid, 49 and bone is less than 1.5 mrem, 9 mrem, and 28 mrem, respectively. 6.3-2 Amend. 49 April 1979
?40
<uO 1 *r I o s.
The operator ingress or egress is to be made by motor vehicle to a point adjacent to the control building portal. The car is assumed to travel at an average speed of 15 mph in both directions. The operator walks between the control building and the motor vehicle at an average rate of 4 mph, and personnel will be equipped with supplied air breathing apparel. 25 49 6.3.1.2 System Design The Control Room Habitability System is designed to provide a safe, comfortable and appropriately equipped location for personnel controlling plant operations during normal times and during accidents. Features incorporated into this Habitability System to assure these aspects are described below. A concrete enclosure and special sealed doors are important features in this habitability system. The details of this shielding and 22 the shield wall thicknesses are described in Section 12.1.2.4. Bases used in the design and analyses are also presented in Section 12.1.2.4. Factors considered in these design analyses include thermal margin beyc. design basis requirements and the associated activity releases 49 and gamla shine from the containment / confinement. In addition, other features, such as two widely separated air intal.es are provided to mitigate the consequences of low probability accidents beyond the design 22 basis. Another important feature of the Control Room Habitability System is the HVAC system. The HVAC system for the Control Room contains two 100: capacity air conditioning units and two 1001 capacity exhaust fans, with one unit each normally operating and the other unit on standby; and two 100? capacity emergency air supply filter units. Complete details of the system are presented in Section 9.6.1. A P&ID of the system is also provided and shown in Figure 9.6-1. This HVAC system contains several aspects that are significant to the Control Room Habit-ability System. One of these aspects is that this system has full capacity redundancy to assure the capability for controlling the environ-ment af ter any single component or subsystem failure. A second aspect is the capability to maintain the Control Room at a slightly greater pressure than any of its surroundings at all tines. A third aspect of interest to the habitability system is the air cleanup units that purify both make-up and recirculated air flows during postulated accidents. Fach air cleanup unit in the HVAC system contains two banks of HEPA filters and a bank of carbon < arbero. The filter capability is discussed in Section 9.6. A fourth aspect is the capability for selecting emergency pressurizing air during accidents from widely separated intakes, one located at the SE Corner of the Control Building roof at approximately elevation 880' and the other at the NE corner of the Steam Generator 4f Building Auxiliary Bay at approximately elevation 858'. This along with instrumentation provided, allows selection of the cleaner air source during such periods. Amend. 49 April 1979 6.3-2a 268 } i >
LIST OF TABLES TABLE NO. PAGE 7.1-1 Safety Related Instrumentation and 7.1-7 Control Systems 7.1-2 List of Regulatory Guides Applicable 7.1-8 to Safety Related Instrumentation and Control Systems 7.1-3 List of IEEE Standards Applicable to 7.1-9 Safety Related Instrumentation and Control Systems 7.1-4 List of RDT Standards Applicable to 7.1-10 Safety Related Instrumentation and Control Systems 49
~
- 7.1-6 Safety Related Electrical Instrumentation 7.1-13 23 and Control Equipment 7.2-1 Plant Protection Syster. Protective 7.2-18 Functions 7.2-2 PPS Design Basis Fault Events 7.2-19 7.2-3 Essential Performance Requirements for 7.2-23 PPS Instrumentation 7.2-4 thru 40 24 Deleted 7.3-1 Containment Isolation System Design Basis 7.3-5 7.4-1 Sequence of Decay Heat Removal Events 7.4-9 7.5-1 Instrumentation System Function', and 7.5-34 Summary 7.5-2 Reactor and Vessel Instrumentation 7.5-39 44 7.5-3 Sunmation of Sodium / Gas Leak Detection 7.5-40 Methods 34 49 7.5-4 Post Accident Monitoring 7.5-42 7.6-1 Use of Refueling Interlocks 7.6-4 7.9-1 rontrol Rooin Arrangements 7.9-8 7-ix 26S ' 7.'
' Amend. 50 June 1979
LIST OF FIGURES F IG . II,0. 7.2-1 Reactor Shutdown System 7.2-69 I 7.2-2 HTS Coolant Pump Shutdown 7.2-70 , 7.2-2A Typical Prinvry PPS Instrument Channel 7.?-70a 1 Logic Diagram
/.2-2AA f.SS Bypass Functien Block niagran 7.2-70a 15 7.2-2b Primary PPS Logic Diagram 7.2-70b i
7.2-2C Typical Secondary PPS Instrurrent Chanr.el 7.2-7Cc > Logic Diagran i i 7.2-2D Secondary PPS Logic Diagram 44 7.2-7Cd 'l 7.2-3 Typical Primary Subsystem 7.2-71 7.2-4 Typical Secondary Subsystem 7.2-72 7 ' '9- 5 Functional Block Diagrams of the Flux-Delayed 7.2-73 Flux, High Flux, Flux-Pressure, and Reactor Vessel Level Protective Subsystems 7.2-6 Functional Block Diagrams of Primary Pump 7.2-74 Electrics and Primary to Intermediate Speed Ratio Protective Subsystems 7.2-7 Functional Block Diagrams of the IF.X Primary 7.2-75 Outlet Temperature and Steam to Feedwater Flow Mismatch Protective Subsystems 7.2-8 Functional Block Diagrams of the Flux-Total 7.2-76 Flow, Startup Nuclear, Modified Nuclear Rate, and Primary to Intermediate flow Rate Protective Subsystems
'7.2-9 Functional Block Diagrams of the Steam 7.2-77 Drum Level and Loss of Condenser Vacuum Protective Subsystems 7.2-10 Functional Block Diagrams of the Evaporator 7.2-78 Outlet Sodium Temperature and Sodium Water Reaction Protective Subsystems 7.3-1 Containment Isolation System Block Diagram 7.3-6 2b8 lhb Amend. 44 April 1978 7-x
N C' as Tatale 7.1-6 (
,y Safety Related Electrical Instrumentation And Control Equipment Measured Temp. Press. Humidity Radiation Chemical Seismic Vibration Equipment Parameter Purpose Location EVST and 39l X X X TBD NA X EVST Outlet Na Safe Shutdown RSB 331, 357 X 501 Thermocoupl es Temperature Phi 360 TB3 47l' Thermocouples OH X N a 0utlet Safe Shutdown RCB 1078 X NA X X X X Te aperature Pki 4/
X X X TBD RCB 103,104 X X NA Primary Sodium None DHRS 39l Makeup Pumps X X X TB3 EYST Cooling RSB 357A 360 X NA X EVST Sodium Pumps None
] EVST and Re- RSS 352A, 353A X NA X X X X TBD EVST NaK Pumps None
- 1
% and Fid. Panels actor cooling X X X X TB3 RSS 35ZA, 353A X NA EVST A3HX and None EVST and Re-actor cooling Field Panels X X TBD X X NA X Reactor CHRS RCB 105A 39l M3keup pump None Service Field Panels X X X TBD RSB 352A, 353A NA X EVST Sodium Pump None EVST Cooling X Field Par,els X NA X X TBD CB 431 X NA Control Room None DHRS & EVST Panel cooling con-trol X X TS3 X X NA X Various RCB 1078 46 Valve Operators None c$
2 m e 7 II b 4 Q, b ( k hib. bli b
TABLE 7.1-6 Safety Related Electrical Instrumentation And Control Equipment Sheet 6 cf 9 Measured Equipment Paraneter Purpose Location Temp. Press. Humidity Radiation Chemical Seismic Vibraticn Na R5B Valve Operators None Various 357A, 350 X NA X X X X TED Na K RSB Valve Operatcrs None Various 352A, 353A X NA X X X X TB3 Local Control DHR5 & EVST RSS 311 Panels N ne cooling control Rcs losV X NA x x x x Taa N 46 m co hwl C:~ CO ('i CD EE c3 ..
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'S L b TO NON-PPS + BUFFER LEVEL DETECTOR ELECTRONICS t.00 P =2
~ i SH r/
SL TO MON-PPS + SUFFER LEVEL @ DETECTuR 100p 3 ELECTRONICS a
+
SH Figure 7.2-9 Functional Block Diagrams of the Steam Drum Level Subsystem. One Channel of Three is Shown. 6678-5 Amend. 47 @ 7.2-77 Nov . 1978 26R 101
10 N0h-PPS e buf ll R fl H:'E i? t.10 P L ' - DIIEClCa - Ellt'RONifS 4 LOOP =l UT TO NDH-PPS + tWI!ER llt'PLP/IURf ~ - DfifC10K ElfCIRONICS UT t 00P *2 T O HON-PPS + 6Uf f[ R 1[tvikAIUht DETEC10R IL(CTFONICS
.\
t00P =3 ;9 EVAPORATOR CUTLET TEMPERATURE TO SWRPRS e BJFf E R (SEE f!G. 7.5-6) StHS0R St:Pt hh! Aif k
- [tfCTROFIff TO SWRPRS + SUII E F Op (SEE FIG. 7.5-6)
$[h:0R ^
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EVAL 0RATOR al J gg BUFFER 10 h0N-FPS TO SWRS e sut r I r -- tea , 10 NON-PP5 (SEE f IG. 7.5-6) B'JF T E R i
- ItECIR0hlCS
[V 20rATOR (? SODIUM WATER REACTION PROTECTIVE SUBSYSTEM Figure 7.2-10 Functional Block Diagrams of the Evaporator Outlet Sodium 7,7,73..; Temperature and Sodium Water Reaction Protective Subsystems, One Channel of Three is Shown Amend. 50 7.2-76 June 1979 onn (_ G 1~ I < . -
provide the required time response. The thermowell is also swaged at the tip. Although The thermocouples are spring loaded against the bottom of the well. failures of the wells are not expected, as confirmed by tests and analysis, the head of the thernowell, including the cable penetration, is sealed to pro-vide a secondary boundary for the sodium. Tests have shown that this system will provide a time response less than 5 seconds. Flexible mica, polyimide and fiberglass insulated thermocouple extension wires in conduit are used to bring the signals out of the Heat Transport System Cell . The signals are then routed to the containment mezzanine into reference junctions and signal con-di tioning equipment. The conditioned signalsTne are transmitted to the control Reactor Shutdown System pro-room for the Reactor Shutdown System logic. vides buffered signals to the PCS and DH & DS. Primary _and latermediate Hot and Cold l edegperature The primary and intermediate hot and cold leg temperatures are measured to determine and record operating conditions and to calorimetrically The measurement is made by two Dj calibrate the pert::anent magnet flowmeters. duplex element resistance temperature detectors (RTDs) per loop, installed in thermowells. Although failures of the wells are not expected, as con-firmed by tests and analysis, the head of the thermowell, including the cable penetration, is sealed to provide a secondary boundary for the sodium. The signals from the RTDs are routed to signal conditioning equip-ment which converts the resistance variation to a standard signal level for transmission to the DH & D$. Primary _ and Intermediate Pun;p Discharge Pressure The primary and intermediate puup discharge pressure measurements monitor pump performance and the primary loop / intermediate loop differential pressure. The measurements are made by pressure elements installed in the elevated section of the drain line from the discharge piping of the sodium pump. NaK filled capillaries f rom the pressure elements are connected to pressure transducers which develop electrical s'qals proportional to the pressure. These pressure transducers provide a sehadary boundary if the bel-lows in the pressure elements should fail. The' and1Noned signal is sup-plied to the DH & DS.
.I_ntermediate IHX Gutlet Pressure 43 The intermediate IHX outlet pressure measurement is used to monitor the loop and IHX operational performance history. The measure-43l ments c.re made by pressure elements installed in the intermediate loop piping between the IHX and the superheater. NaK filled capillaries from the pressure elements are connected to pressure tranducers which develop electrical signals proportional to the pressure. The pressure transJucers provide a secondary boundary if the bellows in the pressure elements should fail. The conditioned signal is supplied to the DH and DS.
Abb ib 7.5-7 Amend. 49 Apr. 1979 h %
%w
I_HX Differential Pressure The primary sodio.a pump discharge pressure and the IHX Interme-diate Loop outlet pressure detectors are used to provide a differential measurement of the IHX Primary / Intermediate pressure difference. The dif ferential pressure measurement is alarmed to alert the operator for corrective action to assure intermediate to primary differential pressure 50l49 is maintained above the minimum required. Intermediate Pump Inlet Pressure The intermediate pump inlet pressure measurements provide a signal to onitor pump performance. Used with the pump outlet pressure, the differential pressure across the puitp is obtained. In the primary loop, the reactor pressure is used for this surveillance. The measure-ments are made by pressure elements installed on the piping between the evaporators and the pump inlet. NaK filled capillaries from the pressure elements are connected to pressure transducers which deveiop electrical signals proportional to the pressure. The pressure transducers provide a secondary boundary if the bellows in the pressure elements should fail. The conditioned signal is supplied to the DH & DS. Intermediate Expansion Tank Level Two separate level measurement channels are provided; both channels are used for indication in the control room and DH & DS and for alarm. Alarm channels provide a broad range measurement that covers possible high and low levels during plant operation as well as the IHTS 49 fill level . The DH & DS uses measurements for intermediate loop sodium inventory (see also Section 7.5.5). The level probes are designed to be replaceable. Evaporator Sodium Outlet Temperature Three thermocouple channels are provided to measure the sodium temperature at the outlet of the evaporators in each loop. The thermo-couples are placed just af ter the pipes from each evaporater join to form two single lines. These three signals are conditioned separately and provided to the Reactor Shutdown Systcm logic. The Reactor Shutdown System in turn provides buffered signals to the DH & DS. 7.5.2.1.2 Sodium Pumps Sodium Level Sodium level is measured in each pump tank. The signal provides indication and alarm. The alarm is used to notify the operator cf abnormal operation and allow initiation of action to prevent pump damage. Tre signal is also provided to the DH & DS where it can be used in calculation of sodium inventory. Amend. 50 7.5-8 June 1979 268 13,
7.5.4.1 Design Description _ The following subsystems make up the FFM system.
- 1. Cover Gas Monitoring 34 s subsystem continuously samples the cover gas and determines, through gamma analysis:
- the concentration of selected radioactive fission gases to inform the plant operations staff upon 34 each instance of core or blanket pin cladding failure.
- the concentration of radioactive fission gases to characterize the failed pins as to burnup and other i n fo rma ti on.
- 2. Reat. tor Delayed Neutron Monitoring 34 This subsystem continuously monitors for the presence of fission products in the sodium coolant which decay with the emission of neutrons. A predetermined increase in the neutron signal from the Primary Heat Transport System sodium, above the nonnal background 34 level, is taken as an indication of fuel contact with sodium.
The Impurity Mor.itoring and Analysis System provides verificatien of fuel exposure to the sodium by removing sodium with a grab sampler and by subsequent laboratory analysis for fuel and fission product material.
- 3. Failed Fuel Location Stable (non-radioactive) xenon and krypton isotopes (that aro not fission products) are placed in each fuel and blanket assembly pin. Each assembly has a unique ratio of isotopes which will be released to the cover gas upon failure of a pin in the assembly.
Analysis of a processed sample of the cover gas, using a mass spectrometer, is used to identify the 34 assembly containing the failed pin. The FFM subsystems are described in greater detail in the following sections. A block diagram of the FFM system is provided in Figure 7.5-3. 7.5.4.1.1 . Cover Gas Moni toring Subsvo t m The Cover Gas Monitoring Subsystern receives sample gas 50 from the Inert Gas Receiving and Processing System. The sampling control system is located in a shielded call in the Reactor Service Building. The cover gas passes through a sodium vapor trap and into 3? charcoal delay bed. ,, Amend. 50
/Od, }j,j_ June 1979 7.5-15
The purpose of the delay bed is to increase the concentration of gas (orimarily O xenon and krypton) isotopes of interest, and allnw the argon cover gas to pass through without delay, thus enhancing signal-to-signal background ratio. A fixed-channel analyzer, located in the reactor service building, analyzes the cover gas sample for radioactive xenon and krypton fission products. If a leak in a fuel pin occurs, an increase in activity will occur. This indication of a leak will be annunciated by the Plant Annunciator. A mul ti-channel 34 analyzer, including a mini-computer, analyses the signal from the detector to display the entire gamma spectrum, thus providing burnup and other information characterizing the failed fuel pins to supplement the failure location subsystem through correlation with core and blanket history.
- 7. 5. 4.1. 2 Reactor Delayed Neutron Monitoring Subsystem The Reactor Delayed Neutron Monitoring Subsystem includes a Delayed 34 l Neutron Monitor consisting of an assembly of three BF 3 -filled gas proportional neutron detectors, mounted in a shielded moderator assembly adjacent to each of the three Primary Hea t Transport System hot leg pipes.
Coolant sodium transported past the detector assembly, is contin-uously monitored for delayed neutrons emitted by decay of radioactive precursors in the sodium. The system sonsitivity is dependent on the signal-to-background ratio of the system. Signal is defined as detected delayed neutrons produced by recoil of precursors from fuel exposed by cladding fail-ure, or from fission of fuel washed out into the sodium through a failure. Background is defined as detected neutrons from known sources which are not initially related to failed fuel (fuel pin contamination, fissionable impuri-ties in core structural materia's, fissionable materials in the sodium, and neutrons from the reactor) The shielding and moderator assembly provides 1) reduction of gamma interference f rom Na-24, 2) moderation of neutrons , 3) capability for remote insertion of a calibrated neutron source, 4) capability for insertion and removal of the detector assemblies from the reactor containment building operating floor without deinerting the PHTS cells. S f yrio T s fruiti the three Dr3 detec tors are routed to individual inputs of a preamplifier. Each input of the preamplifier provides individual electronic discrimination against gamma-caused counts from each detector. After discrimination, the neutron-caused counts from the three BF3 detectors, 1 per coolant loop, are comcined in a summing circuit, and routed to a local control 34 i panel located in tne Reactor Containment Building. The local control panel provides controls for remote adjustment of the preamplifier discriminators, a low voltage power supply, a detector-bias high-voltage supply, a comparator for low detector bias voltage alarn and a front panel function switch to select TEST, CALIBRATE or OPERATE modes. These signals are routed to a panel in the main control room displaying the output on three five decade logarithmic count 34 rate meters and a three pen five decade logarithmic strip chart recorder. [ Amend. 34 7.5-16 Feb.1977
9 7.5.5.3.1 Design Description General Steam or wate. leakage into sodium increases the hydrogen and oxygen concentrate in the sodiim stream. The leak detection is based on measurement of the hydrogen and oxygen concentration in the sodium. 47 0xygen concentration itieasurements in the sodium are complementary to the hydrogen detectors, thus providing a diverse method to ensure early de-tection. To provide a sensitive leak detection capability, the background concentrations of oxygen and hydrogen are maintained as low as practicable. The hydrogen in the sodium is removed by cold trap precipitation of sodium hydride. Oxygen in the sodium is removed in the same way by sodium oxide precipitation. The background concentration of hydrogen in the system is 50 normally maintained below 200 ppb through cold trap operation. Oxygen concentration is maintained at about 2 ppm. 33l31 13 Hydrogen Detectors The measurc,ent of hydrogen in the sodium is performed by allowing the hydrogen to diffuse through a thin walled nickel membrane, and detecting the hydrogen with an ion gauge and an ion pump. The monitor may operate in a static mode using the ion gauge to monitor ,teady state hydrogen con-centration, since hydrogen content is directly related to the pressure reasured in the chamber. The monitor may also operate in a dynamic mode, using the ion pump to coastantly pump the chamber since then the hydrogen concentration is directly related to the ion-pump current. 47 34 0xygen Detectors Oxygen electro chemical cells are used to continuously monitor in-sodium concentration and consist of a reference oxygen electrode separated from the sodium by a solid electrolyte. The electrical potential drop between the reference electrode and the sodium measures the in-sodium oxygen concentration. n f, Amend. 50 d U' b fD, June 1979 7.5-26
De tec tors _ L oca t io n_ The three steam generation loops utilize identical oxygen-hydro-gen (0-H) detector modules at the following locations (a) On one evaporator outlet piping downstream, f mixing tee. This leak detection module under normul operation samples sodium that is a combination of sodiun exiting evaporator A and evaporator B. However, by the use of valves in the sanple line, the leak detection module may be utilized to sample sodium exiting either evcporator A or evaporator B. (b) Superheater vent line. This leak detection module under all conditions samples only sodium exiting the :uperheater vent line. (c) On one evapoi'ator vent line downstream of evaporator A/B vent line junction, This leak detection module under normal operation saf"ples sodium that is a combination of sodium exiting the evaporator A vent line and evaporator B vent line. However, by the use of valves in the sample lines, the leak detection module may be utilized to sample sodium exiting either evap-orator A vent line or evaporator B vent line. (d) On one superhea ter outlet piping. This leak detection module under normal oper ation samples sodium that is a combination of sodium exiting both the suoerheater outlets. However, by the use of valves in the sample lines, the leak detection module may be utilized to sample the sodium exiting each sodium outlet indopendently. 47 Figure 5.l~4 shows the instrumentation location. Indication in Control Room M9asurements from the hydrogen and oxygen detectors are monitored by the Data Handling and Display System. Each channel is limic checked and its trend is limit checked. Low level, intermediate level and high level alarms and channel failure alarms are also provided to the Plant 4/ Annunciator System. Sys tede ra ti on. 47 l The leak detector detects two leak signal categories:
- 1. a strong signal in a single pass, Amend. 47 7.5-27 t;ov. 1978
- O
/ O ()
k)V
47 2. detection of a gradual concentration increase or decrease throuqh several passes through the sodium. Figure 7.5-4 illustrates typical first pass hydrogen concentration change as a function of water ! cal rate- A", illustrated, a change in hydrojen concentration of a few ppb would be indicated at the detector for leak rates in the range of 10-4 lb/sec. Approximately one minute is required for the hydrogen to reach the detector and signal a leak. Detection capabilitj can be extended to smaller leak sizes through the use of a rate of rise detection system. Several passes of sodiun through the systen would be required to allow the hydrogen concentration to build up. The sensitivi ty of this system will allow detection of leaks in the range of 10-b lb/sec. Similarly, Figure 7.5-4A indicates the first pass oxygen 4el47 concentration as a function of water leak rate. Figure 7.5-5 illustrates the hydrogen concentration change with time for various sizes of leaks. 1 7.5.5.3.2 De_ sign Ana ly_s i_s A Steam Generator Leak Detection System is provided to comply with CRBRP General Design Criteria 4 which calls for provision of 6 leak detection in the Steam Generators. In order to show how the criterion will be satisfied, a review of leak damage studies is presented with the resulting instrumentation requirements.
)g Amend. 4g 7.5-27a 2{g ; Feb. 1979
Leak Damage Studies Experimental studies have been conducted in tne United States and in Europe over the past ten years which have given a broad base back-ground to the understanding of the behavior of leakage damage effect. Most of the experimental data taken with 2-1/4 Cr- 1 Mo material have been obtained by injecting water or steam through hole type geometrics at selected target configurations (jet leaks). In general, results of these studies have indicated that both adjacent tube wastace and self wastage 45l13 are possible damage mechanisms, as described in Reference 1, 3 and 4. Adjacent tube wastage will occur with the proper leak size and orientation. For very specific conditions and geometries, some experi-ments have been performed where adjacent tube wastage occurred very rapidly in a localized area. Relating these specific conditions to CRBRP, adjacent tube wastage could occur which would result in tube failure in a very short period of time, less than one minute. However, it is not likely that leaks would be optimized as to leak geometry, location and orientation, as those utilized for the experiments. In the event that this did occur, the steam generator rupture discs provide necessary protection. 13 The second class of damage is self-wastage around the leak site. Some experiments have noted that some very small leaks have experienced a sudden enlargement after a period o f relatively steady operation as reperted in Reference 1. The effect of this type of characteristic has been studied by the GE/ANL Steam Generator Systems Development Program and 13 46 is reported in References 3 and 4. Design Requirements The desian requirements for the Steam Generator Leak Detectors 13 have been selected as described below. SGS Leak Detection Requirements 47l Hydrogen Detectors Oxygen Detectors 24 ppb 50 l Sensitivity 6 ppb 0.04-2 ppm 0.1-10 ppm ! 23 Range Response Time $30 sec. <30 sec. 47 Amend 50 7.5-28 June 1979 268 160
Instrument Sensitivity e The wastage rate studies for jet leaks show that leaks below 10-4 lb/sec persist without major damage for more than one loop transit time.6 Thelooptransittimecanbgcalculated from a 13.49 x 10 lbs/hr flow rate and 4 x 10 lbs sodium 47 inventory in the IHTS loop; the hydrogen generated from the quantity of H2 0 leaked in one transit time divided by the total sodium inventory yields an increase of 6.3 ppb in the 50 471 concentration of hydrogen, thus a 6 ppb sensitivity for the hydrogen detectors. e A resolution cf 3 ppb change in the hydrogen background con-5(j centration ranging f om 60 - 200 ppb (i.e., a change of 3-4%) under steady-state SG operation is a design goal for the leak detector. e The oxygen detector is as sensitive as the hydrogen detector. Taking into account an oxygen background concentration of 47 33 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 Amend. 50 O 7,5-29 qb'3 L lDi June 1979
13 Operation t Requiren,ents e In order to effect an orderly plant shutdown which minimize plant unavailability, the following operator actions are required. Ald'E L ea k_ Si z_e_.( 1 b/ se c ) Operator Ac tion.
-5 Low <2 x 10 Confirm leak Monitor leak data
-5 Confirm leak Intermediate 2 x 10 to 5 x 10~
Initiate orderly loop Shutdown High >5 x 10 -3 Confirm leak Initiate rapid 47 module blowdown 9 For leakages greater than about 0.1 lb/sec of water, the pressure buildup in the system will occur rapidly, causing the Sodium-Water Reaction Pressure Relief System to be activated (See Section 7.5.6). 7.5.6 Sodium-Water _ Reaction Pressure Relief System (SWRPRS) Instrumen_t_a_ tion and Controls 7.5.6.1 DesiE egiy D ti on 7.5.6.1.1 Function The Sodium-Water Reaction Pressure Relief System Instrumentation and Control equipment detects the inception of a large or intermediate sodium-to-water leak in any of the steam generator modules (see Section 5.5.2.6). The following automatic actions in the affected loop are initiated by a large sodium-to-water leak which bursts the steam generator 43 module main rupture discs:
- a. Initiation of reactor and main sodium pump trip.
- b. Trip of recirculating water pumps. , ,,
- c. Isolaticn of steam generaur modules water side. "
Amend. 47 7.5-30
9 7.5.7 Centainment Hydrogen Monitoring The objective of Containment Hydrogen Monitoring is to provide indication in the Control Room of the hydrogen concentration in the upper levels of containment. 7.5.7.1 Design Description The hydrogen instrumentation consists of three self-contair ed electro-mechanical cells located near the top of the RCB with signal lines running outside of containment to preamplifiers. From there, the amplified signal go to the display and alarm panel in the R eactor Cortrol Room so that continuous readout will be provided to the plant operator. 7.5.8 Containment Vessel Temperature Monitorina The objective of Containment Vessel Temperat 'e Monitoring is to provide indication in the Control Room of the cor nment vessel temperature. 7.5.8.1 Design Description The temperature instrumentation consists of thermocouples provided for the outside of steel containment vessel. Signals will be provided to the display and alarm panel in the R eactor Control Room so that continuous readout will be provided to the plant operator. 7.5.9 Containment Pressure Monitoring The objective of the Containment Pressure Monitoring System is to provide indication in the Control Room of the pressure inside the containment above the oeprating floor. 7.5.9.1 Design Description The pressure instrumentation consists of a pressure detector inside the containment vessel. Signals will be provided to the display 25 and e'irm panel in the Control Room so that continuous readout will be I prov a d to the plant operator. 7.5.10 Post Accident Monitoring Table 7.5-4 provides a listing of those parameters which are monitored to assure the plant is maintained in a safe shutdown status. Equipment to condition, display, and record the instrument 50 signals is provided in the Control Room. The instruments which serve the Post Accident Monitoring function are included in those discussed 49 in Sections 7.4.1, 7.5.2, 7.5.3, 7.5.8, 7.5.9, and 7.6.3. The functions of these instruments correspondinq to the parameter of f able 7 5-4 50 are described in the following paragraphs. m , _ Amend. 50 7.5-33b /bb }bs June 1979
The reactor sodium level is monitored to enable the operator to determine whether manual action is necessary to mitigate condition 3 which nay cause low reactor Vessel sodium levels. The operator may also use the reactor sodium level to deter mine <:onditions in the primary sodium systems such as the volume of primary coolant leakage. The IHX (Intennediate Heat Exchanger) inlet and outlets temperatures are monitored to verify the heat transfer from P4TS to the IHTS. These monitors allow the operator to take manual actions r"Jcessary to assura decay r. eat removal from the reactor is maintaire?d within design limits. The DHRS cold leg temperature is monitored to assess the systems' decay heat removal perfonnance so the operator may take manual actions necessary to achieve or maintain core temperature at a safe level. Reactor Containment Building pressure and tenperature monitors are provided to follow pressure and temnerature changes due to an accident, and previde confidence that the accident consequences are within the capability of the Containnent Vessel. Also, these monitors can be used to detect conditions beyond the design basis as discussed in Reference 0 of Sectior 1.6. The PWST (Protected Water Storage Tank) water level is monitored so the operator is assurcJ of the adequacy of that water supply to remove reactor decay heat for an extended period of time. In adaition, the level monitor provides indication of the long tenn need ,, refill the PUST or draw auxiliary feedwater from an alternate source. The Auxiliary r eedwater Flow is monitored to inform the operator of nonnal, abnormal or inadequate flow. He can take manual actions to provide adequate flow and thas maintain adequate decay heat removal through the steam genera'or systen. Steam drum level and pressure are monitored to identify e accident and to allow the operator to take manual actions to initiate and control systems required to achieve and maintain dec.ay heat removal via the S team Generator System. LVST sodium hot leg temperature is monitored tr. enable the operator to make man a l actions during normal, transient, and accident conditions which are necessary to prevent or mitigate the exceeding of 50 Ex-Vessel Storage Tank and stored fuel assembly design limits. Amend. 50 7.5-33c 268 !2
9 References to Section 7 5
- 1. Ford, J. A., "A Recent Evaluation of Foreign and Domestic Wastage Data from Sodium Water Reaction Investigation", APDA CTS-73-05, January, 1973.
- 2. Morejon, J.A., " Sodium-to-Gas Leak Cetection Mockup Tests",
N707-TR-520-C04, September 17, 197 L (Atomics International)
- 3. Greene, D. A., J. A. Gudahl and J. C. Hunsicker, " Experimental Investigation of Steam Generator Materials by Sodium-Water Reactions, Volume 1, GEAP-14094, January 1976.
- 4. Gudahl, J. A. and P. M. Magee, "Microleak Wastage Test Results",
46 GEFR-00352, March 1978. Amend. 50 7.5-33d June 1979 I bf3 !bb
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TABLE 7.5-4 POST ACCIDENT MONITORING Parameter Sensor Location RCB
- 1) Reactor Sodium Level IHX Inlet Temperature RCB 2)
- 3) IHX Outlet Temperature RCB
- 4) DHRS Cold Leg Temperature RCB RCB
- 5) RCB Pressure
- 6) RCB Temperature RCB SGB
- 7) PWST Level
- 8) Auxiliary Feedwater Flow SGB SGB
- 9) Steam Drum Level SGB
- 10) Steam Drum Pressure 50l 11) Deleted 49 12) EVST Outlet Piping Temperature RSB 268 160 Amend. 50
$ June 1979
Rotating Plug drive system /IVTM grapple position Rotating plug drive system /IVTM hold down sleeve Rotating plug drive system /EVTM position EVST drive motors /EVTM grapple position y 7.6.2.2 Desi_gn Analysis Postulated Reactor Refueling System (RRS) accidents with potentially severe consequences were analyzed in detail to determine requirements for safety interlocks. The techniques employed included safety assurance diagrams, fault trees, mechanical and thermal analyses, and radiological release calcu-lations. fione of the analysis results showed off-site doses exceeding those presented in Section 15.5 or 15.7. The off-site doses in Section 15.5 and 15.7 resulting from postulated RRS accidents are all well below the 10 CFR 100 guideline exposures without taking credit for interlocks. It was there-44 fore concluded that the RRS does not require safety interlocks. 7.6.3 Direct Heat Removal Service (DHRS) Instrumentation and Control System 7.6.3.1 Design Description 7.6.3.1.1 Function The DHRS (fluid system and mechanical components as described in Section 5.6, and electrical components as described below) provides a supple-mentary means of removing long term decay heat for tha remote case in which none of the steam generator decay heat removal paths are available. The DHRS Instrumentation and Control System is provided to permit the monitoring of system conditions and to provide alarm indication of off-normal conditions. These are the same instrumentation and controls that are provided for EVST cooling (Section 9.1.3.1.5) and the reactor primary sodium overflow circuits (Section 9.3.2.5) with the addition of a few temperature monitoring instruments located on the flak lines connecting the overflow heat exchanger with the EVST NaK cooling loops (see Figures 9.3-2 and 9.3-3). 7.6.3.1.2 Design Criteria Design criteria that are applicable to DHRS electrical equipment are as follows: A. No single failure of an instrument, interconnectug cable or panel shall prevent a key process variable fromteing monitored. B. DHRS valves shall be manually operated and DHRS electrical equip-ment shall be manually controlled (see 5.6.2) from a panel in the Control Room to provide 1/2 hour start up capability. C. Physical and electrical separation of redundant portions of DHRS (FVS cooling system, primary makeup pumps, instrumentation, and 26 2bb }b9 1 8 7.6-3
controls) shall be provided. D. Electrical power supplied to DHRS electrical equipment shall be independent of off-site power. 46 F. DHRS control instrumentation and DHRS electrical equipment shall function during and af ter an SSE. F. Capability for periodic calibration and testing of DHRS electrical equipment shall be provided. 7.6.3.1.3 Equipment Design As shown on Figure 5.1-7, the DHRS is part of the primary sodium pro-cessing, and the EVS Sodium Processing System. Description of the functioning of these systems for reactor decay heat removal is provided in Sections 9.1.3 and 9.3.2. The P&I diagrams are givan in Figures 9.3-2 and 9.3-3. DHRS electrical equipment meets the design criteria listed in Section 7.6.3.1.2 above in the following manner: A. Control Systems The following DHRS control functions are provided from separate, re-dundant control panels (local and main control room): (1) Remote manual control of voltage to all NaK and sodium pumps. (2) Remote manual control of ABHX dampers and fan speed. (^,) Remote manual override of pump and ABHX interlock circuits. (4) Remote manual control of all valves required to provide DHRS. B. Menitoring Instrumentation Some instrumentation required to monitor the functional performance of the decay heat removal process loops is redundant from the sensor out to and including the readout panel, so that a single failure of an instrunent, inter-connecting cable or panel does not prevent the process loop from being monitored. Ir. those cases where a redundant sensor is not provided, separate indicators on separate panels are provided. Where redundant sensors are not provided, loss of the sensor does not prevent the acquisition of equiva'ent diagnostic information from other sensors on the precess loop. The following EVST cooling and PHRS process variables are monitored with completely redundant instrumentation (sensors, cabling, and panels):
* (1) EVST outlet sodium temperatures SC 46l
- Raquired for post accident monitoring. Amend. 50 26 June 1979 7.6-3a o
,/ h 0 .U
( 2) Overflow vessel sodium level
* ( 3) Reactor overflow sodium inlet temperature
( 4) EVST sodium pump control signal (each pump) (5) EVST NaK pump control signal (each pump) (6) Primary makeup pump control signal (each pump) 46 50 The following EVST cooling and DHRS process variables are moni-tored using a single sensor and redundant cabling and panels. (1) EVST sodium flowrate (each loop) (2) EVST NaK flowrate (each loop) (3) Primary overflow makup sodium flowrate (each pump outlet) 46l (4) EVST airblast heat exchanger fan speed setting (each loop) (5) EVST NaK expansion tank level (each loop) (6) EVST sodium inlet temperature (each loop) (7) EVST sodium and NaK remotely operated valve position indicators (each loop) (8) EVST airblast heat exchanger damper position indicator (each loop) (9) DHRS aeat exchanger bypass valve position indicator 46l (10) DHRS NaK expansion tank level C. Annunciation and Data Handling The following EVST cooling and DHRS process variables are con-nected to the plant annunciator system to alert the plant operator of off-normal conditions: (1) Low sodium and Nar pump gas cooling f!cw rate (each pump) 46 (2) High EVST sodium, EVST NaK and primary makeup pump stator
' temperatures (each loop) 26 7.6-3b Amend. 50
[,$h }{l June 1979
(3) Low EVST sodium, EVST NaK and primary makeup flowrate (each pump loop) (4) High and low EVST sodium inlet temperature (each loop) (5) High and low EVST NaK expansion tank level (eacn loop) (6) Hgih and low EVST sodium level (7) High and low EVST sodium temperature (8) Low sodium valve temperatures (9) High and low DHRS expansion tank level 46 Key process variables that are connected to annunicators are also connected to the plant data handling and display system. D. Other Features Remotely operated valves in EVST cooling and DHRS circuits in-corporate either " fail safe" or " fail in place" features and are provided 46l with direct manual (reach rods on sodium valves) override capability in event of I&C or gas supply failures. Type lE power is supplied to the equipment and instrumentation I required to provide the safety related functions of EVST cooling and DHRS as shown in Figure 7.6-13. This assures independence of off-site power. Functional testing of all portion of DHRS that are not used during the course of normal operations will be tested on an annual basis during reactor refueling. Equipment required to provide power to EVST and DHRS pumps, airblast heat exchangers and the monitoring instrumentation in the control panels shall be designed and tested to Seismic Category I requirements. 7.6.3.1.4 Initiating Circuits Reactor decay heat removal through DHRS is initiated from the Control Room panel as described in Sections 9.1.3 and 9.3.2. 7.6.3.1.5 Bypass and Interlocks When the DHRS is activated, automatic ccr. trol of the EVST air-blast heat exchangers is bypassed so that control of the complete system is manual. All valves in this circuit are also operated on a direct or 26 remote manual basis. The flow in the primary sodium overflow makeup Amend. 46 7.6-3c Aug. 1978 263 172
It is not anticpated that the FHC requires alpha-decontamination after handling each failed fuel assembly since the cell is an alpha-tight facility and can operate with contamination. At regular intervals, swipe samples will be taken at varioun loca-tions in the FHC and particle fallout samples (dishes) will be anal- 3d to monitor the amount of alpha-contamination. The samples will be compared to established contamination levels and will provide warning of any contamination buildup. One method of minimizing the radioactive spread of alpha contam-49l ination through the EVTM and SFSC to other f acilities will be by contin-uously filtering the FHC argon atmosphere. Alpha emitting particles suspended in the argon atmosphere will be removed through a HEPA filter bank installed upstream of the FHC argon gas circulation blowers. If excessive contamination buildup on the FHC liner is indicated, cleaning measures will be initiated. It is, therefore, concluded that:
- 1) Maintenance of the FHC equipment is consistent with the current practice of maintaining equipment in hot cells.
- 2) Alpha-decontamination of the FHC atmosphere wili be carried out on a continuous basis. Removal of alpha-contaminateu de-posits or debris will occur af ter a predetermined contamination 25 threshold has been exceeded.
This section will cover the safety aspects of the FHC containment, 44l and the spent fuel transfer station. Spent fuel decay heat removal by the spent fuel transfer station and by the gas cooling grapple are discussed in Section 9.1.3. 9.1.2.2.1 Design Bases Adequate shielding is provided in the FHC containment structure for j radiation protection outside the cell, to meet the requirements and radiation 491 zone criteria of Section 12.1. Radioactive releases and contamination from spent fuel assemblies that are being prepared for shipment in the FHC are contained within the FHC by proper sealing or closure welding of penetrations. Radioactive leakage and diffusion through seals, in the unlikely event of release of the entire ficsion gas inventory of a fuel assembly, are limited to well below the 44 criteria of 10CFR100. Criticality of fuel assemblics temporarily stored in the spent fuel transfer station is prevented bj physi'al separation and by limitinq their 44 number. 49 an fCO f[l Arend. 49 9.1-13
49 The spent fuel transfer station design considers all normal loadings in 44 combination with the loads from an 55E in nidintaining the necessary physical separation. The FHC roof closure is designed to absorb the load of the heaviest equipment handled by the RSB bridge crane over the FHC: (a) for the main hook, lowered at the maxiinum crane speed (5fpm), and (b) for the auxiliary hook, accidentally dropping Pom the maximum handling height to which it is raised, onto the center of the roof closure without affecting the integrity of the fuel separation lattice. The FHC is located such that heavy equipment not belonging to the fuel hendling and storage system is not carried over it by the RSB bridge crane. 20 The spent fuel transfer station within the FHC is designed so that 44 moventat of the lazy susan will not occur while a CCP is being inserted or wi thd rawn. This design condition prevents mechanical damage to the CCP or its contents. Monitoring instrumentation will be provided for the FHC for condi-tions that might result in a loss of the capability to remove decay heat, and to detect excessive radiation levels. 9.1.2.2.2 Design Description The top of the FHC is located at the operating floor of the RSB, 44,l as shown in Figure 9.1-7. Sufficient shielding is provided so that the r radiation level above the FHC does not exceed the radiation Zone I criteria, 49 sec Section 12.1. This shielding is provided by the cell's roof closure assembly, a load-bearing structure which is part of the RSS operating floor. The FHC side wall facing the operating gallery is shielded by high density concrete tc protect the operating gallery against radiation dose 20 l rates exceeding the radiation Zone I criteria. The other walls and the 49 floor are shielded by conventional concrete to protect the neighboring vaults and the spent fuel shipping cask handling corridor against radia-ation, see Section 12.1. All windows, and port penetrations through the roof, walls, and floor are stepped to limit radiation streaming in the gaps. The main source of radiation in the FHC is spent fuel assemblies in the spent fuel transfer station. g The liner seams on the cell interior walls, roof and floor, and welded penetrations through the FHC walls, roof, and floor are alpha-tight 44lweldedandinspected. Fuel transfer ports, the maintenance and service station port, window seals, and slave manipulator penetrations each have double elastomeric seals buffered with pressurized argon gas. Sealed cover glasses are provided on the interior side of the window penetrations. The spent fuel transfer station within the FHC is shown in Fig-44 ure 9.1-8. A maxin.um of 3 spent fuel asser alies in CCP's can be stored in,' s,i
'y' Amend. 50 June 1979 9.1-14
where contamination can be contained and where modification to shop tools can be made. 3E The welding cart provided by this group is used primarily in the Reactor Containment Buildino. The welding cart is sized such that it can be moved to the job within the Reactor Containment Building. After being lowered through equipment hatches, or transported in the RCB elevators, 35l 4g it can be moved around on various floors of the building. 9.2.1.2.2 Handling Equipment Most of the handling equipment provided is used within the Reactor Containment Building to lif t and transport components. Some handling equipment is also provided for use in the Steam Generator Building. All the principal overhead handling systems for servicing safety-related equipment are designated as Seismic Category I Cranes. All the structura', automatic and manual mechanical and electrical components of the Category I cranes will be designed so that no single failure or malfunction will result in dropping or losing control of the heaviest loads to be handled. Design, fabrication, installation, inspection, testing and operation of the Category I cranes will be defined in the crane specifi-cations to ensure capability of the crane to retain the maximum design load during a Safe Shutdown Earthquake, although the crane may not be operable after the seismic event. For those cranes which may be used for construction prior to start of plant operation, a separate performance specification will be prepared and at the end of the construction period, the crane handling system will be modified to conform to the performance requirements of permanent plant service. Af ter construction use, the crane will be thoroughly inspected using nondestructive examinations and will be 25 performance tested. p~ Lif ting fixtures are provided for the purpose of handling PHTS components and large maintenance equipment. These fixtures interface with the component through the use of adapters. Special lif ting fixtures, for the purpose of installing or removing other components, are the responsi-bility of the component system, and are not provided by the Nuclear Island General Purpose Maintenance Equipment System. Patterned bag containers, for handling sodium-service and/or radioactive service components, are provided. Patters are provided for these baos, and each bag has an er.velope interface with the component to be handled. 2bb l b Amend. 50 9.2-3 June 1979
Transport dollies are provided to move components between the Reactor Containment Building and the Reactor Service Building. A small dolly is provided to transport components which will pass through the service airlock. A large component transporter is provided to utilize the Refueling System gantry rails, and is positioned on the gant.ry rails when components too large for the airlock must be moved between the two buildings. The large transporter is removed from the gantry rails and stored elsewhere when not in use. The large component floor valve is provided t i maintain contain ment of the inert atmosphere and radioactive gases during the performance 43 of general purpose maintenance in the PHTS primary pump, EVST cold trap and the PHTS cold trap access ports. Interface with the building floor at each of the access ports is provided by means of an adapter. A separate adapter will be provided 43 for the PHTS primary pump and a common adapter for both traps. 1 O ocq 1U t:. : .' \ Amend. 43 Jan. 1978 9.2-3a
9.2.2.2 Design Description The Sodium Removal and Decontamination (SR&D) System is composed of three subsystems located as shown on Figure 9.2-1. These are:
- 1) Primary Sodium and Decontamination (PSR&D) System - for 49l removing sodium from fuel handling. primary reactor vessel internals, and reactor enclosure components, and decontam-43l 35l inating the Primary System pump of fission and corrosion product plate-out.
49l 2) Intermediate Sodium Removal (ISR) System - for removing sodium from the modular steam generator and Intermediate System pump.
- 3) Small Component AJtoClave (SCA) - for removing sodium and NaK 49 from Auxiliary Liquid Metal and Fuel Handling System com-ponents.
The Primary Sodium Removal and Decontamination System, shown in Figure 9.2.2, is located in an 83 f t. deep cell that is below the operating floor in the northwest part of the Reactor Containment Building (RCB). This system is remotely monitored and controlled from a control panel on the RCB operating floor. The Intermediate Sodium Removal System, shown in Figure 9.2-4, is located in the maintenance bay of the Steam Generator Building. Most of this system is manually controlled. The critical process parameters are 49l monitored and controlled at the panel board located near the process equip-ment. This facility will be designcd and built af ter reactor startup. The small component autoclave is a horizontally positioned, 4-ft. diameter by 5-ft long vessel with cover. located just outside of the decontamination facility in the Reactor Service Building. The cleaning vessels are designed to provide the process con-ditions required to remove sodium from sodium-coated components by the moist nitrogen process. These conditions include inerting, preheating, 35l reaction with water vapor diluted with nitrogen, water rinsing, and drying. A decontar ' nation phase is proviued in the Primary Sodium Removal and Decontamination System. I4 In the PSR&D System, the moist nitrogen gas and all liquid can be recirculated to enhance procegs control. A vacuum of s1 in. Hg 50 (absolute) and/or hot gas at 180 F is used for drying. The waste water rinse liquids are sent to the Radioactive Liquid Waste System, and the waste gases are sent to the Heating and Ventilation System. The reprocessed waste rinse water (purified by the Radioactive Waste System) is reused in 49 the cleaning operation. 9.2-7 Amend. 50 June 1979 268 17'
In the ISR System, the contaminated water (tritium only) is transferred to the waste water system for disposal, and the supp?y 49 ; water is obtained from the treated water system. O 9.2-7a Amend. 50 June 1979
'/b.O ,
Additional information on controlline and disposing of hydrogen follows. Hydrcqen concentration in the Primary Sodium demoval and Decontamination (PSR&D) System is controlled by: Regulating water vapor concentration in the water vapor nitrogen (WVN) mix'ure introduced to the cleaning vessel. Regulating nitrogen purge rate. In the normal sodium removal process, (1) the vessel is purged with N2 urtil the 0 level is below 10, and (2) saturated steam at 15 psig is-7 metered into the nitrogen purging stream. The steam and N7 gas are mixed in a mixing tee and then introduced into the cleaning vessel where it reacts with the Na on the component. Water vapor concentration is controlled to be between 5-150 by means of remotely operated flow control valves. The lower limit is provided so that Na will effectively react with H70 vapor, the upper limit of 15% is provided to preclude condensation of water vapor in the WVN mixture at 1500F, which is the component temperature. The normal operating condition maintains the hydrogen, which is a reaction product of Na and H 0 in the vessel, to below 1 v/o. WHen the H 7 concentration incre< ses indicating a faster Na-water reaction, the steah supply introduced to the mixing tee is throttled down and the nitrogen purge. rate is increased. Both of the flowrates are controlled remotely from the control panel. The H concentration is 2 monitored by a H7 meter installed in the vent line for the effluent gases. Should the H7 concentration exceed 4 v/o, the meter sensor will set off a high H7 al3rm and at the same time activate a control interlock which automatically shuts off the ? team supply flow control valve while the nitrogen purge continues. The steam flow may be resumed when the excess hydrogen is reduced to below 1 v/o by continuing nitragen purging. In addition to the nitrogen purge feed and bleeding of the gas effluent, the gas mixture in the cleaning vessel is circulatea by a blower through the cooler at the rate of 2,000 cfm during the WVN Na removal cycle. Although the primary purpose of the circulation is to remove the heat of reaction, it also promotes uniformity of the mixture of gases in the cleaning vessel. The pressure in the vessel is kept at 5 psig during operation by means of a back pressure valve. The positive operating pressure is provided to prevent air infiltration in:.o the PSR&D system-Since the maximum hydrogen concentration in the vessel is 4 v/o, any leak of the LCCV gas mixture will ' diluted by air to below the hydrogen flammability limit in air which is v/o. The Primary Sodium Remo 1 and Decontcmination System is desi nonnuclear safety in accordance with PSAR Section 3.2.2.4; namely, (1) gnated as failure of system will not result in exposures at the site boundary or beyond in excess of 0.5 rem whole body or its equivalent, and (2) failure of the system will not damage any in-service safety class components or the plant shutdown capability. P_5 9.2-8 26@ \[9 Amend. 25 Aug. 1976
The equipment of the Primary Sodium Storage and Processing System is mounted in cells that have an inert atmosphere, but are accessible af ter system or component shutdown for inspection af ter de-inerting cells and radioactivity decay. The equipment is mounted or supported so that inspection of vessels, pumps and piping can be accomplished. 9.3.2.5 Instrumentation Requirements Instrumentation and controls (I&C) are provided for operation, performance evaluation and diagnosis of the Primary Na Storage and Processing System. These functions are required for off-normal as well as for the full range af normal operation. Details of the I&C for the sub-system are shown in the piping and instrumentation diagram, Figure 9.3-2.
- 26. DHRS instrumentation is discussed in Section 5.6.2.1.6. The following I&C is required to ensure safe operation of and to prevent extensive damage to the Primary Na Storage and Processing System.
Temperatures at the irlet and outlet of all heat source and sink components, in conjunction with loop flow measurements are provided for all systems to monitor their status. Critical temperatures and flows are alarmed to alert the operator to of f-nonval operations. All EM pumps 461 are provided with winding temperature neasurements and winding coolant low flow indication. These measurements are alarmed for off-normal conditions and interlocket +o automatically shutdown the pump to prevent damaging it. Storage tanks are provided with level measurements, wnich 3re alarmed for abnormal low and/or high level. This infonnation, in con-junction with leak detection data, is required to diacnose external liquid metal leaks. The operator is alerted to Nak to sodium leakage by NaK expansion tank high-low pressure alarms. Differential pressure sensors and 50l 46 I flow meters are provided to alert the operator to possible plugging of the cold traps or insufficient cold trap flow. All the bellows seal valves are provided with leak detectors (Section 7.5.5.1). All valves are provided with position indicators. The stem portion of the sodium valve is monitored and alarmed for low temperature to ensure free operation and 46l protect the valve sodium seal from damage. To provide for continued operation and prevent possible system damage resulting from control system failures, hand controllers are provided for all controllers. The 50l hand controller allows the operator to manually operate the sys.em while the defect is repaired. Redundant temperature sensors are provided for each primary cold trap. High temperatur: conditions in either cold trap are indicated and alarmed in the Control Room to ensure tnat the cold trap is isolated prior to plant cooldown to refueling temperature. Thus plugging from high 36 impur;ty content in the PHIS is precluded. Amend. 50 June 1979 9.3-9 y(. u ) m ivV
O 9.3.3 EVS Sodium Processing 9.3.3.1 Design Basis See Sec tion 9.1.3.1.1. 9.3.3.2 Design Description See Section 9.1.3.1.2. O Amend. 46 Aug. 1978 9.3-9a O w, w
The equipment is mounted in cells that have an inert atmosphere, but are accessible af ter system or component shutdown for inspection af ter de-inerting. The equipment is mounted or supported so that inspection of vessels, pumps and piping for possible deterioration of NaK containment integrity cai. be accomplished. 9.3.4.5 Instrumentation Requirements Instrumentation and controls (IFC) are provided for operation, performance evaluation and diagnosis of the NaK cold Trap Cooling System. These fur.ctions are required for off-normal as well as for the full range of normal operation. Details of the ISC for the subsystems are shown in the piping and instrumentation diagram, Figure 9.3-4. The following I&C is required to ensure safe operation of, and to prevent extensive damage to the NaK Cold Trap Cooling System. Temperatures at the inlet and outlet of all heat source and sink components, in conjunction with loop flow measurements are provided ror all systems to monitor their status. Critical temperatures and fl as are alarmed to alert the operator to off-normal operations. The EM pump is 46l provided with winding temperature measurements and winding coolant low flow indication. These measurements are alarmed for off-normal conditions, anc interlocked to automatically shutcown the pump to prevent danaging it. 46l The storage tank is provided with a level measurement which is alarmed for abnornal low and/or high level, This information, in con-15]metal junction leaks.withTheleak detection operator data, is alerted is required to NaK to sodiumtoordiagnose Dowthem Jexternal liquid 46] to Nak leakage by NaK storage tank high-low pressure alarms, in conjunction with the level measurement mentioned previously (see also Section 9.1.3). All the bellows seals valves are provided with laak detectors. All valves are provided with position indicators. To provide for continued operation and pr' event possibie system damage resulting from control system failures, hand controllers are provided for all controllers. The hand controller ellows the operator to manually operate the systc.o while the defect is repaired. 9.3.5 Intermediate Na Process ng System i 9.3.5.1 Design Basis The system provides the capability to limit the oxygen and hydro-gen concentruion of IHTS sodium to 2.0 ppm and O. 2 ppm, respectively. 50l The system, the tritium e rkingo'inIHTS content conjunction sodium with the primary to levels coldwith consistent traps, limits plant radio-logical release criteria. Amend. 59 June 1979 9.3-12 k,bh no lOc
The system also provides the capability to (1) fill the IHTS loops f rom the sodium Jump tanks (these tanks are part of the Steam Generator Sys-tem), (2) purify sodium in the dump tanks, independently of the IHTS loops,
~
and (3) permit transfer of sodium from one dump tank to another. 9.3.5.2 Desic. cription The Intermediate Sodium Processing System provides purification of the sodium in ea:h of the three IHTS loops. The System does not provide for storage of the IHTS sodium. This capability is provided by the sodium dump tanks, which are part of the Steam Generator System. The Intermediate Sodium Processing System does provide the capability of transferring sodium into the loops from the dump tanks. The same piping network allows the filling of each dump tank with fresh sodium from tank cars or drums at the sodium receiving station. Sodium rerroval from the tanks into tank cars can be accomplished through the same fill lines. The system includes the following components:
- a. Inter.!ediate Sodium Cold Trap Pumps
- b. Intermediate Sodium Cold Traps
- c. Interconnecting Piping and Valves Refer to Figure 9.3-5 for the P&ID and Figures 1.2-8 and 1.2-22 for layout and arrangement.
Each of the three IHTS loops is provided with a separate purifi-cation system consisting of a pump, two cold traps, and the necessary valves and piping. A single trap per IHTS loop is sufficient to remove antici-pated oxygen and hydrogen inleakage and to limit these impurities to a maximum of 2 and 0.2 ppm, respectively. In addition, the intermediate 50 cold traps maintain the tritium level in the intermediate sodium at 0.012 aciT/gm Na by effectively trapping about 981 of the tritium which enters the system by diffusion through the IHX. Most of the remaining tritium, 46 50l0.016 Ci/ day, diffuses through the steam generators and enters the water systcm. Cold trap flow is 160 gpm at normal system operating temperatures. The Intermediate Sodium Processing System is also connected to the sodium dump tanks such that the sodiun in the dump tank may be processed by the system. The intermediate sodium cold trap pumps are used to pump the sodium from the dump tanks into the loops with a small,122 psig, cover 46 gas pressure being maintained on the dump tanks. Sodium can also be transferred from one dump tank to another by gas pressure. 2bb Amend. 50 June 1979 9.3-13
The use rate of argon by these services is variable and is dependent on operator options. Under start-up conditions, the flow will be maxin,um, and a minimum supply capabi!ity of 95,000 scfd of argon is to be provided. Argon is to be used for all services involving sodium wetted components, such as fuel handling sampling, and maintenance services. This gas also is ultimately exhausted through CAPS to the atmosphere. Argon is also to be supplied for purging and inerting IHTS components and for sodium water reaction control purposes. 9.5.1.2 Design Description The argon distr bution subsystem is composed of liquid argon Dewars with vaporizers, gaseous argon bottles, piping, valves, vapor traps, filters, vessels, relief systems, freeze vents, and oil traps as necessary to distribute the argon to meet the requirements described in Section 9.5.1.1. 9.5.1.2.1 Recycle Argon Distribution Argon f rom the primary recycle cover gas storage vessels in the RCB is reduced in pressure to supply cover gas to the reactor vessel, primary sodium overflow vessel, and primary pumps cover gas spaces, which are all interconnected by a pressure equalization line. This cover gas system is maintained at a pressure of 6 in. w.g. by a feed and bleed control system. There is a continuous transfer of argon cover gas from the reactor and the primary pumps via the equalization line to the primary sodium overflow vessel and then through a 5-scfm vapor trap that removes sodium vapor. This vapor trap consists of a vapor condenser and two parallel aerosol filters (one redundant). The gas flows back to RAPS for processing before recycling. A 1-scfm sample of cover gas is taken from the equalization line and is passed through a 1-scfm sodium vapor 50l trap to the Failed Fuel Monitoring System. This gas and the cover-gas bleed from the primary pumps are also returned to RAPS. 9.5.1.2.2 F_resh Argon Supply at RSB Argon for services in the Reactor Service Building (RSB), the Reactor Containment Building (RCB), and the Intermediate Bay (IB) is stored as liquid in two Dewars, located on the RSB pad. These Dewars have a capacity of 1500 gal. each and are equipped with fill and vent lines. Normally, only one of the Dewars is in operation. When it is nearly empty, a low-liquid-level instrumentation signal operates auto-matic controls that shutoff that Dewar and open the other Dewar to the supplj header. A control override allows drawing on both Dewars simul-48 taneously. 9.5-3 Amend. 50 June 1979 740
/uO inf l0i
Two ambient-air vaporizers on each Dewar can evaporate the liquid argon at a nominal maximun, gas flow rate of 2000 scfh each, at 175 psig. With both Dewars on-line, therefore, approximately 8000 scfh of argon gas at 175 psig can be delivered. The argon from the Dewars passes through a filter and is then divided into three main headers that supply argon to the RCB, RSB, and other ex-containment components. 9.5.1.2.3 Fresh Argon: RCB Distribution The RCB neader enters the building with isolation valves on each side of the penecration. This header supplies argon to the primary sodium storage vessel, with a feed and bleed system at a normal pressure of 1 psig, and to the recycle argon storage sessels. The RCB header also supplies argon to the primary sodium plugging temperature indicator, the primary sodium sampling package, the floor / wall service stations, the reactor head inflatable seals, and the IVTM storage facility. The RCB header also supplies argon to the primary sodium line freeze vents, which are furnished argon during startup, maintenance, and sodium drain and fill at a nominal pressure of 5 psig; the pressure can be increased, if needed, to 50 psig. This header also supplies cover gas argon for the NaK system and the make-up pump drain vessel. 9.5.1.2.4 Fresh Argon: RSB Ex-Containment Distribution The RSB ex-containment header supplies make up argon to the ex-containment primary sodium storage vessels in the Intermediate Bay. The normal pressure in the storage essels is 1 psig, but this can be increased to 50 psig during tank drain. These vessels can be vented either through a vapor trap and a pressure control valve to the Cell Atmosphere Processing System (CAPS) or to a vacuum station and then to the CAPS. 9.5.1.2.5 Fresh Argon: RSB Distribution The RSB header supplies argon at the required pressures to the gas chromatograph, the fission gas monitor module, and the gas sampling trap. A branch line provides argon purge to the RAPS cold box. The RSB header supplies argon through regulators to the Aux-iliary Liquid Metal System EVS Na and NaK components and to the Impurity Monitoring and Analysis System EVS sodium sampling package. The sodium lines have freeze vents that are furnished with argon during startup, maintenance, and sodium drain and fill operations at a nominal pressure 48 of 5 psig. This pressure can be increased to 50 psig. 9.6_4 Amend. 4B Feb. 1979 2() lbb
9.5.2.2.1 Nitrogen Supply at RSB The RSB and RCB nitrogen supply is stored as liquid nitrogen in two Dewars, each with 6000 gal. capacity, on the RSB pad. An ambient air vaporizer on each Dewar can evaporate the liquid nitrogen at a nominal flow rate of 15,000 scfh. Normal nitrogen usage is supplied from one Dewar, with a level sensor automatically switching tanks upon depletion of a pre-set level. A control override, however, allows the option of simultaneously supplying nitrogen from both tanks so that doubling the flow rate to meet peak demands is possible. 9.5.2.2.2 Nitrogen: RCB Distribution The header feeding the RCB contains one isolation valve on each side of the containment penetration, providing automatic shutoff capability on either side in the event of nitrogen pressure loss. The header inside containment branches of f into (1) a low pressure header feeding all of the normally inerted cells and pipeways within contain-ment, (2) a high pressure line for actuation of valves in cells that are normally inerted, (3) a line to the CRDM assembly recirculation cooling system, and (4) a line to provide sparging gas to the sodium component cleaning operation. Cells and pipeways containing sodium components in the RCB are normally inerted with nitrogen atmosphere, as is the CRDM cooling system. Each inerted cell or group of cells has inlet and outlet control valves that maintain preset cell pressures, in addition to having automatic cell purging for maintaining required oxygen or water-vapor levels. Purge flow is automatically activated by a cell atmosphere sampling and analysis unit that periodically monitors the 07 and H2 O levels in each cell atmosphere. Radioactivity is also monitored but does not activate purging. The inerting system for RCB and RSB celi (except FHC) is designed for normally controlling the oxygen concentiation within the cells to a maximum of 2 vol %. The design base for the cell gas inerting system is a net inward leakage of air of 1% of the cell volume per day. When the cell is inerted to 2% oxygen, which amounts to 10% of the oxygen content of the inleaking air, the water vapor content in the cell will also be 10% of that in the in-leaking air. The Heating and Ventilating System norgally controls the humidity of the air inBecause the building to 40% the cells are R.H. at 75 F (water partial pressure, 0.12 atm). to be steel-lined, dehydration of the concrete will not contribute directly to the water content of the cell gas, so that the normal partial pressure of water vapor is 0.0012 atm, or 1200 vppm water vapor. During initial warm-up and prior to sodium loading, should the water vapor content of the cell atmosphere (which can be air) exceed the normal maximum value, this water will be removed first by cell purging with air, and then, as the Recirculating Gas Cooling System (RGCS) goes into operation, by condensation on the cooling coils. At steady-state, this ut)it will limit the water vapor content of the cell atmosphere to 98 9.5-7 l } f', Amend. 48 268 Feb. 1979
about 10,000 vppm. Reduction from this value to the 1000 vppm limit will be done by purging with nitrogen. Nitrogen for service maintenance operations is available at service stations located within the RCB. 9.5.2.2.3 Nitrogen: RCB Auxiliary Supply An auxiliary supply of nitrogen gas is stored in high pressure standard cylinders located within a cell in the tornado-hardened RCB. This nitrogen is used to ensure the uninterrupted operability of certain 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 re-mainder of the failed supply circuit. 9.5.2.2.4 Nitrogen: RSB Distribution lhe 150 psig RSB header branches off into several lower pressure headers that service the needs of other systems as well as those of the RAPS and CAPS subsystems within the RSB. RSB cells and pipeways containini sodium components are in-erted with nitrogen during normal operatioi. The cell pressures are maintained by a feed and bleed arrangement and a purge function controls impurity levels. (See 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 Heating, Ventilating and Air Conditioning 50 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 nrecess. 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. Nitrogen gas is provided as a cover gas for the Dowthem tanks used in the chilled .2ter system. 9.5.2.2.5 Nitrogen: RSB Auxiliary Supply An auxiliary supply of nitrogen gas, stored in high-pressure standard cylinders located within a cell in the tornado-hardened RSB, Amend. 50 9.5-8 June 1979 O i\ p> '> ( s Q>
TABLE 9.6-4 (cont'd). LOCATION SAFETY SEISMIC PRIMARY EQUIPMENT TITLE BLDG. ELEV. CLASS CLASS PARAMETER (CFM) RCB Inerted Cells Booster Fan RCB 752'-0" 200 RCB Inerted Cells Booster Fan RCB 733'-0" 200 50 RCB Portable Filter Fan RCB - 1,000 RCB Supply Isolation Valve RCB 816'-0" SC-2 I 14,000 RCB Exhaust Isolation Valve RCB 816'-0" SC-2 I 14,000 RCB Supply Isolation Vahe ANNULUS 8 '. 6 ' - 0 " SC-2 I 14,000 RCB Exhaust Isolation Vahe ANNULUS 816'-0" SC-2 I 14,000 RCB Annulus Air Cooling Fan RSB 816'-0" SC-3 I 133,000 RCB Annulus Air Cooling Fan RSB 816'-0" SC-3 I 133,000
?
RCB Annulus Air Cooling Fan RSB 828'-9" SC-3 I 133,000 [ 828'-9" 133,000 $ RCB Annulus Air Cooling Fan RSB SC-3 I 133,000 RCB Annulus Air Cooling Fan RSB 841'-6" SC-3 I f' RCB Annulus Air Cooling Fan RSB 841'-6" SC-3 I 133,000 755'-0" SC-3 I 18,000 [$ RSB Cleanup Filter Unit Fan RSB 18,000 c: ~ RSB Cleanup Filter Unit Fan RSB 816'-0" SC-3 I 755'-0" SC-3 I 18,000 RSB Cleanup Filter Unit RSB RSB Cleanup Filter Unit RSB 816'-0" SC-3 I 18,000 846'-0" SC-3 I 3,000 RSB Annulus Pressure Main- RSB 49 tenance Fan E'E 8e" O
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9.8 IMPURITY 1,0NITORING AND ANALYSIS SYSTEM The Impurity Monitoring and Analysis System consists of the follow-ing subsystems:
- 1) PHTS Sodium Characterization
- 2) Primary Cover Gas Sampling and Monitoring
- 3) EVST Sodium Characterization
- 4) IHTS Sodium Charatterization
- 5) EVST, FHC, and IHTS Cover Gas Sampling
- 6) Analytical Services Laboratory All components and piping in contact with sodium will be constructed of Type 304 austenitic stainless steel. This includes sampling inserts in the multi-purpose sampler in the sodium sampling packaae. Argon cover gas sampling and monitoring piping and components will be constructed of Type 304 stainless steel.
16 9.8.1 Design Basis 9.8.1.1 Sodium Sampling and Monitoring The s dium characterization (samnling ar,d ronitoring) subsystems are 45 l to be provided a continuous flow of sodium from the systems being monito The subsystems will be designed for automatic, on-line impurity monitoring of this sodium, and for sample collection of aliquots of sodium (or equili-bration devices or particulate samples) for laboratory analysis. These se.'ple streams are monitorea, and samples analyzed, to verify that the circu-45 l lating sodium meets the allowable oxygen and hydrogen levels. Analytical results from these subsystems provide information for the operation of the sodium purification units of the Auxiliary Liquid Metal System.
- 9. 8.1. 2 Cover Gas Sampling and Monitoring Primary cover gar samples are provided by the Failed Fuel Monitoring System. These samples will be processed in the analytical services laboratory 50 to determine imp rities. Sample connections are to be provided for sampling EVST, FHC, and IHTS cover gas with sample bottles whose contents are processed in the analytical services laboratory to determine impurities.
Analytical results of these samples will be used by the Inert Gas Receiving and Processing System to establish and maintain the various argon gas purity requirements within specified limits.
.s ,, 4 n-
@~ / b b) I<' Amend. 50 9.8-1 June 1979
' ' ' ^* Y '" "' '*" ' U-45 3 pace is to te provided in the hot laboratory in the plant services building for equipment +.t. perform vot-of-loop analyses of sodium (and equalibration device, and particulate samoles) and cover aas samples.
I 45 27! 9.8.? Design Description 9.8.2.1 PHTS Sodium Characterization Subsystem The PHTS Sodium Characterization Subsystem, shown in Figure 9.8-1 is provided a continuous sample of the primary sodium from the overflow p" vessel or the storage vessel tanks. The piping is designed to accept sodium from the discharge side of the make-up purns. Th f w rate through the subsystem sampling loop is controlled 45 l dt 10 gpm, and returned either to the primary sodium overflow vessel or to the primary sodium storage vessel tant.s, whichever is the source of the comple. A portion of the prinary sodium sample stream can be diverted through 45 l the primary plugging temperature indicator (PPTI) by the proper operation of valves. A plugging temperature indicator (PTI) is a device to determine the saturation temperature of impurities diss lved in the sodium. Tne results, 45 l however, are not specific for a given impurity, tmt do give an indication of impurity levels which are consistent with the saturation temperature. In parallel with the Primary PTI is the primary sodium sampling 45 l package (PSSP), which nrovides samnles for several types of laboratory anal-yses. A portion of the primary sodium sample stream can be diverted through the PSSP by the proper operation of valves. A continuous flow of sodium at a controlled temperature is then provided to either or both of the '_wo sam-45 l pling devices [ multi-purpose samplers (MPS)] in the primary SSP. by the use of inserts, the MPS can:
- 1) Provide a sodium sample for laboratory analyses of sodium impurities
- 2) Be used to expose foils and/or wires of selected materials which will equilibrate wi .h oxygen, hydrogen, or carbon dissolved in the sodium and which can be analyzed in the laboratory
- 3) Filter kr,s n quantities of sodium, to provide reasure of particulate ir: purities in the sodium.
1no
<[ (< n)Ui / ,.
Amend. 45 9.8-2 July 1978
45136l Cseration of the Primary SSP is manual, and requires remote opera-tion. Master-slave manipulators are provided. Typical laboratory information obtained from these samples are:
- 1) Total coolant impurity levels
- 2) 0xygen, hydrogen, and carbon activity
- 3) Tritium level
- 4) Fission product levels
- 5) Corrosion product levels
- 6) Particulate i:. purity levels 45 l36l27l 7) Other impurities as considered necessary 9.8.2.2 Primary Cover Gas Sampling and Monitoring Subsystem Frimary cover gas samples in shielded, gas sample bottles are provided by the Fuel Failure Monitoring System. These samples will be processed in the gas chromatograph in the analytical services laboratory to determine impurities (helium, hydrogen, oxygen, nitrogen, methane, and carbon monoxide). Samples are also taken for determination of inpurities not detectable by the gas chromatograp:' such as tritium.
50 n
,9 n=
(10 D I.) Amend. 50 June 1979 9.8-3
9.8.2.3 EVST Sodium Characterization Subsystem The EVST Sodium Characterization subsystern is shown in Figure 9.8-2. 45 As EVST sodium is circulated by either EVST sodium pump in the Auxiliary Liquid Metal System, a portion of the sodium is diverted to the EVST l sodium characterization subsystem components. l The ex-vessel plugging temperature indicator (EVPTI) is identical
' to the primary PTI (Section 9.8.2.1) in design and function. In parallel 45 with the EVPTl is the ex-vessel sodium sampling package (EVSSP), identical in function to the primary SSP; only one multi-purpose sampler is provided.
Since EST sodium will become contaminated with radioactive sodium and/or fission products, provision is made for remote operation of the EVPTI and EVSSP. Master-slave manipulators are provided for th; remote manual 45 operational requirements of the EVSSP. 9.8.2.4 IHTS Sodium Cha*acterization Subsysten 45f As shown in Fi ure 9.8-3, the three loops in the Intennediate Heat Transport System (IHTS are sampled and monitored by a three intermediate sodium characterization package (ISCP), each containing a plugging tempe-rature indicator module and a multi-purpose sampler. The inlet to each characterization package is c.onnected to the outlet of the cold t' mp pump in each ! HTS loop and the return line is connected to the returr, lire of the IHTS loop. Sampliny and monitoring can be performed only while the cold trap pump is functioning. Since the sodium in the three IHTS loops 49 ' is not radioac+ive, this subsystem is not designed for remote operations. O 9.8.2.5 EVST, FHC, and IHTS Cover Gas Sampling 45 Provi si onc are m3de in the EVST and FHC cover gas systems and in the IHTS pump wpansion tank pressure equalization lines, to obtain cover gas grab samples, using evacuated gas sample bottles whose contents will be analyzed in the laboratory. Since fission gases can be present in the EVST and FHC cover gas samples, these gas bottles will be shielded for 45 l operator protection. Shielded sample bottles will not be required for IHTS cover gas samples. 45 l 9.8.2.6 Analytical Services Laboratory Space and equipment will be provided in the le labora tory in the Plant Service Building for out of loop analyses of sodium and cover g s samplers. Due to the radioactive nature of many of these samples, the 45 268 1% 9.8-4 jg g
equipment will be designed for shielded and/or remote operotion. Sample prep-aration equipment will include components such as glove goxes, sodium dis-tillation unit and gas transfer >;ystems. Analytical equipment will include components such as gas analyzm anc a gas chromatograph. Controlled waste disposal will be provided for radicactive sample residues. 9.8.3 De_ sign E,atuation The Impurity Monitoring and Analysis System components are designed to accepted industrial and nuclear stand vds to insure structural integrity and operational reliability. The compon ats. opplicable design code and class, plus their seismic category are 'ist<d in Table 9.8-1, 9.8.3.1 PHTS Sodium Characterization Subsystem 45 If the Primary PTI is being repaired or has malfunctioned, monitoring of the sodium it?urity level can be effected by using the Primary SSP, and performing impurity analyses of sodium samples. This subsystem perfonus no control functions, but analytical results are uced to todicate out-of-range 45 conditions, and are also used by the Auxiliary Liquid % tal System to deter-mine primary sodium cold trap operational requirements. This subsystem is not required to funct?cn during an emerg2ncy, nor Isolation valves are is it required for the safe shutdown of t6e reactor,or Primary SSP) from the l sodium providedsamplingto piping. Sodium separate failedlen detectc,rs are provided on bellows components (Printry sodium PTr loop seal valves. The cells containing this eouipment are provided with a6, aerosol leak detectors to signal the relcase of sodium to the inerted cells. 45 The components and piping in this subsystem are contained in lined, inerted cells in the Reactor Containment Building, which can be vented to CAPS of the Inert Gas Receiving and Processing System to prevent the escape of radioactive gases. Suff:cient shielding is provided to prevent radiation overexposure of operating personnel under normal and anticipated faulted conditions. 9.8.3.2 Primary Cover Gas Sampli.ng and Monitoring Subsystem The Failed Fuel Monitoring System gas sampling connections are used to obtain a bulk sample of cover gas in a shielded gas sample bottle which is Sufficient shielding is analyzed in the analytical services laboratory. provided to prevent radiation overexposure of operating personnel under 50 normal and anticipated faulted conditions. 2bb l e 9.8-5 Amend. 50 June 1979
Since there is little delay time and no dilution of the cover gas sar ple, a representative sample is obtained. This subsystem performs no control functions, but analyzed results are used to indicate out-of-range condi-tions, and are also used by the Inert Gas Receiving and Processing System to adjust clean argon purge. This subsystem is not required to function during an emergency, nor is it required for the safe shutdown of the 50 reactor. 9.8.3.3 EVST Sodium Characterization Subsystem This subsystem is not required for the operation of the EVST during an emergency. Isolation valves are provided to separate failed com-ponents (Ex-Vessel PTI or Ex-Vessel SSP) from the sodium sampling loop piping. Sodium leak detectors are provided on bellows seal valves. The cells contairing this equipment are supplied with sodium aerosol leak detectors to signal the release of sodiun to the inerted cell. If the Ex-Vessel PTI is beinq repaired or has nalfunctioned, moni-toring of the sodium impurity level can be accomplished by using the Ex-Vessel SSP, and perf orming impuri ty analyses of sodium samples. This subsystem per-forms no control functions, but analytical results are used to indicate out-of-range conditions, and are also used by the Auxiliary Liquid Metal System to deter-mine EVS sodium cold trap operational requirements. The components and piping in this subsystem are contained in lined, inerted cells in the Reactor Service Building, which can be vented to CAPS of the Inert Gas Receiving and Processina System to prevent the escape of radio-active gases. Sufficient shielding is provided to prevent radiation over-exposure of operating personnel under normal and anticipated f aulted conditions. 45 47 Amend. 50 9.8-6 Jcne 1979 0 mm
9.8.3.4 IHTS_ Sodium _ Characterization Subsys_ tem Continuon availability of this system is not required for the operation of the IHTS or the leak detection instrumentation. The sodium in this component is not radioactive, so that hands-on manual operation is planned. This subsystem is not required for the operation of the IHTS during dn eftergenCy. Isolation Valves are provided in case of piping or component
'ailure. The intermediate sodium characterization packages serving IHTS loops 1, 2, and 3 are installed in cells equipped with catch pans with 49, Class lE leak detection capability.
47 If the PTI is being repaired or has malfunctioned, monitoring of t.1e sodit.m impurity level can be effected by using the MPS, and perfonning im-purity analyses of sodium samples. Sodium sampling and analysis is also re-quired to provide calibration services for steam generator leak detection instrumentation (Section 7.5.5.). This subsystem perfonus no control functions, by anclytical results are used to indicate out-of-range conditions, and are also used by the Auxiliary Liquid Metal System to determine intermediate sodium cold trap operational requirements. 9.8.3.5 EVST, Fn , and IHT5 C_over Gas Sampling Subsystem 45 The shielded gas sample bottles for EVST and FHC cover gas sampling are interchangeable with those used to collect primary cover gas. The shieluing will be suf ficient to protect personnel during sampling, and transfer of the sample to the analytical laboratory. The IHTS cover gas sampling bottles will not be shielded, as this gas is not radioactive. 9.8.3.6 Analytical Services Laboratory The sodium and cover gas analytical equipment and supplies will be 45 ppropriate for analyses planned. Radioactive sodium wastes will be trans-ferred to the Radioactive Waste System for controlled disposal. Radioactive gas wastes will be disposed of under controlled conditions. l 9.8.4 Tests and Inspection 9.3.4.1 Sodium Sampling and Monitoring 45 The PHTS and EVST sodium characterization components are located in inerted cells, so visual inspections will be performed while the system is down, such as during fuel Inading, for the primary sodium components or while the impurity monitoring components are isolated from the rest of the system. Sodium leak detectors will be employed. The IHTS components are located in air atmosphere cells and present no restrictions to direct inspection methods. Amend. 50 9.8-7 ') 4 9
'uv 1 e, j
June 1979
O THIS PAGE IllTEtiTIO!; ALLY LEFT BLAtlK 50 O
, g Anzad. 50 (O f.' Jud6A979 9.8-7a
9.8.5 Instrumentation Recuirements Instrumentation for the Impurity Monitoring and Analysis System provides measurement and control of process variables as required for system operation, performance evaluation, and annunciation of off-nomal conditions. Components will be shut down upon c'eviation of process variables beyond the normal operating limits. 9.8.5.1 Sodium Sampling and Monitoring Sodium sampling loop heating and temperature controls are provided by Piping and Equipment Electrical Heating Control System. Sodium tempera-tures within the PTI's and SSP's are measured by thermocouples which also provide inptt signals to temperature controllers for regulation of process temperatures. Thermocouples are also provided on the magnet of each perma-nent magnet (PM) flowmeter to allow temperature correction of the flowmeter 49 output. Separate control panels are provided for each PTI, SSP, and ISCP. Specific local annunciators are provided on each control panel. The annunciator signals from each control panel are grouped and transmitted to the main control room to indicate abnormal conditions. All sodium valves are provided with leak detectors. heaters and thermocouples, and limit switches. PHTS sodium sampling inlet flow is measured by a PM flowmeter and recorded to assist the operator in flowrate adjustments and to monitor system operation. Two remotely-actuated valves are provided in both the supply and return lines of the sampling loop. The control switches for these valves are interlocked with a high/ low flow alam signal from the inlet flow-meter such that the valves will automatically close upon abnormal flow. The valves are designed to fail closed upon loss of electrical or pneumatic power. A remotely-actuated valve is provided in both the supply and return lines o' the EVST sodium sampling loop. The control switches for these valves are interlocked with high/ low flow alarm signals derived from PTI and SSP in-let PM flowmeters such that the valves will automatically close upon abnormal flow conditions. The valves are designed to fail closed on loss of electrical or pneumatic power. 9.8.5.1.1 Plugging Temperature Indicators Flow control /alves are installed on the PTI inlet and bypass lines. Pe al r .unted, hand-indicating controllers provide remote control of these vai /es uased on PM flowneter signals. Remote manual and automatic control of sodium temperatures are provided in the PTI for determination of the plugging (saturation) tempera-ture. Automatic cycling control is accomplished by using high- and low-flow signals for control of the cooling gas blower. At the high-flow setpoint, the blower is energized to cool the sodium until the flow decreases to the
- 9. 8- 8
?b0 l3]
Amend. 49 Apr. 1979
low setpoint, where the blower is deenergized. With the blower off, the sodium tanperature and flow increase and the cycle is repeated. Sodium temperature at the PTI orifice and PTI outlet flow are recorded on the same recorder to allow fast direct comparison and interpretation of the data. Manual PT! control is accanplished by manually switching the blower on and of f as the flow variations are observed on the recorder. Ala rms are provided to indicate that the high- or low-flow and high orifice temperature setpoints have been exceeded. 9.8.5.1.2 Sodium Sampling Packages Sodium flow rates through the MPS and SSP bypass are controlled manually. PM flowmeter signals are recorded and also indicated at the manipulator station. MPS furnace temperatures are controlled in three zones: (1) one three-mode controller (proportirnal plus rate plus reset) is used to control all heaters in the upper furnace, which is the main heat zone for the sampler cup, (2) two two-mode controllers (proportional plus reset) are used to control the two axial zones in the lower furnace. The remaining heaters on the MPS assembly on the argon-vacuum line, and inlet and outlet sodium lines and valves, are controlled by automatic on-off con-trollers. A high/ low temperature alarm is provided for the upper furnace. 45 O S0 Amend. 50 June 1979 9.8-9 7 g r) ,.q L. v ' .) . ) ,)
TABLE 9.8-1
, IMPURITY MONITORING AND ANALYSIS SYSTEri STRUCTURAL DESIGN CRITERIA Design Code System Component Seismic Code Class Category Primary Na Charac- Piping to/from Auxiliary Liquid Metal ASME Section 1 I terization System and isolation valves III Sampling loop piping and valves ASME Section 3 I
.III 48 Primary plugging temperature indicator 45 ASME Section 3 I III
? 45 l Primary sodium sampling package E 48 ASME Section 3 I III 36 '
Electric Hoist Coml - III i Master Slave Manipulator Comi - II 48 45l c Radiation Shielding Coml - I ca 36 y ca 8e F 50 36 $8
TABLE 9.8-1 (cont'd.) IMPURITY MONITORING AND ANALYSIS SYSTEM STRUCTURAL DESIGN CRITERIA Design Code Seismic S_ystem Component Code Class Category ___ 45l 50 EVST Na Character- Piping to/from duxiliary Liquid Metal Sys. ASME Section 3 I ization III 47l44 Sampling loop piping and valves ASMF Sec.III Ex-vessel plugging temperature indicator ASME Section 3 I III
- 49 48
'o c 36 Ex-vessel sodium sampling package ASME Section 3 I L TII Electric hoist Coml. - III Master-slave manipulator Com1. - II Radiation shielding Coml. - II 36 EVST Ar Sampling Ar sampling piping ASME Section 3 I III Ei N : 88~
-F3
-u '
i 6 . 0 0 0
h 1 VAPOR l w/ l TRAP F AILED FUEL !=l]s MONITORING SYSTEM l I l- , 7 I i 4 l l l PRIMARY COVER l 3 i g GAS SAMPLE I PPTl PSSP l I i l I I
\ \"/
REACTOR l 1 I l , r i i e . P I I I
- l rs a, N
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/T_L. i i i MA KE UP L i
i F 7,
.y PUMPS I { _g g f s! - 1
'3 OVERFLOW C3 VESSEL l I _ __ _ _ _
s
/
(Ii \ n s /
- o. Rt R T A y* VE SSE L 38 Figure 9.8-1. PHTS Sodium Characterization and Primary Cover Gas Sampling and Monitoring Subsystems
=
$ EVST g - - _ _ ,
is I . I i l I l l l l 1 1 __ _ _ _ q ,_____ 1 -
; 4 l
l euMe f'k l N' l euce l l l f ~ ~l 9 f 3~ ~I l P _. l- -- I _ _I l li
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w 3 I L 4 I l l l EVPTl EVSSP pp rxnc2c. cho WEN Figure 9.8-2 EVST Sodium Characterization Subsystem N C3 _ _w 8 9 e
9.9 SE,RVICr WATER SYSTEMS 9.9.1 . Normal Plant Service Uater System 9.9.1.1 Design Basis The Normal Plant Service Water System 1s a non-safety related system designed to provide cooling wai.er for the Normal Chilled Water System chiller condenscrs, the Secondary Service Closed Cooling Water 33 System and other equipment listed in Table 9.9-1 during normal plant operation and planned outages. The system will be desianed according to the ASME Section VIII/ ANSI B31.1 requirements. 9.9.1.2 System Description The Normal Plant Service Water System is shown in Figure 9.9-3. The system consists of two (approximately 26,600 GPM) 100 percent 43 l 15 capacity electric motor driven deep well turbine pumps and the required piping, valves and instrumentation. The Normal Plant Service Water is pumped from the basin of the Circulating Water System cooling tower to the equipment to be cooled, and is returned to the cooling tower
! return header. The pumps are located in the Circulating Water Pump-431 house. Normally, one punp is operating with the second pump in an 33 auto-standby mode.
A g The components served by the Normal Plant Service Water System are listed in Table 9.9-1. Design data for the major system components are listed in Table 9.5-2. 15l 9.9.1.3 Safety EvaiJation The Normal Plant Service Water System is a nonseismic, non-safety class system. Cocling water for the Emergency Chilled Water System chiller condensers, and standby Diesel Generators, for the safe shutdown and the maintenance of the safe shutdown condition, is provided 15 by the Emergency Plant Service Water System as described in Section 9.9.2. 33 l 9.9.1.4Tests and Inspections 15 The Nor mal Plant Service Water pumps are tested at the manu-facturer's facility and retested in the system prior to continuous plant operation. The operation of the pumps will be rotated to equalize wear. 9.9.1.5 Instrumentation Application 15 Indication of the Normal Plant Service Water header pressure is provided in the Control Room. Normal Plant Service Water aischarge 9.9-1 94n ,n)- Amend. 50 fuO U. June 1979
header pressure is annunciated in the Control Room. A logic circuit is available to automatically start the standby pump when the operating pump motor trips or is inadvertently stopped. 15 l33 l 9.9.2 Eme_rgen_cy_ P l a n t Se rv i ce Wa te r Sys t em 9.9.2.1 Desian Basis The Emergency Plant Service Water System is designed to provide 43 l suf ficient cooling water to permit the safe shutdown and the maintenance of the safe shutdown condition of the plant ir. the event of an accident resulting in the loss of the Normal Plant Service Water System or the loss of the plant AC power supply and all offsite AC power supplies. The Emergency Piant Service Water System is not t. sed during normal plant operdtion. The system provides the Emergency Chilled Water System 43 33.15 chiller condensers and the Standby Diesel Generators with cooling wat.er. The Emergency Plant Service Water S stem includes the Emergency Cooling Towers and Emergency Cooling Tower Basin, as described in Section 9.9.4. The Enorgency Plant Service Water System is designed to Seismic Category I requirements as defined in Section 3.2. Pumps, valving and piping required for the safe shutdown of the plant are designed to 43 ASME Section Ill, Class 3 requirements, as defined in Section 3.9.2. All electric motors serving the system are connected to the Class lE onsite power supply. In case of loss of plant and offsite power, these motors are switched automatically to the onsite power supplies. 15 33 The piping and equipment for each redundant loon of the system is - physically separated or protected with a barrier to conform to common mode f ailure cri terion. System piping is below ground. The Emerooncy 50 Cooling Tower structure is tornado missile hardened as described in Section 9.9.4.1. 9.9.2.2 Sy_s tem Descrip_ tion The Emergency Plant Service Water System (EPSW) consists of two 100 percent capacity fully redundant cooling loops. Eech cooling loop includes one circulating pumo, one make-up pump, one emergency cooling tower and associated piping, valves, instrumentaticn and controls. Figure 9.9-4 shows the various equipments and represents the system 50 component configuration and relationship. The components served by the Emergency Plant Service Water Systnm are listed in Table 9.9-3. Design data on the major system com-43 33 ponents is listed in Table 9.9-4. Upon loss of Normal Chi . ;ed Water or upon start of the Standby DieselGenerators,theEPSWpumps,EPSWmakeuppumps,andCoofingTower 50 fans will automatically start and provide cooling water at 90 F maximum 9.9-2 Amend. 50 June 1979
):ov LUv t m
?nt
to the Emergency Chil!ed Water Chiller Condensers and the Standby Diesel Generators. The EPSW pumps take suction from the Emergency Cooling Tower operating basins which are located directly below the pumphouses and adjacent to the common storage basin. During system operation the EPSW makeup water pumps will transfer wat !r from the common storage basin to the redundant operating basins i.o compensate for evaporative and drif t losses from the towers. Cooled water from the Emergency Cooling Tower operating basins is pumped via underground supply mains to the emergency loads in the DGB and SGB. After cooling the emergency chillers and the standby diesel generators, warm water is returned, also through underground mains, to the Emergency Cooling Towers. To account for seasonal temperature variations, temperature control valves served by electro-hydraulic operators bypass a portion of the returning water back to the pump suction. A temperature indicator controller automatically adjusts U the v9 ves as required to maintain supply temperature above 55 F, the minimum required for chiller operation. In addition to cooling chilled water system and electric power system loads, each loop of the EPSW System provides a connection to supply water to the Non-Sodium Fire Protection System. The EPSW pumps and the Emergency Cooling Tower Basin are designed to allow fire protection operation while maintaining the capability for supplying 100 percent cooling to the emergency loads. The fire protection pumps are provided with a flow totalizer that will automatically terminate operation when a prescribed amount of water has been used (see Section 9.13). This ensures that the guaranteed 3n day supply of water for EPSW system operation will not be compromised. In addition, this system is connected to the EPSW loops in such a manner as to preclude a single failure from compromising the capability of the EPSW system to perform its 50 required function. 9.9.2.3 Safety Evaluation The EPSW system is a Seismic Category I, safety related system designed to have 1007 redundancy in both active and passive components. The system is provided with AC power from the Class lE power sources. EPSW Loop "A" is supplied from Class lE Division 1 and Loop "B" is supplied from Class lE Division 2. This arrangement assures that 100 percent cooling capability will be available even if one of the Standby Diesel Generators or one of the EPSW loops should fail. The EPSW system is a fully automatic system, normally controlled from the Main Control Panel in the Control Room. Should the Control Room become uninhabitable for any reason, such as smoke or fire, reu .n-dant controls have been provided that will allow full operation of the 50 system from a control panel in the Diesel Generator Building. 9.9-3 Amend. 50 June 1979 200 2 [] ['
During the initial phase of recovery from an accident, one Emergency Plant Service Water loop satisfies the cooling of the Standby Diesel Generators and the Emergency Chilled Water Chiller Condensers. The Emergency Plant Service Water System is capable of accommo-dating any single component failure without affecting the overall system capability of providing cooling water to achieve a safe shutdown con-di tion. 15 l 9.9.2.4 Tests and Inspections The system components will be tested at the manuf acturer's facili-ties, and a complete system test will be accomplished prior to plant operation. The EPSW System does not operate during normal plant operations. However, the system, including all active compenents will be operated periodically during the year in conjunction with the Standby Diesel Generator testing program as outlined in USNRC Regulatory Guide 1.108. The system c .i be proven operable at any time by manual initiation. Inservice inspections will be conducted according to ASME Section XI, as described in Section 9.7.2.1.g. In~ addition, isolation valves and pressure test connections 50 on the supply and return headers in the pumphouses and the DGB permit inservice inspection of the buried piping by hydrostatic testing. 9.9.2.5 Instrumentation Application InstrJr.entation will be provided for local and/or remote (Control Room) indicat:or of the following parameters as indicated: pump discharge pressure (local / remote) 50 supply temperature (local / remote) storage basin level (local / remote) diesel generator and emergency chiller flow rate (remote) diesel generator and emergency chiller supply temperature (local) diesel generator and emergency chiller return temperature (local / remote) diesel generator and emergency chiller supply and return pressure (local) 50 perating basin level (local / remote) makeup water flow (local / remote - alarm on low) A flow switch, located in the return line from each diesel generator and emergency chiller will detect an abnormal low flow condition 43 3,c and energize an annunciator in the Control Room. Secondary Service Closed Cooling Water System 15 l 9.9.3 The objective of the Secondary Service Closed Cooling Water (SSCCW) System is to provide cooling to auxiliary equipment located in the turbine building. Amend. 50 ) June 1979 9.9-4 2 g [] [))C 9 t
9.9.3.1 Design Basis The Secondary Service Closed Cooling Water (SSCCW) System is desig. to provide adequate cooling water supply for the power generation equipmen during normal plant operation. The SSCCW System is designed in accordance with ANSI B31.1 and is not safety related. 15 9. 3.2 System Description N The SSCCW System is shown in Figure 9.9-S and consists of a single closed loop with two 100 percent capacity centrifugal pumps in pa ra l l el . The system utilizes two 100 percent capacity SSCCW heat exchangers are cooled by the Normal Plant Service Water System. The cooling water for the SSCCW discharges into o common discharge header where the SSCCW pumps take s tion. The SSCCW System provides cooling 15 water to the equipment listeu in Table 9.9-5. A surge tank, located above the SSCCW pump suction, accommo-dates system volume changes, and maintains static head on the pumps in the SSCCW System. Makeup water to the SSCCW System is supplied by a connection from the condensate pumps to the SSCCW pumps common suction header. Tank level is maintaned automatically by means of level trans-mitters and controllers moun ted locally. A signal from these transmitters opens the level control valve on the conderisate line to maintain the surge tank at the desired level. The surge tank is readily accessible during operation for manual level adjustment if desired. A bypass line is provided around the SSCCW heat Exchangers and includes a temperature controller installed at the discharge manifold downstream of the heat exchangers to regulate the bypass flow thereby providing a tempering effect to maintain a costant 95 degree F coolina water. 9.9.3.3 Safety Evaluation 15 The Secondary Service Closed Cooling Water (SSCCW) System is not a safety related system and is not required during an emergency shutdown of the plant. 15 9.9.3.4 Tests and Inspections Pumps for the SSCCW System are tested prior to installation and again prior to plant operation. System subsections normally closed to flow are tested periodically to ensure their operability and integrity of the system.
'. r Amend. 50 5UCq LUk>no 9.9-5 June 1979
15 9.9.3.5 Instrumentation Applications The conmon discharge header of the SSCCW pumps is monitored for low pressure and alarred in the Control Room. Pressure indicators are Tempera-43 l provided at eachispump ture indication discharge located on eachand each line suction heatand exchanger outlet. dis-on the conson charge manifold of the SSCCW pumps. 9.9.4 Emergency Cooli_ng Towers and EneroencL C oolinq Tower Basin 9.9.4.1 Desinn Baris The Emergency Cooling Towers (Figure 9.9-4) operate as part of the Emergency Plant Service Water Systen (Section 9.9.2) to provide cooling water for the Emergency Chilled Water System chiller-condensers, and for the Standby Diesel Generators. Uninterrupted cooling water supply is required for the above equipment. The failure of the Normal Plant Service Water System requires the operation of the Emergency Cooling Tcwers for the safe shutdown and the maintenance of the safe shutdown condition of the plant. The Emergency Cooling Towers do not operate under normal plant conditions except for routine testing. The Emergency Cooling Towers and Erergency Cooling Tower Storage Basin are designed according to the applicable requirements of Regulatory Guide 1.27< The integral pipir.g associated with the cooling towers is 3J designed according to ASME Section III, Class 3 requirements. The capa-city of the Energency Cooling Tower Basin is sufficient to permit the uninterrupted operations of the Emergency Plant Service Water System for that period of time (minimun of 30 days) needed to evaluate the situation, to take corrective action to mitigate the consequences of an accident, and to take any neCessary measures to permi t water reple-nishnent. The storage capacity of the Emergency Cooling Tower Storage 50 Basin is based on the historical regional measurements, combining the worst recorded 30 day average period of maximum difference between dry bulb temperature and dew point temperature (t.T) and the highest wind speeds recorded during the same 30 day period, such that the combination of tT and wind speed occurring simultaneously results in the maximum amount of evaporation and drif t loss of water from the cooling tower. The Emergency Cooling Towers are designed not to exceed the maximum permissible cooling water supply temperature, using the worst one day and worst 30 day periods of regional meteorological records when the heat transfer to the atmosphere is minimized and maximum cooling water supply temperature is induced. The worst one day neriod of the 43 record is assumed the first day of the worst 30 day period. Amend. 50 9.9-6 June 1979 U > .. s lj
The Emergency Cooling Towers pumphouses, operating basins and "0 storage basin are designed to withstand the most severe natural phenomena (e.g. , Safe Shutdown Earthquake, tornado, tornado missiles, wind, Probable Maximum Flood or drought). The design has the necessary redundancy of components. 50 l Electrical power for the Er..ergency Cooling Tower fans, pumps, and control equipment is provided from the Class lE AC power supply. One fan is provided with electrical power from System Class lE Division 1 50 and the other from System Class lEB Division 2. 15l43 9.9.4.2 Design Description The Emergency Cooling Tower Structure consists two of pumphouses (containing the pumps and piping of the EPSW System, Section 9.9.2) located directly above the operating water storage basin. The cooling towers, pumphouses and operating basins are 1001 redundant Seismic Category 1, Tornado protected structures. The common storage basin is a Seismic Category I, flood and tornado protected structure. The storage basin has sufficient storage capacity for 30 days of operation, including 60,000 gallons of water storage for the seismic Fire Protection System plus adequate allowance for drift and evaporation losses. Each cooling tower is designed to achieve +he required heat dissipation rate at any time, approximately 2.36 x 107 BTU /HR at the maximum Emergency Plant 50 Service Water Flow of approximately 3600 gpm. The change in water chemistry due to the absence of blow-down from the cooling toaers has minimal effect on operation of the Emergency Plant Service Water System. Proper selection of the Emergency Plant Service Water components and applied biocide additives provide compen-sation for the increased tube fouling, resulting from the change in the water chemistry. The maximum nakeup water required after 30 days of operation is approximately 100,000 gallons per day. In case the make-up water is not available af ter 30 days, make-up water can be supplied by either truck, rail or temporary piping from the Clinch River or from the nearby potable water systems. The top elevation of the Emergency Cooling Tower Basin is 818 f t. which is 9 ft. above the probable ma (imum flood level. The basin maximum water level is at 810 f t. ele"ation. The entire basin and the cooling tower supports are founded on siltstone. The basin is a below grade reinforced concrete structure. For further details on the 50 basin, refer to Section 3.8.4.1.5. Each Emergency Cooling Tower consists of a single cell, pro-vided with an induced draft fan system. Each cooling tower is enclosed 50l43 in a Seismic Category I, tornado missile protected structure. The water 9.9-7 Amend. 50 June 1979
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intake and discharge piping are located within the tower or safely below the ground for tornado missile protection. The water intake and dis-charge pipir.g and the internal distribution piping are Seismic Category I, ASME Section III, Class 3 design. Each Emergency Cooling Tower 50l has a design flow rate of 3600 GPM. The Emergency Cooling Towers are of counter flow, induced mechanical draft design. The internal distribution piping distributes the intake water evenly over the fill area so that sufficient water area is exposed to the counter flow air to provide evaporation for the required heat removal. The air counter flow is provided by the induced draf t fans. Drift eliminators are located above the internal water distri-bution piping and below the induced draft fans. The drift eliminators are a zigzag pattern of channels which prevent water carryover through the fan stack. The Emergency Cooling Towers are supported by the reinforced 50 concrete storage basin. The top of the cooling towers is approximately 44 f t. above the maximum water level. 50l The Emergency Cooling Tower Basin is filled with potable grade water which is treated for bacteria control. The quality of the stored water is analyzed at reguiar intervals and the required biocide additive is injected manually in quantities required to control seasonal variations of the bacteria growth. The Emergency Cooling Towers anc Emergency Cooling Tower Basin will be seismically analyzed as described in Section 3.7. 9.9.4.3 Safety Evaluation The Emergency Cooling Tower structure consists of two 100 50 percent capacity cooling towers pumphouses, and operating basins and one 100 percent apacity below grade cooling water storage basin. The entire structure h Seismic Category I, tornado, and flood protected. Piping, associated with the Emergency Cooling Tower is designed to ASME Section III, Class 3 requirements. The structure can withstand the most severe natural phenomena expected, and other site related events, such that the Emergency Cooling Tower cooling capability is assured under required conditions. The method of analysis is similar to that used 50l for other Seismic Category I structures. The entire structure is designed to withstand the Safe Shutdown Earthquake. The fill, drift eliminators, motors, mechanical drives, piping, electrical conduit, cables and supports will be seismically analyzed in accordance with the 43 procedures discussed in Section 3.7. Amend. 50 June 1979 9.9-8 m{ Lou m L. s u t
The Emergency Cooling Towers and operating basins are above the probable maximum flood level. The flood level considerations are discussed in Section 3.4. The Emergency Cooling Tower pumphouses, except for the make-up pump pits which extend down to elevation 771'-0", are also above the probable maximum flood level. However, the Emergency Cooling Water 50 Make Up Pumps are submersible thereby providing system flood protection. The Emergency Cooling Tower structure is designed to withstand tornado windforces and tornado missiles and the cooling tower internals 50 are protected by the enclosing structure. The tornado and wind load;ngs j and the Missile Protection are discussed in Sections 3.3 and 3.5 respectively. All materials used for the Emergency Cooling Tower Structure 43'* are designed to be non-flammable in order to negate the possibility of loss of the cooling function due to fire. 50l In order to evaluate the capability of the Emergency Cooling Towers and Emergency Cooling Tower Basin to act as an ultimate heat sink for the Emergency Plant Service Water System for a minimum period of 30 days, a detailed analysis will be done using the following conserva-tive assumptions:
- 1. The Emergency Cooling Tower Structure is subjected to the maximum probable heat load. This load corresponds to the heat removal duty of the Emergency Plant Service Water System to control a postulated design basis accident and is listed on Table 9.9-3. During all other modes of operation the Normal Plant Service Water System removes the heat loads.
- 2. The postulated design basis accident is assumed to occur under canditions that minimize the heat removal rate, and maximize the water usage as follows:
- a. Meteorological Condition for Minimum Heat Removal Rate.
The meteorological condition for minimizing heat removal rate is the highest wet bulb temperature that may occur at the inlet to the cooling tower. Wet bulb temperature is the only meteorological condition significantly affecting the water temperature produced by mechanical draft cooling towers. Each Emergency Cooling Tower is designed to dissipate the maximum expected heat load during the first 24 hours after a design basis accident assuming average wet bulb temperature for the worst aay of record. Amend. 50 June 1979 9.9-9
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- b. Water Usage Maximizing Conditions The conditions for maximizing water usage for 30 days may be summarized as follows:
(1) Wet bulb and dry bulb temperatures for the worst 24 hours on record are assumod for the first 24 hours af ter design basis accident. For the following 29 days, the worst month of record is assumed. U (2) 90 F initial basin water temperature. 50 (3) Maximum specified cooling tower drif t loss of .01; maximum flow. (4) All pumps and fans operating in the active trains. 43, (5) 60,000 gallons of water for fire protection use is not 50 considered available for cooling.
- 3. The maximum water usage based on the above assumptions will be cal-culated by a computer program that mode' time history of the heat loads and the cooling tower heat removal capability. Normal com-ponent leakage and losses due to a postulated pipe rupture will also be taken into account.
U. S. Department of Comnerce weather data for Oak Ridge, Tennessee Township and area stations for the years January 1951 through December 1971 will be used in the analysis. Evaporation rate from the Emergency Cooling Tower is calculated using the heat balance across the Emergency Cooling Tower. 50 9.9.4.4 Test and Inspec tion The Emergency Cooling Tower fans will be tested prior to installation of the manufacturer's facilities. After construction of the Emergency Cooling Tower structure is completed, but prior to normal plant operation, the cooling towers will be tested for cooling perfor-mance, evaporation and drif t rates according to the Standards of the Cooling Tower Institute. 5d The applicable Emergency Cooling Tower components will be tested periodically in conjunction with the Emergency Plant Service 43 SC W ter System according to the requirements of ASME, Section VI. Amend. 50 9.9-10 June 1979 L,
9.9.4.5 Instrumentation Apolication The following instrumentation is provided at the Emergenci/ Cooling Towe. structure with signals transmitted to the Control Roon:
- a. Level transmitters, for storage basin level indication readout.
- b. Temperature sensors (for each cooling tower) for discharge cooling water temperature readouts.
- c. Temperature sensor (in the storage basin) for discharge cooling water temperature readout.
- d. Air flow switches (at the cooling tower fan discharge) to 43 33 indicate croper fan operation by status l'chts.
15l 9.9.5 River Water Service The River Water Service supplies Clinch River water as makeup to
- 43) the Main Cooling Tower, Emergency Cooling Tower structure and Plant Water Treatrent Facility during normal operation. A basic flow diagran of the River Water Systen is provided in Figure 9.9-6.
15 9.9.5.1 Design Basis The River Water Service (RWS) is designed to provide adequate river water to replace circulating water lost from the Main Coolinc Tower during normal operation due to drif t, evaporation and blowdown. The RUS 33l lso supplies the Plant Water Treatment Facility to meet all process ara potable water demands during normal operation. Design flow rate for che RWS is 9,000 gom. The RWS piping is designed and tested in accordance with ANSI B31.1 and is not safety related. The River Water Intake design incorporates two subrerged, perfo-rated pipe intakes which are specifically designed to minimize their impact upon the aquatic life present and eliminate interference with commercial river traffic in the Clinch River. 15 l 9. 9. 5.2 Rystem Description The RWS consiste # a pump house located at the shore of the Cli:ich River, two perforated pipe inlets, two River Water Service pumps 111 designed for 9,000 gpm each and the associated piping and valves necessary to provide river water to meet the plant demands. 9.9-11 Amend. 43 Jan. 1978 2b'3 2:;
Two backwash 1ines are provided to allow removal of debris collected on the perforated pipe inlets. A recirculation line is provided for the river water service pumps to preclude low flow problems associated with the pumps. 15 9.9.5.3 Safety Evaluation The RWS is not a safety relat<>d system and is not r equired during an emergency shutdown of the plant. 9.9.5.4 Tests and Inspections 15 River Water Service Pumps are tested prinr to installation and again prior to plant operation. The RWS is normally in service. 15
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C " Flow, pressure, and alarms are provided as required on the RWS. Pump discharge flow will be regulated by level control of the main cooling tower basin. O
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m u m u m u moe rct w P P P P oeno NSI L a a a a N N N N s r e e e e o t t t t - - - - r t a a a a e a i i is i g r d r d d r d n ee e ee e a p ml m ml m h rp r rp r W c eu e eu e WCLL s x t o t t o t HCLL r E nC ns nC ns CSAA e I I r I I r NSI L l t d e d e T l a &i &l &i &l N i e u o u o E h H yl yo yl yo N C rF rC rF rC @ O P M O W H C W C C S A i a m r i a m r i a m r i a m r E T O C N S P P P P N 1 5 3 3 1 3
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O 15 TABLE 9.9-2 NOPJ1AL PLANT SERVICE WATER SYSTEM MAJOR COMPONENTS APPROX. NPSW FLOW DESCRIPTION QUANTITY FOR EACH COMPONENT
)g Normal Plant Service Water Pump 2 26,600 43 50 NOTE: NPSW - Normal Plant Service Water (rigure 9.9-3)
Amend. 50 9.9-14 June 1979
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15 TABLE 9.9-3 COMPONENTS SERVED BY EMERGEhu'( PLANT SERVICE WATER SYSTEM Component Location Component Service Requirements Flow *gWT BTU /HR Component Bldg. Cell Elev. GPM F 0 (X10 ) Sta-fby Diesel Generator A DGB 511 816'-0" 1500 90 13.2 Standby Diesel Generator B DGB 512 816'-0" 1500 90 13.2 Emergency Chilled Water e, System Chiller A SGB 216 733'-0" 2l00 90 10.4 6' Emergency Chilled Water U 50 System Chiller B SGB 217 733'-0" 2100 90 10.4 SC P') c7s ':..tering Water Temp. co N
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O 15 TABLE 9.9-4 EMERGEf4CY PLANT SERVICE WATER SYSTEM MAJOR COMP 0f1ENTS Design Data For Description Quantity Each Component Emergency Plant Service 2 3600 GPM Water System circulating pump 110 ft. total head 43 33 Emergency Cooling Tower 2 3600 GPf1 Emergency Plant Service 2 150 GPM 50 Water Make-Up Pumps 92 ft. total head 43 33 O Amend. 50 9.9-16 n June 1979
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292fi Water Chargas for Private Fire Protection,1974 801 Facilities Handling Radioactive Materials,1975 802 Fire Protection Practice for Nuclear Reactors,1974 901 Uni form Coding for Fire Protection,1976
- 10. ASME Boiler and Pressure Code, Section III and Sectian X (For Containment penetrations andlsolation valves) ,
48 9.13.1.2 System Description
- a. The Non-Sodium Fire Protection System consists of the following:
Wdter Supply System tiet Sprinkler System Preaction Sprinkler System Water Spray System Carbon Dio;;ide Gas Blanketing System Halon 1301 Gas Blanketing System Standpipe System Portable Fire Extinguisher System Fire Detection System The general description of the above systems is provided in Table 9.13-4. The P&I Diagram of t' Water Supply System (Figure 9.13-1) shows the physical location e 'a lant buildings and the water supply system. The P&ID for ti. r 1301 Gas Blanketing System 50 is shown in Figure 9.13-la. The f .: prevention and protection systems to be provided for all the areas associated with the safety relate a ntructures, systems anc components are listed in Table 9.13-3.
- b. The Control Room fire protection and extinguishing system consists of Halon 1301 portable gas fire extinguishers for manual fire-fighting operations at indicated in Table 9.13-3. The design des- ,1 3
"'otion of these systcms is provided in Table 9.13-4.
_ standpipes are provided outside the Control Room far manual - 48 wecer hose spray protection. The Computer Room represents a separate firr zone within the Control Room, therefore, the supply and exhau.t ducts to the Computer Room are provided with motor operated fire dampers. The automatic or remote manual operation of these dampers provides sealing for the Computer Room. 20 Amend. 50
- 9. i 3-ll June 1979 y n
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The Computer Room fire protection and extinguishing system consists of the Halon 1301 gas blanketing system with portable fire extinguishers 48 for manual fire-fighting operations as indicated in Table 9.13-3. During a fire, the Computer Room will be completely isolated from the Control Roe.n by the Control Room HVAC System. After the fire is extinguished, the Control Room HVAC system will be placed in a 100% exhaust mode and the Computer Roc., exhaust duct damper will be opened. The negative pressure established by the exhaust mode will clear out the majority of the combustion cod Halon 1301 decomposition products. After some time the supply air o,mper will be opened to speed up the clean up operation. Since the Computer Room ventilation rate is approximately 8 air changes per hour, within a two hour period, the room atmosphere will be sufficiently cleaned to permit personnel entry without dangee to the entering operator. The Halon 1301 storage tanks are located outside of the Control Room area. To limit the potential areas affected by leakage, the Halon 1301 distribution system is isolated from the stoiage tanks by redundant isolation valves. At the vicinity of the sterage tanks where the probability for small leaks exists, redundant detectors will be provided to alert plant operating personnei of the presence of Halon 1301 in the tank area. Since the Halon 1301 storage tanks are located outside of the Control Room are, the distributing pipeing is isolated from the storage tanks by double isolation valves and the Halon-1301 system will be manually acti-vated by Control Room personnel. For these reasons, an accidental leak of Halon 1301 into the Control Room or its HVAC System is highly unlikely. Therefore, no Halon 1301 detectors have been planned for the Control Room Ventilation System. The entire Halon 1301 system is located in a tornado missile proof and flood protected structure. The Halon 1301 storage system up to and including the storage system insolation valves will be designed to Seismic Category I requirements. The detailed system description 20 including any required redundancy will be provided in the FSAR.
- c. The design features of the Fire Detection System are provided in Table 9.13-4. The alarm system is designed such that the failure of single fire detection device does not affect the operation of remaining detection devices connected to the same detection zone.
The interconnecting circuitry between the detection devices within a zone is continuously supervised, and a break in the circuitry is annunciated both locally and in the Control Room. 13 9.13-12 Amend. 48 Feb. 1979
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CHAPTER 11.0 - RADI0 ACTIVE WASTE MANAGEMENT TABLE OF CONTENTS PAGE 11.0 RADI0 ACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS li.1-1 34 11.1.1 Modes of Radioactive Waste Production 11.1-1 11.1.2 Activation Product Source Strength Models 11.1-2 11.1.3 Fission Product and Plutonium Release Models 11.1-5 11.1.4 Tritium .oduction Sources 11.1-7 11.1.5 Summary of Design Basis for Deposition of 11.1-7 Radioactivity in Primary Sodium on Reactor and Primary Heat Transfer Surfaces and within Reactor Auxiliary Systems 11.1.6 Leakage Rates 11.1-10 11.2 LIQUID WASTE SYSTEM 11.2.1 Design Objectives 11.2-1 11.2.2 System Description 11.2-2 11.2.3 System Design 11.2-4 11.2.4 Operating Procedures and Performance Tests 11.2-5 11.2.4.1 Operating Procedures 11.2-5 11.2.4.2 Performance Tests 11.2-6 11.2.5 Estimated Releases 11.2-6 11.2.6 Release Points , 11.2-6 11.2.6.1 Nuclear Island 11.2-6 11.2.6.2 Balance of Plant 11.2-7 30 11.2.7 Dilution Factors 11.2-7 11.2.8 Estimated Doses 11.2-8 m ,o onr DUO Amend. 34 ll-i Feb. 1977
PAGE 11.2.8.1 Doses from Exposure to Liquid Effluents 11.2-9 50 Appendix to Section ll.2A 11.3 GASEOUS WASTE SYSTEM 50 11.3.1 lesign Base 11.3-1 11.3.2 System Description 11.3-1 11.3.2.1 Process Flow 11.3-1 11.3.2.2 Gaseous Radioactive Waste Inputs to System 11.3-6 11.3.2.3 Activity Inventories and Concentrations 11.3-6 11.3.2.4 Release Path Calculations 11.3-7 11.3.2.5 Activity Release Tabulations 11.3-8 11.3.2.6 Radioactive Gaseous Site Boundary and Restricted Area Concentrations 11.3-9 11.3.3 System Design 11.3-10 11.3.3.1 General 11.3-10 11.3.3.2 Equipment 11.3-10 11.3.3.3 Instrumentation 11.3.11 11.3.4 Operating Procedures and Performance Tests 11.3-11 11.3.5 Estimated Releases 11.3-14 11.3.6 Release Points 11.3-15 11.3.6.1 Nuclear Island 11.3-15 1 11.3.6.2 Balance of Plant 11.3-16 31 34l 11.3.7 Dilution Factors 11.3-17 11.3.8 Dose Estimates 11.3-17 11.3.8.1 Dose Rate Estimates ,q ,, 11.3-18
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50 Appendix to Section ll.3A Amend. 50 ll-ii June 1979
@ PAGE 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEM 11.4.1 Desigr Objectives 11.4-1 50 11.4.2 Continuous Monitoring / Sampling 11.4-2 11.4.2.1 General Der,cription 11.4-2 11.4.2.2 Gaseous System Description 11.4-2 11.4.2.2.1 Fost Accident Containment Atmosphere Monitors 11.4-2 11.4.2.2.2 Reactor Containment Isolation Monitors 11.4-2 11.4.2.2.3 Building Ventilation Exhaust Monitors 11.4-3 50 11.4.2.2.4 Condenser Vacuum Pump Exhaust and Deaerator Continuous Vents Tritium Sampler 11.4-4 11.4.2.2.5 Control Room Inlet Air Monitors 11.4-4 11.4.2.2.6 Inerted Cell Atmosphere Monitors 11.4-4 11.4.2.2.7 RAPS and CAPS Monitoring 11.4-4 11.4.2.3 Liquid Systems Description 11.4-5 11.4.2.3.1 Radwaste Disposal System Liquid Effluent 50 M nitor 11.4-5 34 11.4.2.4 Mainter.ance and Calibration 11.4-5 11.4.3 Sampling 11.4-5 11.4.3.1 Process Sampling 11.4-6 11.4.3.1.1 Intermediate Level Activity Liquid Waste Collection Tanks 11.4-7 11.4.3.1.2 Process Distillate Storage Tanks 11.4-7 11.4.3.1.3 Low Level Activity Liquid Waste Collection Tanks 11.4-7 50 11.4.3.1.4 Low Level Activity Distillate Monitoring Tanks 11.4-8 11.4.3.1.5 Ccncentrated Waste Collection Tank 11.4-8 2bb 22,~ Amend. 50 Il-lll... June 1979
PAGE 11.4.3.2 Effluent Sampling 11.4-8 11.5 SOLID WASTE SYSTEM 11.5.1 Design Objectives 11.5-1 11.5.2 System Inputs 11.5-1 11.5.3 Equipment Description 11.5-1 11.5.4 Expected Volumes 11.5-3 11.5.5 Packaging 11.5-4 11.5.6 Storage Facilities 11.5-4 1 11.5.7 Shipment 11.5-4 11.6 0FFSITE RADIOLOGICAL MONITORING PROGRAM 11.6.1 Expected Background 11.6-1 11.6.2 Critical Pathways to Man 11.6-2 11.6.2.1 Doses from Gaseous Effluents 11.6-3 11.6.2.2 Internal Doses from Liquid Effluents 11.6-4 50 11.6.3 Sampling Media, locations, and Frequencies 11.6-4 11.6.4 Ana lytical Sensitivity 11.6-4 11.6.5 Data Analysis and Presentation 11.6-5 11.6.6 Program Statistical Sensitivity 11.6-6 Amend. 50 ll-iv June 1979
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LIST OF TABLES PAGE 11.1-1 Activation Products in Primary Sodium Coolant 11.1-12 11.1-2 Activation Products - Core Structural Material (Stainless Steel 316) 11.1-13 50 11.1-3 Summary of Significant Information Used in CRBRP Corrosion Release Estimates 11.1-14 11.1-4 Fission Product Escape Fractions 11.1-15 ll.1-4A Comparison of Assumed Escape Fraction from Failed Fuel Rods with Measured Escape Fractions in B9D 11.1-16 11.1-5 Parameters Assumed for Plutonium Release to Prinary Coolant 11.1-17 50 11.1-6 Radioactive Argon Cover Gas Activation and Fission Products Release Rates and Resulting Isotopic Release Rate 11.1-18 11.1-7 Activation, Fission Product and Plutoniwn Isotope Concentrations in the Primary Sodium Coolant During 30 Year Plant Lifetime - Design Values 11.1-19 11.1-8 Significant Corrosion, Fission Product and Released Fuel Activities on Wetted Surfaces at Reactor Shutdown 11.1-22 11.1-9 Radioisotope Inventory in Primary Cold Traps 50 After 15 Years Operation 11.1-24 11.2-1 Design Annual Released Activity Inventory 11.2-12 17.2-2 Design Annual Concentration of Low and Inter-mediate Activity Level Input Streams 11.2-14 11.2-3 Design Activity Inventory Stored Af ter Processing 11.2-16 11.2-4 Concentration of Radionuclides at Discharge to Clinch River Design Values 11.2-18 11.2-5 Equipment Description of Liquid Radwaste System 11.2-20 268 L.?'O L j ll-v Amend. 50 June 1979
PAGE ll.2-5A Indoor Radioactive Waste Tanks - Provisions , to Prevent and Control Overflow Conditions ll.2-20a 25 11.2-6 Liquid Radwaste Radioisotope Annual Inventory
- Design Values 11.2-21 50 11.2-7 Design Effluent Concentrations, Maximum Permissible Effluent Concentrations (MPC) and Fraction of MPC at Discharge to River 11.2-23 11.2-8 Estimated Cleaning Process Data-Intermediate Level System 11.2-25 50 11.2-9 Dose Rates Received By An Individual Via Exposure to CRBRP Liquid Effluents (MREM /YR) 11.2-26 ll.2A-1 Wai.er Body Widi.h Foctors for Estimating Gamma Exposure from Contaminated Sediment ll.2A-8 ll.2A-2 Dose Conversion Factors Fc. Total Body and Skin Exposure Via Immersion In Water and Shoreline Deposits ll.2A-9 ll.2A-3 Dose Conversion Factors for Exoosure Via Ingestion of Water to RaJicactive Materials Released From The CRBRP Radwaste Systems 11.2A-12 11.2A-4 Bioaccumulation Factors for Freshwater Aquatic Foods ll . 2A-15 ll.2A-5 Summary of Variables Used in Radiological Dose Evaluations Which Are flot Specific to the CRBRP ll . 2A-17 ll.2A-6 Maximum Permissible Concentration In Water, MPC w , for Continuous Exposure to Radionuclides Released from the CRBRP Radwaste Systems ll.2A-18 11.3-1 Radionuclide Input Rates to Reactor Cover Gas 11.3-19 11.3-2 Gaseous Radionuclide Concentration in Reactor Cover Gas 11.3-20 11.3-3 Activity Inventories in RAPS Process Vessels 11.3-21 50 11.3-4 Activity Concentrations in RAPS Process Gas Streams 11.3-22 ll-vi Amend. 50 June 1979 268 C
PAGE 11.3-5 RAPS Cryostill Performance Characteristics 11.3-23 11.3-6 Activity Inventories in CAPS Process Vessels 11.3-24 11.3-7 Activity Concentrations in CAPS Process Streams 11.3-25 11.3-8 CAPS Decontamination Factors 11.3-26 11.3-9 Radionuclide Release Rates and Release Paths for the Design Value Service Condition 11.3-27 11.3-10 Radionuclide Release Rates and Release Paths for the Expected Service Condition 11.3-29 11.3-11 Annual Activity Release Rates for the Design Service Condition 11.?-31 11.3-12 Annual Activity Release Rates for the Expected Service Condition 11.3-32 11.3-13 Radionuclide Concentrations in head Access Area 11.3-33 11.3-14 Radionuclide Concentrations at Site Boundary from RSB Exhaust 11.3-34 11.3-15 Radionuclide Concentration at Site Boundary from RCB Exhaust 11.3-35 11.3-16 Tritium Concentration at Site Boundary from T.G. Building Exhaust 11.3-36 19 50 ll.3-16a Tritium Concentration at Site Boundary from IB Exhaust 11.3-76 11.3-17 Design Parameters of RAPS and CAPS Process Vessels 11.3-37 11.3-18 External Doses at Site Boundary to an Individual via Gaseous Effluents from CRBRP Design Release Pnints 11.3-38 SC 11.3-19 Release Poir. Description of Internal Whole Body Doses Due to CRBRP Gaseous Effluents 11.3-39 3 11.3-20 Effluent Release Po. .ts Design Data ll.3-39a ll-vii Amenc. 50 qn ,- . June 1979 LC0 LJl
PAGE ll.3A-1 Radiological Data For Isotopes Released from CRBRP Radwaste Systems ll . 3A-12 ll.3A-? Dose Conversion Factors for Exposure to Radioactive Materials Released from the CRBRP Gaseous Radwaste System l l . 3A-15 ll.3A-3 Maximum Permissible Concentration in Air (MPC# ) for Continuous Exposure to Radionuclides Released from the CRBRP Gaseous Radwaste System ll.3A-16 ll.3A-4 Whole Body Dose Cu.. version Factors for Tritium Released During Normal Operating Conditions ll.3A-17 ll.3A-5 Environmental Half-Lives of Radioisotopes Released from CRBRP Radwaste Systems ll.3A-18 ll.3A-6 Summary of Variables Used in Radiological Dose Evaluations which are not Specific to the CRBRP ll.3A-20 11.4-1 Process and Effluent Sampling and Monitoring 11.4-10 11.5-1 Design Objectives in Solid Radwaste System 50 11.5-6 Per Year in Terms of Annual Quantities ll.5-1A Conceptual Primary Cold Trap Removal Sequence ll.5-6a I 11.5-2 Inventory of Design Annual Activity of 50 Shipped Waste 11.5-7 11.5-3 Solid Radwaste Shipments Per Year 11.5-9 44 Amend. 50 II- i 2@l))e[9I7('
LIST OF FIGURES PAGF 11.2-1 Liquid Radwaste System Flow Diagram 11.2-27 11.2-2 IALL Collection and Neutralization Liquid Radioactive Waste 11.2-28 11.2-3 IALL Evaporation Liquid Radioactive Waste 11.2-29 11.2-4 IALL Demineralization and Discharge Liquid Radioactive Waste 11.2-30 11.2-5 LALL Collection and Neutralization Lie;id Radioactive Waste 11 .2-31 11.2-6 LALL Evaporation Liquid Radioactis > Waste 11.2-32 S0 11.2-7 LALL Demineralization and Discharga Liquid Radioactive Waste 11.2-33 11.3-1 Radioactive Gas Flow Paths 11.3-40 11.3-2 Schematic Diagram of the Recycle Argon Circuit 11.3-41 50 11.3-3 Schematic Diagram of CAPS Process Flow 11.3-42 11.3-4 Flow Diagram - RAPS 11.3-43 11.3-5 Flow Diagram - RAPS 11.3-44 11.3-6 Flow Di. gram - CAPS 11.1-45 11.3-7 Flow Diagram - CAPS 11.3-46 11.3-8 RCB Nitrogen Samoling and Analysis 11.3-47 11.3-9 CRBRP Design Radiological Release Points to Atmosphere 11.3-48 11.3-10 P&I Diagram Radioactive Argon Processing System 11.3-49 11.3-11 P&I Diagram Radioactive Argon Processing System 11.3-50 11.3-12 P&I Diagram Cell Atmosphere Processing System 11. 3-51 11.3-13 P&I Diagram Cell Atmosphere Processing System 11.3-52 11.5-1 Solid Radwaste System Flow Diagram 11.5-10 ll-ix 2b 2 -] Amend. 50 June 1979
CHAPTER 11.0 - RADIOACTIVE WASTE MANAGEMENT LIST OF REFERENCES SECTION NO. PAGE NO. 33,3 11.1-11 11.2 11.2-11 11.3 11.3-18c 50 11.5 11.5-5 11.6 11.6-7 9 ll-X Amend. 50 June 1979 7g ,. 00 cJr
TABLE 11.1-9 RADI0 ISOTOPE It4VENTORY IN PRIMARY COLD TRAPS AFTER 15 YEARS OPERATION Percentage Activated Of Isotope Activity in Primary Cold Trap (Li) Corresion Released in Products -Shutdown 10 Days After Shuth wn'-- Cold Trap _ Co60 10 19.6 19.6 CoSS 10 47.9 43.4 54 Mn 50 239 231 Tal82 10 3. 5 3.3 Cr51 10 3.9 3.0 50 I 49 fe lo 0.37 0.32 Fission Freducts Sr89 10 70 60 o S r^n 10 25.9 25.9 10 20.1 17.6 Zr 95 10 37.3 33.5 t.b95 10 37.3 33.5 Rul03 10 53.2 44.6 l 0C, Ru 10 38.0 37.2 106 Rh 10 33.0 37.2 Sbl25 90 2710 2680 127 Sh 5 123 20.5 Te l27m 5 133 125 Tel27 5 146 125 Tel ?M 10 400 376 Tel29 10 400 376 1 131 5 1675 750 Te l32 5 4g 1116 132 Amend. 50 June 1979 11.1--24 2b f3 h!UI)
TABLE 11.1-9 (continued) Percentcge Activated Of Isotope Attivi'v in Prinnry Cold Tract 'C1) Ce ri cs ion Released in Prod; cts Cold leap Shutdow. 10 Days After S utd. ;n F.ssion ProJocts I13~) 5 1110 132 Csl34 60 2155 2135 Cs l36 5 583 342 O Amend. 49 April 1979 11.1-25 ,s O c;u 9 t' O ij
11 2 LIQUID WASTE SYSTEM 11.2.1 Design Objectives The Liquid Radwaste System is designed to process contaminated liquids from the Clinch River Breeder Reactor Plant prior to reuse or release into the environment. The design objectives are to purify and reuse waste liquids where possible and to minimize the total activity in liquid effluents and the total volume of concentrates that will require drumming. The basic approach is to process liquid radwaste so that virtually all of the radioactive material is contained in solid material, to load all solid radioactive material into containers that meet Depart-ment of Transpcrtation regulations, and to transfer the containers to a licensed contractor for processing or disposal. The design objective is to release as little radioactivity as is practicable into the envi-ronment under normal operating conditions. The radioactivity released will be as low as reasonably achievable and less than the limits set 49 by 10CFRZO. Table 11.2-1 -hows the estimated design annual inventory of radioactivity by nuclide which may be discharged after diluhion into the Clinch River. The first column lists the isotopes, the second column gives the half-lives. The third column lists the design acaual activities 49l System. released The fourthfrom the lists column monitoring tanks the annual of the Low activities Activity whicF, may beLevel Liquid (LALL) releasea from the storage tank of the Intermediate Activity Level L: quid (IALL) System. The fif th column lists the sum of the discharged activities per isotope. The two systems are described in detail in the following 49 S ction 11.2.2. Activity in the LALL System comes from sodium spillage washed into the plant drains. Listed activities in the IALL System come from fission products, plutonium, and corrosion products that have plated out or deposited on components washed in the Large Component Cleaning 49l Vessel (LCCV) and the small component autoclave (SCA). Activity levels of the fission products are based on an assumption of 1.0% failed fuel, 30 year irradiation and 10 day decay timc. Corrosion product activity is based on 30 year irradiation and 10 dag decay. The activities listed in the column represent a reduction of 10 by decontamination proceaures of all isotopes entering the liquid radwaste system except for iodine and tritium for which the decontamination factors are T and 1 respectively. A plutonium limit of 100 ppb in the primary sodicm is assJmed. The esF ~ ,of the released radioactivity levels shown in 49l Table 11.2-1 . vat k . The decontamination factors (DF) are conservativc _stimated from op m ting experience of Light Water Reactors. Collection ( activity in the LALL System is based conserva assumption of a 850 gallons per day drainage containing 10~jively pCi/cc.on the This activity is due to sodium removed from the reac. tor for chemical 49lanalysisorduetospillsandcleanupduringnormalplantoperations. The amount of activity stored and released from the LALL System is consavatively estimated in assigning a single value to the fraction of 11.2-1 -) y q ,, Amend. 50 LdO (J, June 1979
the activity removed in the water washes. Both systems conservatively assume that all of the fluids are processed 10 days af ter 30 years of i r ra d i a ti on . The assumption of 10 days decay is connected to the assumption of no spare parts availability. Since, in fact, it is planned to have an inventory of spare parts, a more likely der.ay period is from one to three months. Finally, it should be noted that no planned release 49 of activity from the I ALL System is contemplated under normal operating procedures. It may also be noted tht the activity associated with the I ALL System accounts for the major fraction of the total activity listed in Table 11.2-1. The estimated cleaning process data is provided in Table 11.2-8. The decontuninated water in the IALL System will be recycled 49l'foruseinwashingnomponentsintheLCCV. Processca water in the LALL System monitor tanks will not be reused, since to reuse it would mean injecting radioactivity into the laboratories and showers used by plant personnel. In addition, reuse would require control of the water being supplied to many areas which is both complex and costly. Instead, the liquid radwaste, af ter monitoring to assure compliance with all appropriate Federal and State reguitions, will be released into a diluting stream or 49 used as makeup water i tne IALL system. 11.2.2 System Description Figure 11.2-1 shows the schematic for the liquid radwaste system. The system consists of two subsystems. The first is designed to process liquids with intermediate levels of radioactivity that are reused; the second is designed to process liquids wi' ' iow levels of radioactivity 49lthataredilutedbywaterfronpersonnelshoworsandothersourcesandthen released into the Coo!ing Tower Blowdown Stream (CTBS). The decontamination of liquid radwastes in both systems is carried out in the following sequence: The liquids are collected, neutrallied if necessary, and then treated in batches to one or more cycles of filtration, evaporation, and deminera-49 lization. Normally, the stream in the LALL and IALL Systems are kept completely separa te. However, the 1esign provides the option of utili-zing the equipment in either system as a backup for the equipnent in the other system through a variety of interconnections shown on the schemagic. The Radwaste System provides a gross decontamination factor (DF) of 10 4 49 except for iodine and tritium where the decontamination factors are 10 and 1 respectively. These DFs are based on operating data compiled in WASH-1258. The concentrated liouid radwaste is collected from the bottom of the evaporator and transferred to a solid radwaste system for solidi-fication and disposal by burial of f-site. The contaminated resins in the demineralizer and the contaminated filters are also transferred to the solid radwaste system for packaging and disposal. Table 11.2-2 shows the design annual concentration of activi-49 l ies by isotope flowing into the LALL and I ALL systems. t The fission 50 produc t activities are calculated on the basis of 1.0 failed fuel, 11.2-2 }gjkme2d)b50 June 1979
TABLE 11.2-1 DESIGN ANNUAL RELEASED ACTIVITY INVENTORY (I) Low Level (2) Intermediate (3) Total Isotone ia lf-l i fe(5) Activity (Ci) Level Activity (Ci) Activity (Ci) H- 3 (6 ) 12.3Y 2.86(-3)* 3.31(-2) 3.60(-2) Na-22 2.6Y 5. 34 (-8 ) 6.24(-7) 6.77(-7) na-24 15H 6.65(-9) 7.76(-8) 8.43(-8) C r- 51 28D - 5.14(-7) 5.1d(-7) Mn-54 312D - 1.61(-5) 1.61(-5) Co-58 710 - 5.37(-5) 5.37(-5) Cc-60 5.2Y - 8.37(-6) 8.37(-6) Fe-59 450 - 5.92(-8) 5.92(-B) S r- 39 51D 1.48(-9) 6.58(-7) 6.59(-7) Sr-90 28.8Y 1.04(-9) 4.73(-7) 4.74(-7) Y-90 f>4. l H 1.04(-9) 4.72(-7) d.74(-7) Y-91 53D 1. t -9 ) 1.93(-7) 1.9a(-7) N b- 95 35D 2.2b(-9) 1.26(-5) 1.26(-5) Zr-95 64D 2.28(-9) 1.33(-5) 1.33(-5) flo-99 67D - 4.35(-7) 4.35(-7) Ru-103 40D 3.36(-9) 1.26(-5) 1.26(-5) Ru-106 lY 4.23(-9) 2.89(-5) 2.89(-5) Rh-106 2. 2H 4. 23 (- 9 ) 2.89(-5) 2.89(-5) Ag-111 7. 5D 1.27(-8) 1.27(-8) St-1 ?5 2.7Y 7.40(-9) 8.63(-8) 9.37(-8) Te-:27m 109D 3.0C(-9) 1.35(-6) 1.35(-6) Te-127 9.35H 3.00(-9) 1.14(-6) 1.14(-6) Te-129m 34D 1.20(-8) 4.05(-6) 4.06(-6) Te-129 70M 1.20(-8) 4.05(-6) 4.06(-6) Te-132 78H 6.4)(-9) 2. 91 (-6 ) 2.91(-6) I-131 8.lD 3. 23 (- 6 ) 7.99(-6) 1.12(-5) I-132 2.3H 6.09(-7) 7.12(-6) 7.73(-6) Cs-134 2.lD 3.40(-8) 7.21(-7) 7.55(-7) Cs-136 13D 1.96(-7) 1.92(-6) 2.11 (-6 ) Cs-137 30Y 1. 30(-6) 1.54(-5) 1.i7(-5) Ba-140 12.80 5.86(-10) 2.55(-6) 2.55(-6) La-140 40H 5.86(-10) 2.55(-6) 2.55(-6) Ce-141 32.5D 1.55(-9) 4.24(-7) 4.26(-7) Ce-143 33.7D 5.08(-10) 2.28(-7) 2.28(-7) Pr-143 13.70 5.08(-10) 2.28(-7) 2.28(-7) Ce-144 e5D G.90(-10) 3.11(-7) 3.11(-7) Pr-144 17t1 6.90(-10) 3.11(-7) 3.11(-7) Nd-147 ll.lD 2.97(-10) 9. 58(-8 ) 9.60(-8) Pm-147 2.7D 3.93(-10) 1.74(-7) 1.74(-7) Eu-155 1.8Y - 1.80(-8) 1.80(-8) Ta-182 115D - 9.97(-7) 9.97(-7) Pu-238 86Y 2.46(-10) 3.86(-9) 4.11(-9) Pu-239 2.0(4)Y 6.54(-11) 1.03(-9) 1.09(-9) Pu-240 6.7(3)Y 8.54(-11) 1.34(-9) 1.42(-9) Pu-241 13Y 7.0S(-9) 1.14(-7) 1.21(-7) 49
! Pu-242 3.8(5)Y 1.81(-13) ,
.89(;12) 3.07(-12)
D LJ 11.2-12 Amend. 49 Apr. 1979
TABLE 11.2-1 (Continued) Low Level (2) Intermediate (3) Total Isotope Hal f-li fe(S ) Activity (Ci) Level Activity (Ci) Activity (Ci) Np238 2D 2. 77 (-15.) 4.23(-14) Np239 2.4D 1.27(-11 2.05(-10) 4.51(-14{ 2.17(-10 Am- 241 433Y 2.52(-11) ) 4.01(-l2) 4.26(-10) Am-242m 152Y 9.95(-13) 1.58(-11) 1.67(-11) Am-242 16H 9.95(-13) 1.58(-11) 1.67(-11) Am-243 7.4(3)Y 4.07(-13) 1.85(-11) 1.89(-11) Cm-242 163D 1.77(-11) 2.84(-10) 3.02(-10) Cm-243 30Y 2.45(-15) 3.89(-12) 3.89(-12) Cm-244 18Y 5.12(-l2) 8.14(-11) 8.65(-11) 49 (1) 1.0% failad fuel for fission products and 100 ppb Pu in the primary sodium. 30 year irradiations and 10 day decay for fission and 49l attivated corrosion products. (2) Total discharge of s10-4 uCi/cc at 850 gallons per day. (3) 10$ of the annual stored inventory is released. (4) 5 1.11 exceptreleased activity for iodine (DF=10"hag)been decontaminated and tritium (DF=1). by a factor of 10 (5) Y= years, D= days, H= hours, M= minutes. (6) The tritiug values de not include Balance of Plant tritium release of 50 8.45 x 10 - Ci/ day. See Section 11.2.6.2. (9
- 2.86(-3) = 2.86 x 10-3 a n n ' r.
L O' O L U Amend. 50 11.2-13 June 1979
DOL vel \T PAGE _
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J
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A 'I ha l 79 07 02 019 l NO. OF PAGES A REAS,0N: C PAGE ILLEGIBLE: C HARD COPY FILED AT: PDR CF OTHER C BETTER COPY REQUESTED ON / / O PAGE TOO LARGE TO FILM: O HARD COPY FILED AT: PDR CF OTHER C FILM 2D ON APERTURE CARD NO. 268 2,;;
11.3 GASEOUS WASTE SYSTEM 11.3.1 Design Base The design objective of the Gaseous Waste System is that the levels of radioactive materials in the plant effluents to the environment shall be kept as low as reasonably achievable. Extensive efforts shall be directed toward the development of system designs that will result in minimizing or eliminating planned releases of radioactive material to the environment during normal plant operation. Plant design objectives include conformance with the requirements of 10 CFR 20. The design base and the expected values of the annual activity release for each gaseous radionuclide are listed in Tables 11.3-11 and 50 3 11.3-12, respective'y. The summed annual dose at the site coundary from gaseous radioactive effluent for the design base conditions is approximately 4.5 mrem total skin dose (which includes both a and s), and 50 0.66 mrem whole body gamma dose (see Table 11.3-18). This annual dose is well below the requirements of 10 CFR 20 for unrestrictcd areas. Dose rates within accessible restricted areas are below the requirements of 10 CFR 20. The Radioactive Argon Processing Subsystem (RAPS) shall maintain the primary sodium system cover gas at an acceptable level of radioactivity and shall provide a source of low-radioactivity gas for use in reactor seals. The cover gas is to be contained within the RAPS circuit during its radioactive decontamination and reuse, except for leakages and cover gas samples taken for failed-fuel and impurity monitoring. The Cell Atmosphere Processing Subsystem (CAPS) shall process plant effluents that contain or might potentially contain radioactivity, in order to reduce the radioactivity to levels that are as low as reasonably achievable during the normal range of routine plant operations. During off-normal operations, the CAPS function is to continue to prevent, to as great an extent as is practicable, the release of radioactivity. 11.3.2 System Description 11.3.2.1 Process Flow The origin, flow paths, and release points of the gaseous radioactive waste system are shown schematically in Figure 11.3-1. The radioactive gases generated in the reactor (see Sections 11.3.2.2 and 11.3.2.3 for composition, flow rates, and concentrations) consist of tritium and noble gas isotopes. The latter, and some of the tritium, migrate to the reactor and Primar; Heat Transport System (PHTS) cover-gas spaces. The Radioactive Argon Processing Subsystem is an essentially closed internal system which continuously processes the cover gas to reduce its activity and then returns the " recycle" argon to the seals and cover-gas spaces. An expanded schematic diagram of the RAPS is 49 lshown in Figure 11.3-2. 11.3-1 Amend. 50 n n*9. June 1979
.C0 L4-
No significant quantities of iodine nor particulate forms of radioactive isotopes, excluding those daughter products associated with noble gas decay, are expected to be present in the Gaseous Radwaste System. Although some vaporization of nongaseous isotopes from the liquid sodium into the reactor cover gas may occur, all cover gas entering the system is processed through two vapor traps, which are expected to remove essentially all nongaseous isotopes, including trace quantities of sodium iodide. Continuous radiation monitoring of the gases is provided by the process monitoring of RAPS and CAPS and monito -ing of the CAPS exhaust. Radioattive gas leakages into the inerted cells of the Reactor Containment Building (RCB) and Reactor Service Building (RSB) are collected and processed through the CAPS before release to the environment. An expanded schematic diagram of the CAPS is shown in Figure 11.3-3. 50 l M st (99.85) of the tritium generated forms a hydride in the sodium; it is then partially removed from solution in the sodium by cold trapping. A very small portion diffuses into the cells of the Intermediate Bay of the Steam Generator Building. A detailed description of all the identifiable leakage and discharge paths is given in the following paragraph. In Figure 11.3-1, certain paths have been assigned " numbers" that correspond with the following discussion: Pa th la. Reactor cover gas is conservatively estimated to diffuse through the reactor head seals at the rate of 0.012 scc / min. g This leakage diffuses into the head access area and is discharged g to the atmosphere through the RCB heating and ventilating exhaust duct. Path lb. The buffered head seals are expected to leak (to the head access area) a maximum of 7 scc / min of recycle argon cover gas. Pa th 2. Although the cover gas lines connected to the reactor and other components in the Primary Heat Transport System are not expected to leak, a leakage of I scc / min has been assumed, for the purpose of conservatism, in the design basis evaluation. Also, tritium dissolved in the sodium in this system will diffuse through the hot pipe walls into the RCB cells. These two leakages are considered to diffuse into the RCB cell atmospheres, which are collected and processed by CAPS and are discharged to the CAPS heating and ventilation exhaust. Path 3. Although the RAPS and CAPS piping and components are also not expected to leak, a leakage equivalent to 1 scc / min of RAPS Cold Box influent gas has been assumed, for the purposes of conservatism, in the design basis evaluation. This assumed leakage is considered to diffuse into the RSB cell atmosphere , 49 which are processed by CAPS and are discharged to the CAPS heating and ventilation exhaust.
,n , , _
dOO L9s Amend. 50 June 1979 11.3-2
Path 4. Ar 39 and Kr collected and stored in RAPS (see further below) are bled into CAPS; these also are discharged to the CAPS
" eating and ventilation exhaust.
Path 5. Tl.t .a iled Fuel Monitoring System discharges reactor cover gas samples to CAPS. After processing, this gas also is discharged to the CAPS heating and ventilation exhaust. Pa th 6. Other plant systems, specifically Refueling, Maintenance and 50l Auxiliary Liquid Metal intermittently discharge radioactive or potentially radioactive gases through CAPS to the CAPS heating and ventilation exhaust. Path 7. Tritium dissolved in the sodium of the PHTS will transfer to the Intermediate Heat Transport System (IHTS) sodium by diffusing through the intermediate heat exchanger (IHX) tube _lls. A very small but finite amount will then diffuse through the hot leg piping in the cells of the intermediate bay (IB) and steam generator bay of the Steam Generator Building and will mix with the ventilation streams in that building. Radioactive gases are thus released to the Heating and Ventilation Systems of the IB, t" RCB, and the RSB (Paths 7, 8, and 9 on Figure 11.3-1). The discharge of these streams to the environment is discussed in Section 11.3.2.6. Balance of Plant (BOP) tritium release (Path 10, Figure 11.3-
- 1) is discussed in Section 11.3.6.2.
A schematic diagram of the process flow in the cover-gas recycle system, which includes the reactor, the overflow vessel, and the PHTS pump cover-gas spaces, the oil traps for the pumps, the Failed Fuel Monitoring System, the recycle argon vessels, and RAPS equipment, is shown in Figure 11.3-2. The recycle system components, distinguished by solid-line blocks, constitute the collection, control, and principal processing portion of the system, although isotooe decay occurs in all parts of the system. The function of this system is to continuously draw radioact.ve gases from the cover-gas spaces, so that noble gas isotopes, both stable and radioactive, are extracted from the cover-gas spaces by condensation in a cryostill, and then to return the purified argon to the cover-gas spaces as a " recycle" argon purge. The activity in the cover gas is thus dependent on the production rate in the reactor, the purge rate, the holdup time, the hal f-life of each isotope, and the cryostill ef ficiency. Argon flows from the recycle vessels nominally at 5.15 scfm: 0.5 scfm to each of the PHTS pumps and 3.65 scfm to the reactor cover gas space. The PHTS pumps gas effluent is divided equally (by design), so that 0.75 scfm (total) passes through the three shaft seal spaces and the three oil traps and enters the RAPS input (vacuum vessel); the other 0.75 49 scfm bleeds to the common pressure-equalization line that joins the reactor, Amend. 50 [, d ,3 7 j June 1979 11.3-3 Lr 4
, the reactor overflow vessel, and the PHTS pumps' cover gas spaces. From this pressure-equalization line, 1.0 scfm of the gas passes through a sodium vapor trap and through the Failed Fuel Monitoring System before entering the RAPS input; the remaining 3.4 scfm goes through the overflow vessel cover gas space, then through a sodium vapor trap to RAPS.
RAPS continuously processes a flow of 10.0 scfm which is made up of the 5.15 scfm input and 4.85 scfm of recirculated throughput. The output of RAPS, 5.15 scfm, is delivered to the ret ycle argon vessels. The RAPS cryogenic distillation column operates with liquid argon as the still bottom. This liquid absorbs the krypton and xenon isotopes and permits their separation by draining, evaporating, and transferring them to the noble gas storage vessel. The transfer to the noble gas storage vessel is to be an annual procedure. During the subsequent year, the transferred gas will be bled at a controlled, low rate from the noble gas storage vessel into CAPS, and through CAPS to the RSB CAPS H&V exhaust. The release process will occur over a period of several months. The Cell Atmosphere Processing Subsystem process flow circuit is shown in detail in Figure 11.3-3. The individual inputs to CAPS, grouped as showr. in the five upper boxes in the diagram, are as follows: I) Cells and Pipeways - During normal operation, there will be a small but finite diffusion and leakage of radioactive g ses through the piping and components. This source of SGI activity will be accumulated in the atmospheros of respec-tive RCB and RSB cells and when the cells are purged to CAPS, the contained radioactivity will be collected and processed in CAPS. It is conservatively assumed for calculational purposes that an average 1 scc / min of reactor cover gas and 1 scc / min of RAPS cold-box process gas will be leaked into the cells; further, they will be 50l exhausted to CAPS without delay, except that Ne23 is assigned a delay of 8 minutes. The cell atmosphere sampling and analysis units will divert to CAPS for processing any activity concentration exceeding a pre-determined setpoint value. For other than RCB cells, normal cell-atmosphere nitrogen, when it is not radio-active, passes directly to the heating and ventilation exhaust system, bypassing CAPS.
- 2) Mass Spectrometer - This equipment, part of the Failed Fuel Monitoring System, periodically sarrples reactor cover gas and discharges portions of the samples into 49 CAPS.
Amend. 50 June 1979 11.3-4
,9 m ..
Cb L 't b
- 3) Gas Services Exhausts - Intermittently, CAPS will receive exhausted nitrogen, argon, or air from vessel cover gases, cooling gases, cleaning, bagging, and fuel handling opera-tion, and other services. These are only infrequent, potential carriers of radioactivity; they will not normally contribute a significant amount of radioactivity relative to the first three sources.
- 4) RAPS Cold Box - lhese CAPS inputs include RAPS cold box components overpressure relief, purgt of RAPS for componen!
maintenance, RAPS cryostill nitrogen coolant effluent and 50 the noble gas bleed from the noble gas storage vessel. The noble gas bleed is normally continuous; the others will be used only in case of a malfunction in the RAPS circuit, and only for short periods of time, as for a repair or correction. 50l With the exception of the noble gas bleed, these sources will not normally contribute a significant amount of radio-activity relative to sources (1), (2), and (3). The nominal volume input of gases to CAPS :s the tine-averaged sum of the inputs listed. The CAPS design flow rate is 38 sctm-A recirculation loop, shown by broken-line in Figure 11.3-3, will return the CAPS output to the vacuum vessel if radioactivity above an acceptable level is detected by the effluent monitoring system. The tritium-water removal process uses an oxidizer and a freeze-out dryer; it oxidizes tritium, collects the resultant tritiated watcr, and passes it to the Radioactive Liquid Waste System, where it is incor-porated into a solid form for off-site disposal. CAPS incorporates two cryogenically-coo'ed charcoal delay beds and a tritium-water removal unit. In the beds, the short-lived gaseous radioactive species are adsorbed and then decay; they are thus removed from the process gas stream. RAPS and CAPS have different process methods, i.e. , the distillation-process removal of noble gases in RAPS rather than the delay beds in CAPS, and the oxidation-process removal of tritium in CAPS. In each subsystem, however, the input is collected in a vacuum vessel, from which it is transferred anu stored under pressure in a surge vessel. It is then treated in the respective cold box. The recirculation-loop feature of these cryogenic subsystems increases their effective capacity to reduce activity In RAPS, it also permits maintaining a 49 steady throughput under conditions of changing input. @ Amend. 50 11.3-5 June 1979 n n ,. lo0 ~tG -
11.3.2.2 Gaseous Radioactive Waste Inputs to System The radioactive waste gases consist of noble gas radionuclides and tritium that are generated by fission and/or neutron activation. The noble gas radionuclides migrate to the reactor cover-gas space, although a time lag occurs in the leakage from ruptured fuel and in the movenent to the cover gas. The tritium remains primarily in the sodium, from which it is removed by cold trapping. The tritium concen-tr ation in the cover gas will be affected by the sodium temperature, the cover gas temperature and pressure, the cold trapp"ng efficiency, and the concentration of hydrocen in the sodium- The latter factor, in turn, depends on the diffusion rate of hydrogen from the steam-generator tubes into the intermediate sodium and the subsequent di ffusion of the hydrogen into the nrimary sodium system through the IHX tube walls. Table 11.3-1 lists the radionuclides of concern, their half-lives, decay constants, and design base input rates to the cover-cas space at normal reactor power level (975 tB!t). The noble gas input rates to the cover gas are adjusted for decay during their release from failed fuel (modeling described in Section 11.1). The assumed condition of li failed fuel is the de>ign base point for the radioactive gas pro-cessing systems. 11.3.2.3 Activity Inventories and Concentrations
- a. Reactor Cover-Gas Space - The steady-state inventory of a specific radionuclide in the bulk volume of the reactor cover gas can be calculated from the following formula:
i I=
+ FW *(l)
WhereI=inveptory(Ci),i=inputrate(Ci/nin),A= decay constant (min ) (0.693 ' half-life),T = processing efficiency i-factor (typically taken as unity), F = purge rate (3.65 scfm), and V = cover-gas m ce volume (410 scf). F/V is the " purge factor" The concentration of a radionuclide in the cover-cas spaceisitsinventorydivfdedbythetotalgasvolumeadjusted to standard conditions (68 F,14.7 psia). Table 11. 3-2 lists, for each isotooe of concern, the in-ventory concentration in the reactor cover gas for the design-base condition of 15 failed fuel,
- b. RAPS Process Stream - Tabie 11.3-3 lists the inventories in the principal RAPS vessels, and Table 11.3-4 lists the concentrations 49 of activity at selected points in the RAPS process stream. The
#cend. 49 April 1979 11.3-6 A 0 l? k ,~'
11.3.2.6 Radioactive Gaseous Site Boundary and Restricted Area Concentrations Radioat tive gaseous concentrations at site boundary have been calculated for five Heating and Ventilating System air stream sources; these are conpared to 10 CFR 20 unrestricted area MPC limits. The air streams are:
- a. Reactor Containnf nt Building Vent - Cover gas diffusion through the reactor head seals and recycle argon gas leakage through the buffered seals mix with the H&V air stream and are exhausted through the main RCB exhaust duct to the exhaust opening located on the top of the Confinement Building. (The release point is
#5A on Table 11.3-20 and on Figure 11.3-9.) The flow rate through the exhaust is 14,000 cfm. Associated activity concent. tions are shown in Table 11.3-13.
- b. Reactor Confinement Building Vent - The Annulus Air Cooling System is provided as a means to mitigate events beyond the design basis.
Activity will only be discharged through relaase point *13, lo-cated at the top of the Reactor Confinement Building, in the event of very low probability accidents beyond the design basis. Thermal Margins Beyond the Design Basis are discussea further in Reference 10 of PSAR Section 1.6.
- c. CAPS H&V Exhaust - CAPS effluent discharges into th? CAPS H&V ducting and is released to the environment through a missile protected exhaust structure located on the roof of the RSB.
The air stream flow rate is 3000 cfm, and the release is within the restricted area (release paint 6 ). Associated activity concentrations are shown in Table 11.3-14.
- d. Intermediate Bay Vent - The exhaust duct located in the IB re-ceives ventilation exhaust air from the Intermediate Bay area.
The flow rate is 64,000 cfm with release within the restricted area (release point dl). The associated activity concentration i s s hown i n Ta bl e ll . 3-16a .
- e. Turbine Generator Building Vent - BOP Tritium discharge will be released to the environment from the TG Building H&V vent (release point d7).
49 The flow rate is indicated in Table 11.3.20. Associa-ted tritium concentration in the release is given in Table 11.3-16. Amend. 49 April l! '9 11.3-9 9 n - . :.
- Do na
! The two restricted-area locations that present potential occu-pational exposure to airborne radionuclides are the IHTS piping cells and the head access area. The estimated leakage of tritium into the IHTS piping cells is 1.6 x 10 -4 C i /c'ay. This is diluted in 1000 cfm of ventilgting air, resulting in an exuected concentration of 3.9 x 10-g 50 i.Ci/cm3 , less than 0.1" of the MPC (occupational) concentration of 5 x 10 -6 pCi/cm The normally accessible area with the largest potential atmos-perhic radionuclide concentrations is the head access area (HAA). As showninTable11.3-13,theconcentrationsinthisyegicafgroperation with 1 failed fuel will be approximately 1.3 x 10- pCi/cm , which 50 results in a sum of the fractianal MPC's (occupational) of 0.05. Also shown on Table 11.3-13 are the expected concentrations for operation with 0.1: failed fuel, which results in a sum of the fractional MPC l
(occupational) value of 0.00/. Particulates are not expected to be discharged from the design release points of the CRBPP. However, as discussed in Section 11.4, monitors will be provided, as appropriate, to ensure the capability of detection. 11.3.3 System Design 11.3.3.1 General The RAPS and CAPS Sy cem designs emphasize all-welded con-struction, wherever practicAle, and bellows-sealed valves throughout, so that leak-tightness is enhanced. There will be no field-routed piping in eitnei system. 11.3.3.2 Equipm ' t_ PAPS and CAPS flow diagrarns are shown in Figares 11.3-4 through 11.3-7. The design parameters of the major equipment ccmponents shown in the diagrams are listed in Table 11.3-17; this table sunmarizes the design codes, the seismic categories, the operating pressures and tempera-tures, the actual volumes of the components, and their capacities t.ader normal operating conditions. Experience with components is limited. Some testing has been performed as part of the development of FFTF. These components are all located within the RCB and RSB, which are Seismic Category I structures, and are tornado-protected by the building. Consequences of equipment failures by rupture of leakage are 49 discussed in Section 15.7. 11.3-10 Amend. 50 June 1979 2bb 2h)
11.3.3.3 Ins trumer,ta tion Process instrumentation is to be installed in RAPS, CAPS, and the inert gas distribution systems in order to effect the control, gene ally in the automatic mode, of pressures, temperatures, radio-activity concentrations, and flow rates (see Section 9.5.5). The normal pressures and temperatures for the vessels are listed in Table 11.3-17.
"g
- Radioactivity concentrations and flow rates have been discussed in previous sections of Section 11.3.
Radiation monitoring for the nitrogen-inerted cells in the RCB and RSB is provided by two separate multi-channel sampling and analysis units, typically piped as shown in Figure 11.3-8. The individual cell atmospheres are continuously withdrawn but are sequentially subjected to analysis for detection of radioactivity, water vapor, and oxygen. Detection of oxygen concentration in excess of the high set point will automatically initiate purging of the violated cell with fresh nitrogen to reduce the cacentration to the low set point. Initiation for automatic purging to reduce water vapor concentration rather than oxygen is an operator option, selectable by a hand switch. Detectio' of radioactivity in the cell atmosphere automatically directs the cell effluent to vent to CAPS if the radioactivity is above the set point; otherwise, it is vented to HVAC for direct release. This option is provided for all the inerted cells in the RSB, but not for the inerted cells in the RCB. The effluents from the RCB cells are always vented to CAPS during normal plant operation. Two radiation monitors in series are provided in a common RSB inerted-cell-vent header before the cell gases are Mscharged into the HVAC ducting. This provides continuous monitarir.g of the vented gases, and a high-radiation signal provid- automeiic. closure of a common header-isolation valve located downstream of the radiation monitor. The signal also closes all the HVAC vent valves to the individual cells, to prevent release of radioactive gases. 50 The question of whether or not available RSB radiation-monitoring equipment provides adequate discrimination to guard against excessive releases, is addressed in the following sample calculation: If a 100,000 ft of radioactivity detectioncellisat,butnog)above,thethreshold (1 E-6 Ci/cm and is then purged within one day to correct its oxygen concentration, the purge flow for other than cells in the RCB will enter tge H&V effluent duct. 3Undgrtheseconditjons,anominal(lE+5ft)(2.832 E+4 cm / f t ) (lE-6 Ci/cm ), or 0.0028 Ci, will be in the sell. In the worst case, all of it (0.0028 Ci/dari could be released in the H&V effluent. This is only 1.1% of the normally 49 expected daily plant release rate. 11.3-11 M ] j: ,,u Amend. 50 June 1979
11.3.4 Operating Procedures and Performance Tests The gas inputs to CAPS (listed in Section 11.3.2.1) are drawn into a vacuum tank by one or more of four 25-scfm compressors, depending on input flow rate, which are instrumented to automatically maintain a 7.7 to 12.7 psia pressure in the tank. The compressors are arranged in parallei and their controls are such that one starts when the vacuum pressure reaches 12.7 psia, and others start in sequence if the pressure 49 is not held below 12.7 psia. All compressors stop when the vacuum O Amend. 50 ll.3-l'a June 1979 9 268 251
reaches 7.7 psia in the vacuum tank. If the setpoint vacuum pressure exceeds 13.7 psia, a high alarm is triggered. If the temperature of the effluent from the compressors exceeds the high setpoint, indicating inadequate cooling, can low alarm will be triggered to alert the opera-50 tor to the abnormal condition. RAPS and CAPS are independently operated, with process control being automatic and with control-room as well as local provisions for overriding automatic controls if conditions so dictate. Both subsystems have control and alann instrumentation and instant data retrieval avail-able in the control room; this provides the operator with information that ensures proper system operation. The effectiveness of operating procedures has been demonstrated as part of the FFTF development and, to a limited extent, analogous systems used in licht-water reactors. The receiver (surge vessel) of the CAPS compressor (s) normally operates at about 40 psig but can be operated in the range of 35 to 135 psig. The outlet flow from the surge vessel is regulated by a flow control valve. When the surge vessel pressure reaches the nominal 40 psig, the flow valve will permit gas to flow to the processing equipment at a flow rate that increases linearly with surge-vessel pressure differ-ential above 40 psig. If there is a high gas inflow to the surge vessel and the pressure rises above the nominal setpoint, the outflow from the vessel is increased to accommodate to the increased inflow rate. If the inflow exceeds the maximum processing capability, the surge vessel will act as an accumulator and its pressure can increase to 135 psig, at which pressure the compressors are automatically shut down and a high alarm is triggered. In the event of a compressor diaphragm failure, the failed compressor can be isolated and repairs can be made without shutting down the remainder of CAPS. Similarly, individual radioactive gas filters, upstream of each compressor, can be isolated and replaced when a high pressure-drop alarm is triggered f t om excessive particulate buildup. One of two parallel tritium-water removal trains is always on line while the other is being regenerated, with the switchover being automatically controlled by a sequencing timer. In the regeneration cycle, any CO, ghich hag been frozen out of the process gas is sublimed off between -20 F and 0 F and is released through the CAPS effluent to Then, ice formed from tritiated water vapcr is melted and, 9 H&f. 40 F and 70 F, it is drained to the CAPS rad-water holding vessel. this vessel, it is periodically transferred to the Radioactive Waste System by manual actuation of the transfer valve. The waste is then 49 converted to non-compactible solid form for ultimate offsite disposal. 11.3-12 Amend. 50 June 1979 779 2d h> c r
The RAPS design is based on operation with 1% failed fuel. Nomal operation and the expected releases are based on ooeration with 0.1" failed fuel . The estimated radioactivity release rate from the gaseous waste and the H&V systems after a 1-year period under average operating conditions are shown in Table 11.312. The release rates based on the design condition of 1.0% failed fuel are presented in Table 11.3-11. 11.3.6 Release Points 11.3.6.1 Nuclear Island There are e total of eight design release points for the Nuclear Island Buildings. The location, height, discharge flow rate, discharge velocity, discharge air temperature, and size and shape of tre discharge orifice for each release point are presented in Table 11.3-20. Ventilation from the Steam Generator B ilding Intermediate Bay cells is ducted to a single exhaust point located in the Steam Generator Building Intermediate Bay. (Release Point 1 of Figure 11.3-9). There is a separate exhaust point for each af the Steam Generator Loop cells. Ventilation from each of the three Steam Genera-tor Loop cells is du & . te its respective exhaust point located in the Stean Generator Builaing Auxiliary Bay (release Pcints 2, 3, and 4 of Figure 11.3-9). Levels of radioactivity in these areas will make no significant contricution to offsite dose rates. There a e two design release points provided for the RSB. One design release point exhausts the Radwaste Area. This area in-volves decontamination of non-volatile isotopes and is not expected to result in the release of activity to the exhaust. However, as described in Section 11.4, monitoring of this release point will be provided. An additional release point for the RSB is provided for the exhaust from the CAPS, which is exoected to release activity to the exhaust. Per Section 11.4, a monicor will also be provided for this exhaust. The locations of the RSB CAPS H&V exhaust and Radwaste Area exhaust are Points 5 and 6, respectively, of Ficure 11.3-9. Ventilation from Reactor Containment Building H&V system ar.d from the annulus pressure #iltration system is ducted to a single ex-49 haust point (release point Sa of Figure 11.3-9) located on the top of the Reactor Confinement Building. Amend. 49 April 1979 11.3-15 268 2L
Two release points (points 20 and 21 in Figure 11.3-9) associated with Thermal Margins Beyond the Design Basis design featurn receive exhaust from the Annulus Air Cooling System and the Containment Cleanup System (These systems are described in Section 9.6.2). These systens are not required to operate during normal operations or to mitigate the consequences of any accidents in the Desiqn Basis. Activity would only be released from these points in the event of very low probability accidents beyond the design basis, och as a hypothetical core disrup-tive accident. The Containment Cleanup System exhausts throuoh a release point (21) near the top of the Reactor Confinement Puilding. Before being exhausted to the atmosphere, the Containment reaction products pass through one of two filter trains, which consist of an air washer, a sodium scrubber and water separator, a heater, a prefilter, a high efficiency particulate air filter (HEPA), an adsorber bed, and an af ter-HEPA fil ter. Particulates, radioiodines, radiogases, and plutonium are monitored continuously in the effluent stream. 11.3.6.2 Balance of Plant _ A small fraction of tritium produced in the fuel and control rods passos into the steam-water system by diffusion throuah stainless steel in the IXH and through chronallov in the steam generators. Tritium is expected to be in the steam-water sy' stem in the form of tritiated water. The condenser air removal system removes non-condensible gases (vapors) from the condensing steam. Tritiated water vapor, present in the off-gas flow, constitutes the only expected qaseous release contri-bution from the balance of plant. Mechanical vacuum pun.r s will remove the vapors together with the non-condensible ga os and will discharge then to the exhaust plenum of the Turbine Generator Building (exhaust point 7 on Figure 11.3-9). The vapors will mix with the exhaust air. The resulting caseous tri-tium release from the TGB is provided in Table 11.3-16. B0P tritium contribution is included in the dose calculations presented in Section li.3-8. Balance of Plant tritium release is based on the following assumptions: (1) Plan' Capacity Factor of 0.68, (2) Vacuum Pump Operating Factor of 0.85, (3) Radioactivity Input to Steam-Water System 0.016 Ci/ day, and (4) Condenser off-gas removal 7 scfm. The design value release of tritiated water vapor amounts to 6.3 x 50 10-5 Cifday. 49 Description, design bases, and evaluation of the BOP design are provided in Section 10. 11'3-16 Amend. 50 June 1979 mm c;a COO
Average effective grazing area of beef cattle and dairy cows is assumed to be 45 m' Assuming that 100% of the tritium ingested is evenly distributed within the meat and the average weight of a steer is 500 kg, the cumulative fraction of tritium transferred per kg of beef is 1.0/500 kg or 0.002/kg. Althocqh some grains produced outside of the immediate area are used to supplement their diet, dairy cows on the farm nearest the CRBRP site are allowed to graze outside all year. It is therefore assumed that 1005 of the diet of both cows and cattle comes from the fields. All other variables, such as total intake of beef or milk and elapsed time between butchering and ingestion, are provided in Table ll .3A-6 to the Appendix of Section 11.3. Maximum total annual whole body internal dose from exposure to gaseous effluents to the hypothetical individual who eats only leafy vegetables grown in the closest home garden to the site, eats only the meat from beef cattle grazing in the closest field to the site, drinks one liter of milk per day taken only from the closest known cow to the site, drinks water from the nearest reservoir, and lives in the closest 50 ! house to the site, is 0.021 mrem /yr. The estimated annual internal dose from natural radiation to an individual is 18 mrem /yr (ref. 2). There-fore, the maximum internal dose to an individual from exposure to "BRP 50l gaseous effluents is approximately 0.12 percent of his internal e from naturally ocgurring radiatio.1. The maximum internal dose - o 50 1 f ctor of 4 x 10 ~ of the 10 CFR 20 annual dose limits for unres :r c r od areas. External and internal doses resulting from exposure to daughter products of gaseous effluents have been included in the dose evaluations. Analysis of doses due to gaseous fallout on the Clinch River assume'., an annual mean depth of a reservoir from which water is taken for diinking supplies downstream of the plant site to be 4.8 m. Approxi-50l mately 0.5 days are required for processing the Clinch River water into potable water. Internal whole body doses to persons drinking water is calculatgdwiththea3sumptionthattotalintakeduring24hoursis 2,200 cm and 1100 cm for an 8 hour working day (Ref. 3). Period of exposure is assumed to be 260 days for workers and 365 days for public consumption of water. A conservative dilutio., factor for gaseous deposi-tion into the river, due to flow of fresh watei into a reservoir, is derived as flow oJtifloW in. An additional dilution factor (0.16) results from the ratio of the river flow relative to the deposition velocity of the gaseous release. Maximum whole body internal dose to an 50l individual from ingestion of water is 2.0 x 10 -4 mrem /yr from the gaseous fallout. The analysis conservatively assumes, on an annual basis, one month of zero flow condition of the Clinch River at the plant site and 49 11 months of summer average flow (4777CFS). 11.3-18a 0 2)b Amend. 50 June 197
Population dose assessments for the area within the fif ty mile radius of the plant site have been performed. The pathways assessed considered all those previously discussed in tnis section. Population dose assessments via the ingestion of milk containing fallout utilized the inpact for milk cow populations from the counties within a 50 mile radius of the CRBRP as provided in Ref. 4. The same gereral approach was applied to the population dose associated with the ingestion of beef containing fallout, i.e., a survey of the beef cattle (Ref. 4) was incorporated. The population dose associated with the ingestion of leafy vegetables utilized an extrapolation of nearby land usage from the plcnt site to the 50 mile radius area. Ingestien of water containing fallout effluent considered nearby reservoirs. Inhalation dose assessments were also included. The man-rem internal dose values from all the 50 pathways considered sum to a value of 0.017 man-rem per year. The assessments t.tilized the model specifics of section ll.3A and 50 mile 49 radius population figures associated with the year 2010. O Amend. 50 June 1979 qr . ( 0 (q) { ' e ?.
REFERENCES
- 1. U.S. Department of Defense, The Effects of Nuclear Weapons, (Glassten S., Editor), Revised Edition, USAEC, Washington, D.C.,
1962.
- 2. Effects on Population of Exposure to Low Levels of Ionizing Radia-tion, Report of the Advisory Committee on the Biological Effects of Ionizing Radiations, National Academy of Sciences, National Research Council, Washington, D.C. 20006, November, 1972.
- 3. Recommendations of the International Commission on Radiological Protection, Report of Committee II on Permissible Dose for Internal Radiation, ICRP Publication 2, Pergamon Press, New York, New York, 1959.
49l 4. Tennessee Statistical Abstract - 1971 Edited by M. G. Currence, Center for Business and Economic Research, University of Tennessee, 50 Knoxville, Tennessee. Amend. 50 ll.3-18c June 1979 b O 25,'
TABLE 11.3-1 RADIONUCLIDE IriPUT RATES TO REACTOR COVER GAS
- Decay Constant input Rate Isotope Hal f-L i f (day ~) (Ci/ day) 131m 4 91 Xe 11.96 day 0.058 112 133m Xe 2.26 day 0.306 3,760 1
Xe 5.27 day 0.131 65,100 135m Xe 15.7 min 63.6 9E,600 ll5 334,000 4g Xe 9.16 hr 1.81 133 170,000 Xe 14.2 min 70.2 83m Kr 1.86 hr 8.98 16,400 85m Kr 4.4 hr 3.78 30,000 85 Kr 10.76 years 1.77E-4 2.05 87 Kr 76 min 13.1 52,000 8F' Kr 2.79 hr 5.96 64,400 Ar 39 269 years 7.0E-6 0.129 41 Ar ^ 110 min 9.07 3 14 23 Ne 38 sec 1576 1.42E+9 50 H 3 12.5 years 1.52E_4 1.95E-7*** TOTAL; 832,000 ** 49l
- For the design condition
** Exclusive of ile23 49l ***The rate at which tritium diffuses into the cover gas to rep' ace losses and maintain eouilibrium concentrations at the liquid metal-50 to-gas interface.
Amend. 50 June 1979 11.3-19 nrn 9On EJ0 (O
TABLE 11.3-2 GASE0US RADIONUCLIDE CONCENTRATION IN REACTOR COVER GAS
- Inventory Concentri; ion Isotope (pCi/ scc)
(Ci) l31m 8.6 0.74 Xe l33m 2.8E+2 24. 50 l Xe Xe l33 5.0E+3 4.3E+2 l35m 1.2E+3 1.1E+2 Xe Xe l35 2.3E+4 1.9E+3 Xe l38 2.0E+3 1.8E+2 83m 64 Kr 7.5E+2 85m 1.5E+2 Kr 1.8E+3 85 1.4E-2 Kr 0.lf 07 2.0E+3 1.7E+2 Kr 00 2.9E+2 Kr 3.4E+3 39 0.783** Ar 9.09** 4I 14.4 1.2 Ar 23 7.7E+4 Ne 8.9E+5 3 1.5E-5 49l H 1.7E-4
- For the design condition 49 l ** After 30 years' operation Amend. 50 June 1979 y
2bf3 2bf/
TABLE 11.3-3 ACTIVITY INVENTORIES IN RAPS PROCESS VESSELS RAPS Vacuum RAPS Surge Recycle Argon Vessel Vessel RAPS Cryostill Vessel s Isotope Design Expected Design Expected Design Expected Design Expected (Ci) (Ci) (Ci) (Ci) (Ci) (Ci) (Ci) (Ci) l31m Xe 1.2 0.12 28 2.8 1.9E+3 1.9E+2 1.lE-3 1.lE-4 l33m Xe 38 3.8 8.2E+2 82 1.1E+4 1.lE+3 3.lE-2 3.lE-3 Xe l33 6.9E+2 69 1.6E+4 1.6E+3 4.7E4 5 4.7E+4 0.61 6.lE-2 135m Xe 24 2.4 32 3.2 2.0 0.20 6.6E-5 6.6E-6 l35 Xe 2.5Et3 2.5E+2 4.0E+4 4.0E+3 8.8E44 8.8E+3 1.1 0.11 Xe l38 35 3. 5 44 4.4 z.5 0.25 8.2E-5 8.2E-6 d 83m Kr 50 5.0 3.6E+2 36 1.6E+2 16 3.9E-3 3.9E-4 85m k' Kr 1.7E+2 17 2.0E+3 2.0E+2 2.lE+3 2.lE+2 3.8E-2 3.8E-3 Kr 2.2E-2 2.2E-3 0.52 5.2E-2 7.2E+2 72 2.lE-5 2.lE-6 87 Kr 1.lE+2 11 6.0E+2 60 1.8E+2 18 5.0E-3 5.0E-4 00 Kr 2.7EF2 27 2.5E+3 2.5E+2 1.7E+3 1.7E+2 3.7E-2 3.7E-3 39* Ar 3.5 3.5 81 81 28 28 49 49 4I 7.9 7.9 1.5 1.5 1.3 1.3 Ar 1.1 1.1 23 Ne 17 17 3.97 0.97 3.4E-4 3.4E-4 1.3E-3 1.3E-3 3 H 6.6E-5 6.6E-5 2.4E-3 2.4E-2 3.0E-2 3.0E-3 1.6E-3 1.6E-3 50 49 Total 3.9E+3 4.lE+2 6.2E+4 6.3E+3 5.8E+5 5.8E+4 52 50 g
- After 30 years' operation 5 $g 2
e
'O C s
~
CP O 1-o e e
TABLE 11.3-6 ACTIVITY INVENTORIES IN CAPS PRCCESS VESSELS CAPS Vacuum Vessel CAPS Surge Vessel Isotope i Design Expected Ocsign Excected 131m Xe 3.9E-3 3.8E-3 5.8E-2 5.7E-2 133m Xc 5.0E-3 2.3E-3 7.5E-2 3.5E-2 133 Xe 0.18 0.13 2.7 1.9 135m Xe ;,.8E-2 2.7E-2 0.11 0.11 135 Xe 0.23 3.9E-2 3.2 0.55 138 Xe 1.1E-3 1.5E-4 4.1E-3 5.5E-4 Kr 8' 4.1E-3 7.1E-4 4.5E-2 7.7E-3 85m Kr 1.5E-2 2.6E-3 0.19 3.4E-2
'5m Ki 9.1E-3 4.1E-3 0.14 6.1E-2 Kr 87 8.2E-3 1.4E-3 8.0E-2 1.4E-2 88 Kr 2.4E-2 4.1E-3 0.28 4.9E-2 Ar 39* 1.0E-3 6.7E-4 1.bE-2 1.0E-2 Ar 41 6.1E-4 1.6E-4 6.6E-3 1.7E-3 23 Ne 9.1E-6 9.1E-6 2.0E-6 2.0E-6 50l H 2.9E-5 2.9E-5 4.4E-4 4.4E-4 Total 0.50 0.22 6.9 2.9 49 *After 30 years' operation Amend. 50 June 1979 11.3-24 4 U b [, O ,'
TABLE 11.3-7 ACTIVITY CCNCENTRATIONS IN CAPS pRCCE55 STREAM Influent to Vacuum Effluent from the Vessel Second Charcoal Bed Isotope (uCi/ scc) (uCi/ scc) Design Excected l Cesign l Excected ie 131m 8.7E-4 8.5E-4 3.9E-10 3.8E-10 Xe'3 ' 1.1E-3 5.2E-4 133 *
- Xe 4.0E-2 2.9E-2 135m *
- Xe 7,.E-3 7.3E-3 135 9.CE-3 *
- Xe 5.1E-2 12 *
- Xe"8 3.0E-4 4.0E-5
<r- 9.5E 4 1.5E 4 3.0E-10 5.2E-11 Kr SE' 3.4E-3 5.9E
- 5.4E-5. 9.4E-7
' r85 2.1E-3 9.2E 4 1.6E-3 7.3E-4 37 *
- Kr 1.9E-3 3.3E-4 8 2.4E-7 4.2E-8 Kr 5.4E-3 9.4E-4 Ar 39** 2.4E-4 1.5E-4 1.9E-4 1.2E 4 Ar 1
1.4E-4 3.6E-5 4.2E-5 1.1E-5 23 *
- Ne 1.1E-5 1.1E-5
~ ~ - ~
50 Total 0.12 5.0E-2 1.9E-3 8.7E-4 49 *<<E-10
**After 30 years' operation Amend. 50 June 1979 2- [", -.0 0e, 11.3-25 c O ,_
? GE D I V 6 --.
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TABLE 11.3-12 ANNUAL ACTIVITY RELEASE RATES FOR THE EXPECTED SERVICE CONDITION Main RCB H&V RSB H&V Intermediate Total Radionuclide Exhaust Exhaust Bay Leakage Release (Ci/ year) (Ci/ year) (Ci/ year) (Ci/ year) Xe 31m 4.8E-4 2.7E-4 0 7.5E-4 l33m Xe 1.6E-2 2.9E-31 0 1.6E-2 l33 49 Xe 0.27 1.lE-10 0 0.27 l35m Xe 6.9E-2 + 0 6.9E-2 l35 49] Xe 1.2 + 0 1.2 l38 Xe 0.11 + 0 0.11 83m Kr 4.0E-2 3.7E-5 0 4.0E-2 85m Kr 9.9E-2 0.67 0 0.77 85 Kr 8.9E-6 5.2E+2 0 5.2E+2 07 Kr 0.11 8.5E-8 0 0.11 49 88 l Kr 0.19 3.0E-2 0 0.22 39 Ar 2.9 85 0 88 4 91 Ar 4I 8.4E-2 7.7 0 7.8 l Ne 2.0 + 0 2.0 3 H ++ 5.5E-5 7.lE-3 5.8E-2 6.5E-2 50 49 Totals 6.9 6.lE+2 5.8E-2 6.2E+2
+ E-45
++ BOP Tritium Release (0.9 Ci/yr) from T-G Building Exhaust not included.
Also, allowance for 2 weeks per year bypass of the oxidizer unit (amounts to 2.7E-2 curies of tritium exhausted to the RSB H&V exhaust) 49 is not included. 2bj 2b Amend. 50 11.3-32 June 1979
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- s R o o I 3 5 5 8 m *
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o s l 3 3 3 3 3 3 l l l e e e e e e r r r r r r r e 1 l 8 E5 7 8 9 8 8 8 3 4 2 I 3 t o 0 1 ees e Nhi H I X X X X X X K K K K K A A N H T ** Pd 5 (* i O 9 4 >12 gM 2 k' jmm ;rCL"
TABLE 11.3-16 TRITIUM C0iCEtiTRATIOi AT SITE B0ctiDARY FROM T..G. BUILDIfiG EXHAUST MPC* Concentra tion Concentration (uCi/cc) (pCi/cc) + MPC
~
49
- TABLE l'. 3-16a TRITIUM C0?iCEtiTRATI0!i AT SITE SOUtlDARY FROM IB EXHAUST MPC* Concentra tion Concentration 6tCi/cc)
- (uCi/cc) + MPC 2E-7 1.0E-12 5.0E-6 49 10 CFR 20, Appendix B, Table II (unrestricted area).
4, e, _U' -) O V (. Amend. 49 April 1979 11.3-36
T/af 11.3-17 ESIG" P/P/EEPS T P/PS RD C/ES PROESS \ESSElS a -
"w + **
, Design Nor-al Capacity at a
N Items
!urber Fequired Oesign Ccde Seismic Category Ircssure (psig)
Operating Pressure Mign
'errerature Coerating Terrerature Vol tr e Operating Pressure and Materials of (psig) (Ci ) (cF' (scf) Ter:re ra ture Construction (scf)
Storage Vessels, 2 III-2 I 200 35 250 80 to 120 720 2200 Carbor Steel Recycle Argon (total)
%?S Ccyegenic 1 III-3 I -14.7. 200 32 -320 -282 3.6 58 Stainless Steel Distillaticn r:e t Veasel RAPS Vacuum 1 III-3 I -14.7, 200 -7 to -2 250 120 261 125 to 206 Carbon Steel Vessel FAPS Surge i l-3 I -14.7, 200 103 250 120 500 3600 Carben Steel Ves:.el RA?S Storage 1 III-3 I -14.7, 200 35 250 70 260 CSD Carbon Steel Vessel, t ble Gas CAPS Charcoal 2 III-3 I -14.7. 200 34 -320 -134 64 DNA* Stainless Steel 6l 2ed Vessels (totall CAPS Vacu r. 1 III-3 I -14.7, 200 -7 to -2 250 120 260 124 to 204 Carbon Steel Vessel CAPS Surge 1 III-3 I -14.7. 200 35 to 135 250 70 o 120 693 2360 Carbon Steel 50 49 Vessel 9
*0ces not apply because of adsorption variable
**AS';E Secticn III, Class 3 - III-3
+0esign Code listed nay be higher for reasons other than safety
++Saturnion 7emperature at ncrmal operating pressure c- >
c B b$ N a .' O' w on M OO N O
~
O O O
TABLE 11.3-18 EXTERNAL DOSES
- AT SITE BOUNDARY TO AN INDIVIDUAL VIA GASEOUS EFFLUENTS FRIJM CRBRP DESIGN RELEASE POINTS Total Skin Whole Body Gamma (mrem /yr) (mrem /yr) 4d ' Main RCB H&V Exhaust 0.25 0.13 CAPS RSB H&V Exhaust 4.2 5.3E-1 Intermediate Bay Leakage 4.lE-6 0.
Sfi- T-G Building Exhaust 1.7E-6 0 49 bl Total (mrem /yr) 4.5 0.66
- Design condition 41l 268 260 Amend. 50
@ June 1979 11.3-38
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Figures 11.3-10 thru 11.3-13 have 50 been deleted. 2bb $ r 11.3-49 Amend. 50 June 1979
three gas monitors. Their output will also be provided to the PPS for initiation of contair. ment isolation when a preset radiation level is 6 reached by two of the three detectors. A particulate filter will be in-stalled in line with the gas chamber to prevent buildup in the sample chamber. The monitoring system will be designed to comply with IEEE 279-1971. The overall containment isolation system design and protection logic is discussed in Section 7.3. Figure 12.2-1 shows a typical block 6 diagram of these channels and Figure 7.3-1 shows the trip logic configur-ation. 11.4.2.2.3 Building Ventilation Exhaust Monitors The building exhaust plenums from which potentially radioactive plant gaseous release may emanate are: one in the Intermediate Bay, three 49 in the Steam generator Building, cne in the Plant Service Building, twelve l8 in the Turbine Generator Building, one in the Radwaste Building, Reactor Service Building exhaust (s), the common RCB H&V and Annulus Pressure Maintenance and Filtration System exhaust and the Containment Cleanup System and Annulus Air Cooling System exhaust. Continuous monitoring will be performed at those exhausts which could conceivcbly undergo a signifi-cant increase in detectable levels in radioactivity. The remaining ) exhausts will be sampled periodically. "8 l8 The exhaust plenum located in the IB receives ventilation exhaust air from the Intercediate Bay area. A continuous air monitor (CAM) will be provided to detect particulate, radiciodine and gaseous activity in the effluent stream. The air sample will be obtained isokinetically from the exhaust, on a continuous basis. The operation of the three-channel CAM unit is described in Section 12.2.4.2.1. 18 ieleases, The RSB exhaust (s) will be continuously monitored for radioactivity i 49 1 The three SGB exhausts receive ventilation exhaust from the indi-vidual steam generator cells, and IHS cells in the Intermediate Bay. Eacn exhaust will be sampled for tritium activity using silica-gel dessicants, and analysis of samples will be aerformed by liquid scintilation techniques. The exhaust samples will be obtainM isokinetically, and flow thrcugh the silica-gel column will be maintained constant uf a. regulated pump assembly. 49 The exhaust fans in the TGB receive ventilation from the various Turbine Generator Building operating areas. These exhausts points will also be sampled for tritium activity as described above. 49 TheRoom. tory / Counting exhaust in the PSB receives ventilation from the Hot Labora-6 Samples will be collected isokinetically by a parti-culate (and iodina if required) filter and analyzed for isotopic content @ 268 275 11.4-3 Amend. 50 June 1979
6 l in the Counting Room. The Containment Cleanup System and Annulus Air Cooling System exhausts at the top of the the Reactor Confinement Building. Particulate, radiciodine, gaseous and plutonium activity in the effluent stream will be monitored continuously. The common RCB H&V and Annulus Pressure Maintenance 2 ,. and Filtration System exhaust will be continuously monitored for particulates, radiogases, and radiciodine in the effluent stream. G Amend. 50 ll.4-3a June 1979 2b8 3 h
The reporting of effluent radioactivity r. leased f om the CRBM will be consistent with the guidelines establ led in Regulatory Guide 1.21. This reporting will be based upon the r - alts of Counting Rnom analysis of ef fluent samples obtained at each locati]n listed above. u 11.4.2.2.4 Condenser _ Vacuum Pump Exhaust and Deaerator Continuous 49 Ven t_s J_r i t_i um Sa mp_l e r A continuous gas sample will be .vithdrawn from the condenser vacuum g 4g pump air and dederator exhaust into a silica-gel dessicant column to enable a deternination of tritium activity in order to indicate unacceptable tritium diffusion in the steam generators. The sample will be analyzed using scintillation techniques, 11.4.2.2.5 Cc.itrol Room Inlet Air Monitors The main and remote control room air intakes will be continuously monitored for gaseous radioactivity to determine which intake should be used during the Control Room isolation condition. Details concerning the sequence of operation during Control Room isolation are given in Section 9.6.1.3.4.B. A three channel (particulate /radioiodine/radiogas) CAM will be installed downstream of the parallel (redundant) HVAC nake-up air filter trains to check on the performance of these high-ef ficiency HEPA filter trains. A detailed description of the operation of each of these three CAM units is given in Section 12.2.4.2.1, 6 11.4.2.2.6 Inected Cell Atmosphere Monitors /% $ The capability for monitoring the atmosphere of each individual inerted cell for high radioactivity will be accomplished by three methods. One method is the sequential sampling of groups of cells with three on-line gas monitors as described in 3.A.l.3. Each monitor shall have a trip signal determined by the process system to initiate activation of cell purging equipment. In addition, mobile partict late and gas monitors are provided to sample any individual inerted cells atmosphere, as described in 12.2.4. Finally to provide a sensitive method of sodium leak detection, particu-late monitors are provided for continuous monitoring of selected inerted cells within the RCB containing components contacting radioactive sodium. These moni-tors will alarm for activated sodium present in the cells atmosphere. The indi-vidual irerted cells that are continuously monitored for sodium leak detection will bc listed in the FSAR. 11.4.2.2.7 RAPS and CAPS Monitoring Gas monitors will be provided for the Radioactive Argon Processing System (RAPS) and Cell Atmosphere Processing System (CAPS). These monitors will be located at the inlet to these systems for controlling the rate of radioactivity input. Monitors will also be located at the output of these systems to ascertain that the radionuclide activity of the processed gas is within limits for reuse in RAPS or within 10CFR20 limits for those gases exhausted to the H&V system by CAPS. @ 2bb Arend. 49 11.4_4 Apr. 1979
Table 11.4-1 (Cont'd.) Drocess and Effluent Sampling and Monitoring Sc e r, s E -=: tad P:-fterie; er Cete:*c- . Ci/cc C: :e-tre ces ce C.aa ', Oe::rir!'ca Sidg. E'ev. 53 t1- !>;e D t'vit/ C:se cata: Was o . L - * --* 505' f flunt Activity R:3 931 Centiason .. c:rticaiate /, 'cietili tt:, o ,,,t?37 Cs '!S Orrss crec f Pa:10,-1tre Swee y 'ti
- illy ttm if r-(! ?! !)
f-Scin'ill.it't, if (E5 Fr)
"r]
'"O Cr:ss Cm:
C .ss Cm: N Tv 04 Eftlae** Activity R:8 931 Ccntinuous -7 articulate *
/C -Sti
- illatic, '0 ,W,U 37 Cs) T*O Grc ss (c %.
% :icacjir t -sciatillatics !^ 131 !) in Grcss Cor.c.
I.a n , f-fiatillati;n1Y]'(R;rr) ( '0 Onss Cc-c. T?'503 ("1. er.t FCE 931 Coatinse n
? Nic r. i / activity ,,
Di utoa t e oc -5cietillatic,10 ^ 7(23') pu) IBO Grcss Conc. Niayed 50 l40 1 ee tec tic, tec e m es FD Oy I ed s C 32 N' C R C U D ic 'a CD s.-o O n 'J- N U1 s DQ
. i C'-
Table 11.4-1 (Cont'd.) Process and Effluent Sampling and Monitoring Ccatiruous Erpected Mc9 iter'. ; ce Cetector 6Ci/cc Ccace*.tratio*s or Ouaatity Cescriptic9 Bldg. Elev. Sa ple Tyre 5ersiti.ity Cose Fates Veasu ed Re arks NDS a-d CDs Frecess R58 765' Continuous p !cietillation 10br-CS Sea Sectic, Raitoring Cross Conc. Citeous 11.3.2.3 3 Pfar.t Cischerge Efflueet PSS . Sa ple Libid C0rposit? Sce %ction : ctre,1c C:ac. Sa r'a 3-slyzed in dcwnstrea- of low A Activity System 11.2.5 c: etie; r m using E trc :eticr.al and w li v d s.intil12 tion ccu-:ers ard V* s;cctrcsc3;y systcm. Irerted Cell A bcs-
- prere l'.onitors 6 %e.,st,oispcs.i .
Lgrs3 f.C3 reettc* Centain cnt ?uilding RSE - Peact7r Sarvice Puildir.g lB - Ir.t? :diste Bay 25 - Tu-bir$ C,. erater Ea11 ding C93 - Ccoteol Dcom Euildir.g
!r,for t.t'en will te provided in the F5AR dw 6lI V h m
- a. 3 b
M
@M %
NO N O w
- n. :.
,,,,.5 O
e O O
TABLE ll .5-1A (CONT'D) 46 14* Using remote tools from the Maintenance Platform, place and ooerate automatic welding equipment to cap the cold trap c rys tali.7er inlet and outlet sodium pipes. 46 l 15. Shut off NaK coolant flow to the cold trap crystallizer and drain the tbK to the lowest drain point on the cryctalizer. 46 l 16. Using renote tools from the Maintenance Platform, cut and cao the NaK coolant lines and cocolete preparation for removal of the crystalizer. 46 l 17 Close Floor Valve and remove Maintenance Platform. 46 1 18. Install Cold Trap Removal Cask to the Floor Valve. 46 l 19. Coen Floor Valve and remove Clod Trao crystalizer into Cold Trap Removal Cask. 46 l 20. Close Floor Valve and remove cask from Floor Valve containing cold trap crystalizer.
+To be preceded by appropriate decay time.
*This step can possibly be accomplished following Step 14., depending on pipe accessibility.
1 n
/ OO ._ ; v Amend. 50 June 1979 12-iv
LIST OF TABLES Continued fable No. Page 50 12.1-29 Fuel Handling Cell Argon Circulation System Source 12.1-60 Terms 12.1-30 Control Room Shield Design Source Term 12.1-61 12.1- 31 Isotopic Inventory Released to Environment 12 l-62 Release Includes Bypass Leakage and Annulus Leakage 12.1-32 Isotopic Curie Content-Equilibrium Core-EOC 12.1-63 33 CRDL Activation Source Term 12.1-64 d.1-34 Primary Pump Intermediate Level Waste System 12.1-65 Radioisotope Inventory 12.1-35 Low Activity Level Liquid Waste Collection Tanks 12.1-66 Radioisotope Inventory 12.1-36 Low and Intermediate Level Liquid Waste Collection 12.1-67 Filters Radioisotope Inventory 12.1-37 Intermediate and Low Activity Level Evaporator 12.1-68 Filters Radioisotope Inventory 12.1-38 Intermediate and Low Level Liquid Wash Evaporators 12.1-69 Radioisotope Inventory ' 12.1-39 Intermediate and Low Level Demineralizers 12.1-70 Radioisotope Inventory 12.1 40 Intermediate and Low Level Distillate 12.1-71 Demineralizer Resin Traps Radioisotope Inventory 12.1-41 Intermediate Level Liquid Waste Precoat 12.1-72 12.1-42 Intermediate AcHvity Waste Filter Sludge Receiver 12.1-73 Radioisotope Inv tory 12.1-43 Solid Radioactive Waste Concentrated Waste Tank 12.1-74 Radioisotope Inventory 12.1-44 Solid Radioactive Waste Decanting Tank Radio- 12.1-75 isotopic Inventory 12.1-45 Solid Radioactive Waste Decant Filters Radio- 12.1-76 isotope Inventory 12.1-46 Solid Radioactive Waste Radwaste Drum 12.1-77 50 Radioisotope Inventory Amend. 50
,funelg7_9, d
12-v 200 50'
LIST OF TABLES Continued T a b l e Np._ Page 12.1-47 Solid Radioactive Waste Radwaste (Spent Resins- 12.1-78 Filter Sludge) Radioisotope Inventory 12.1-48 Area Monitor Descriptions 12.1-79 12.1-49 Dose Rates Selected Plant Location Due to 12.1-80b CRBRP Direct Radiation 50 12.1-Al Experimental Program Related ta CRBRP Shielding 12.lA-6 12.2-1 Leakage Rates into Normally Accessible Areas 12.2-7 12.2-2 Radiocuclide Concentrations in Head Access Area - 12.2-8 Design Values 12.2-3 Location of Continuous Air Monitors 12.2-9 12.2-4 Expected Annual Exposure in Normally Accessible Cells 12.2-11 12.3-1 Typical Portable Health Physics Equipment 12.3-8 12.3-2 Typical Health Physics Laboratory Equipment 12.3-10 12.3-3 Estimated Man Hours of Access to Radiation Arcas During Normal Operation and Operational Occurrences 12.3-11 12.3-4 Estimated Man Hours of Access to Accessible Radiation Zones 12.3-12 '25 12.3-5 Personnel Protection Monitors Area Monitors 12.3-13 50 12.3-6 Personnel Protection Monitor:ng Continuous 12.3-15 Air Monitors q on
,bb
( nuo Amend. 50 June 1979 12-v'
LIST OF FIGURES FIGURE NO. PAGE 12.1-1 Plant Radiation Protection 12.1-81 12.1-2 Plant Radiation Protection 12.1-82 12.1-3 Plant Radiation Protection 12.1-83 12.1-4 Plant Radiation Protection 12.1-84 12.1-5 Plant Radiation Protection 12.1-85 12.1-6 Plant Radiation Protection 12.1-86 12.1-7 Plant Radiation Protection 12.1-87 12.1-8 Plant Radiation Protection 12.1-88 12.1-9 Plant Radiation Protection 12.1-89 12.1-10 Plant Radiation Protection 12.1-90 12.1-11 Plant Radiation Protection 12,1_91
'2.1-12 Plant Radiation Protection 12.1-92 12.1-13 Plant Radiation Protect;an 12.1-93 12.1-14 Plant Radiation P,rotection 12,1_94 0 12.1-15 Plant Radiation Protection 12.1-95 12.1-16 Plant Radiation Protection 12.1-96 12.1-17 Plant Radiation Protection 12.1-97 12.1-18 Plant Radiation Protection 12.1-98 12.1-19 Plant Radiation Protection 12.1-99 1 .' ! 1.u.t "a 'i a ti c: I m s tior. l',1 age 12.1-19b Plant Radia tion Protection 12.1 n9L lt .i-lh Plant Radiation Protection 1?,1-qor 50 12.1-19d Plant Radiation Protection 12.1-99d 12.1-20 Control Room Shielding Perspective 12.1-100 12.1-21 Functional Block Diagram of an Area Radiation 12.1-101 Monitor 12.lA-1 Flow Chart of General Anal3 'ical Method 12.lA-8 Amend. 50 12-vii 2Og ;j j June 1979
12.2-1 PPS Containment Exhaust Radiation 12.2-12 Monitoring Channel 12.2-2 Non-PPS Air Radiation Monitoring Channel 12.2-13 268 2';0 Amend. 50 June 1979 12-viii
CHAPTER 12 - RADIATION PROTECTION LIST OF REFERENCES Section Page No. 12.1 12.1-27 12.lA 12.lA-5 O 12-1x na m
TABLE 12.1-8 INVENTORY OF RADI0ACiiVE PRIMARY SODIUM IN THE PPTI+ AND PSSP++ Inven tory _(Ci ) Isotope PPTI PSSP 24 Na 103 121 c2 Na l37 1.2 x 10-2 1.4 x 10-2 Cs 0.29 0.36 l36 Cs l34 6.1 x 10
-2 7.1 x 10 -2 Cs l25 7.8 x 10-3 9.3 x 10-3 Sb 1.7 x 10-3 1.9 x 10-3 1 131 0.17 0.20 l32 0.013 Te 0.015 6 l2 0.12 0.13 If29m Te 3 -2 3.0 x 10 -3 Te l29 2.5 x 10 3 3.0 x 10 Sr 89 2.5 3.8 xx 10 10-4 4.6 x 10 -4 Sr 90 2.4 x 10-4 -4 2.8 x 10 ly 2.4x10j 1.1 x 10 2.8x10j Y
95 1.3 x 10 -4 Zr 95 2.1 x 10-4 2.4 x 10 -4 2.1x10) Nb l03 - 2.4 x 10 Ru l06 2. 9 x 10 3.4 x 10-4 Ru 06 2.1 x 10-4 2.4 x 10-4 Rh 4 2.1 x 10- 2.4x10j Ba l0 2.1 x 10- 2.7 x 10 -4 La 2.3 x 10-4 2. 7 x 10 -4 Ce 2.7 x 10-4 3.1 x 10 C' 4f 4 1.6x10j 1.8x10j Pr 1.6 x 10 1.8 x 10 Pr l I4 1.9 x 10 -4 2.3 x 10 -4 Nd l7
-5 9.0 x 10 -5 1.1 x 10 -4 Pm 60 9.0 x 10 1.1 x 10 -4
-7 Co 8.0 x 10-6 9.5 x 10 -6 1.5 x 10 -5 S8 Co 54 1.S x 10 -
Mn 5.9 x 10-5 7.0 x 10 ^ Pu 238 9 5.7 x 10 -0 6.8 x 10 -5 E Pu 1.5 x 10- 1.3 x 10-5 Pu 240 241 2.2x10l 2.3 x 10 _5 3 Pu 1.7 x 10-3 Pu 242 4.1 x 10- 2.2 x 10 5.2 x 10- 8 4gl H 3 8.2 x 10-3 9.7 x 19 -3 50l Tc 103.09 121.8
+ Primary Plugging Temperature Indicator ') { g
++ Primary Sodium Sample Package - '? n' '-")
12.1-37 Amend. 50 June 1979
50l TABLE 12.1-34 A PRIMARY PUMP INTERMEDIATE LEVEL (IALL) WASTE SYSTEM RADI0 ISOTOPE INVENTORY Inventory (Curies) 10 Days After Isotope Reactor Shutdown 51 Cr 3.92 S4 Mn 119.3 59 Fe .43 S8 Co 63.2 0 C0 31.7 89 Sr 78.8 90 Sr 57.7 y 90 57.7 y 91 22.9 95 Zr 43.2 95 Nb 43.2 99 Mo 4.8
,4.103 57.2 6
Ru 46.3 6 Rh 46.3 Ill Ag 1.5 Te l27 157.6 l27m Te 157.6 129m Te 475.5 Te l9 475.5 Te l2 338.8 131 7 0.23 ' 132 1 338.8 Amend. 50 12.1-65a Jura 1979
50 l TABLE 12.1-34 A (Continued) Inventory (Curies) 10 Days After Isotope Reactor Shutdown l34 Cs 0.085 l 36 Cs 0.42 7 Cs 3.2 l40 Ba 31.2 l40 La 31.2 4l Ce 51 .4 l4 Ce 23.0 Pr l4 23.0 l44 Ce 36.7 Pr 36.7 I fid 11.1 Pm" 20.8 9 Pm .61 Eu 2.7 6 Eu 1.02 182 Ta 4.13 238 Pu ,47 Pu .12 24 Pu .16 241 Pu I4 I Pu .0003 ,n ') O E' 238 m/ t, G L' 74p ,00001 4g 239 74p .07 Amend. 50 12.1-65b June 1979
50 l ' TABLE 12.1-34 A(Continued) Inventory (Curies) 10 Days After Isotope Reactor Shutdown g 241
.13 g 242m .005 g 242 2.3 x 10' g 243 .002 242 Cm .095 Cm .001 49 Cm .028 ont
,2bd ti , J 12.1-65c Amend. 50 June 1979
TABLE 12.1 47 SOLID RADI0 ACTIVE WASTE RA0 WASTE DRUM (SPENT RESIllS - FILTER SLUDGE) RADI0 ISOTOPE IflVENTORY ISOTOPC INVCUTORY (CURIE 3) ISOTOPE INVEUTORY (CURICS) Cr-51 2.31(-1) Cc-141 2. 9 9 (- 2) Mn-54 7.00( 0) Cc-143 1.34(-1) Fe-59 2.53(-2) Pr-143 1. 3 4 (-1) Co-58 3.72( 0) Cc-144 2.14, (.-1) Co-60 1.86( 0) Pr-144 2.14(-1) Sr-89 4.58(-2) Nd-147 6. 41 (- 2 ) Sr-90 3.3 6 (-1) Pm-147 1.21(-1) Y-90 3.36(-1) Pm-149 3.55(-3) y-91 1. 34 (-1) Eu-155 1. 57 (-2) 2r-95 2.52(-1) Eu-156 5. 94 (-3 ) 2.52(-1) Nb-95 Ta-l82 2.s43(-1) y,o-99 2.80(-2) Pu-238 2.74 (-3) ku-103 3.34 (-1) Pu-239 7.00(-4) Ru-106 2.70(-1) Pu-240 9.35(-4) Rh-106 2.70 (-1) Pu-241 8.17(-2) Ag-lll 8. 7 6 (-3 ) pu-242 1. 7 5 (-6) Te-127 9.17 (-1) Np-233 8. 7 6 (-8) Te-127m 9.17 (-1) Np-239 4. 08 (-4 ) Tc-129s 2.77( 0) Am-241 7.53 (-4) Te-129 2. 77 (__0) Am-242m 2.91(-5) Te-132 1.98( 0) An-242 1.34(-10) I-131 1.34(-3) Am-243 1.16 (-5,) 1-132 1.98( 0) Cm-242 1. 4 2 (-4 ) Cs-134 4. 94 (-3) Cm-243 5.84(-6) Cs-136 2.45(-3) h-244 1. 63 (-4 ) Cc-137 1. 8 6 (-2) , Da-140 1. 82 (-1) 268 (_/u 19 La-140 1.82(-1) TOTAL 28.4 Amend. 49 April 1979
TABLE 12.1-48 AREA MONITOR DESCRIPTIONS Location Area Design Instrument Basis Cell Dose Rate Range for Area Bldg. No. mR/hr mR/hr Location I&C Cubicle
- 49) El. 824'-3" RCB 162 0.2 0.01-103 1, I&C Cubicle 491 El. 824'-3" RCB 163 0.2 0.01-103 1, I&C Cubicle 50 l 49l El. 824'-3" RCB 164 0.2 0.01 -10 1.,3.,4.
Refueling Hatch El. 816' RCB - 0.2 0.01-107 1.,3.,4. Airlock El . 816 ' RCB - 0.2 0.01-103 1, Operating El. 794' RCB 105V 2.0 0.1-104 1.,2. Operating El. 780' RCB 105U 2.0 0.1-104 1.,2. Operating El, 766' RCB 105M 2.0 0.1-104 1.,2. Operating Ei. 766' RCB 105N 10. 0.1-104 1.,2.,5. Operating El. 766' RCB 105R 10. 0.1-104 1.,2.,5. Operati ng El. 766' RCB 105T 10. 0.1-10 4 1.,2.,5. Operating El. 766' RCB 105P 10. 0.1-104 1.,2.,5. Operating El . 766' RCB 105Q 2.0 0.;-104 1.,2. Operati ng 1 El. 733' RCB 105A 2.0 0.1-104 1. ,2. '/ (; 8 ,2 '-) 12.1-79 f..und. 50 June 1979
TABl.E 12.3-3 EST1.'iATED Mail HOURS OF ACCESS TO RADI AT10!1 AREAS DURIflG I;0P?iAL OPERAT10:1 A!!D OPERATIODAL OCCURREfiCES Operating Area Man Hours / Quarter
- 1. Reactor Containment Building a, Operating Floor and Adjacent Balconies (Zone 1) 1275
- b. Cells Celow the Operating Floor in f1E, "E, 950 and SW Corncrs and Cell 152 (Zone 2)
- c. Pump Wells (Zone 2) 580
- d. ilead Access Area (Zone 2) 220
- e. All remaining Accessible Cells (Zone 3) 365 Total Reactor Containnent Duilding 3390
- 2. Reac*or Service Building
- a. Operating Floor, Dalconies, and Refueling Cask 7560 Shaft and Corridor (Zone 1)
- b. CHRS Cells (Zone 1) 70
- c. Access Corridors and Adjacent Sa:r.pling and Value 2490 Gallery Cells Below the Operating Floor (Zone 2)
- d. Fuel llandling Cell (Zone 1) 5720
- e. Rad.vaste Processing Area (Zones 1 & 2) 1315 Total Reactor Service ;uilding 17155
- 3. Provision for Rc> quired Health Physics Coverage 3640 Total man hours in accessible areas of Reactor 24185 man hours Containment and Reactor Service Building quarter 36% '398 12.3-11 Amend. 49 April 1979
O TABLE 12.3-4 ESTIt',/JED IVJ4110'JRS OF tCCESS TO ACCESSICLE PADI ATION ZONES Zone ihn floors / Quarter Zone I 17850 Zone II 5740 Zone III 595 49 Total 24185 0
-r'd n rj
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t Amend. 50 12.3-12 June 1979
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- T G r 1 1 1 1 1 0 0 0 0 0 0 0 0 0 T Nh 1 1 1 0 0 3,
1 1 1 1 I 1 1 1 1 E 1 1 1 A/ - - - 1 N DRP - - - - - - - - - - - - - 2 O n 1 1 1 1 1 1 M ( 0 0 0 0 01 , 1 1 1 1 1 1 1 1 1 1 1 1 1 1 0 0 0 0 0 0 0 0 0 0 0 0 , E N 0 0 0 L O @TB A T C E T I P E Y D B D B D B D B D B B D D D D D D D D D D D D D D O T T T T T T T B B B B B E B B B B B B D R T T T T T T T T B B T T T T T T P L E N N O a a a a a a a S R E P a G m m G mmmmmmm m a a G G a G a G a G a a m a a mmam mmm a a a a a a m a n a a ma ma ma a a a mm a m a a a r m m a a m a P CE G C G G G G G G G TP t t t G G G G t t t t t t t I Y c c c c c t t t t t N r' e e c c c c c c c c t t t t t O e e e e e e e e c c c c c c c M r r r r r r r r r r e r e r e e e e e e e e i i i i i i r r r r r r r r D D iD iU D D i i i i D D D' D D D D Di i i D D i i i i D D D D R h n a OSD c e e r r
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m o l o o AOT d g le g g g l R . RI n n ur n n n r g APN a F o i i i l b d E O H it d it t t A o a R M a r ti a a a r L lB . A l nr r r r r t gg e e er e e e e n t tdd u p po p p m o o g F O SC @ U O E C H nl l EBB m d aer lB co l a i t e t vagc L A F A B nrrdi L 8 9 3 3 3 2 3 1 oeel v E 3 3 1 1 1 e 5 3 - CS nB r C 3 3 3 3 3 2 2 4 e e 4 r rGl S O' oo o I V ' tt mrt T E 1 '9 '9 '5 '3 '6 '4 '6 '6 ccat n A L 8 7 4 6 3 1 9 1 1 aaena C E 7 7 7 7 / 8 7 8 8 eet ol O D RRSCP L N G E - - - - - D B B B B B B B B B G L D S R S R S R S R S R G S G S C S P E B38BC l lO3 ,G3 L CSGC~' t h c R RS 94 1 NN M,O O @ yWC ,.- ()Y)
9 TABLE 12.3-6 PERSONNEL PROTECTION MONITORING - CONTINUOUS AIR MONITORS LOCATICS AREA AND/OR DETECTOR PROCESS MONITOR "LARM I MONITOR MONITCRED
TYFE T"E SENSITIVITY BACKGROUND SETPT.(s) OUTPUT REMARr5 ELDG. ELEV. CELL (tCi/cm) (mR/hr) pCi/cc l(note 2)
_ _ _ _ __ l 816' Operating Floor -10 FCB - Particu~ ate 8 SCINT 10 Cs-137 0.2 TBD A fiobile Mcnitor - Pro-
-10 vides continuoas Radioicdine a SCINT 10 I -131 0.2 TBD nonitoring of con-
-6 Gaseous SCINT 10 Kr-85 0.2 TBD 816' Particulate - SCINT -10 RCB -
Operating Floor 10 Cs-137 0.2 TBD A Mobile Monitor - can
-6 be used to monitor Gaseous s SCINT 10 Kr-85 0.2 TBD individual inerted cells 2
PCB 816' - Operat.ng Flocr Alpha a SCINT 10 Pu-239 0.2 TBD A Mobile Monitor - pro-vides continuous monitoring of con-tainment atmospnere h u FCB 766' 105M Operating Floor Particulate > SCINT 10
-10 Cs-137 2.0 TBD A Mobile Monitor - can be us.ed to monitor
-6 4
m Gaseous o SCINT 10 Kr-85 2.0 individual inerted cells Operating Floor -10 RSB 816' 364A Particulate B SCINT 10 Cs-137 0.2 TBD A Mobile Monitor - can
-10 be used to monitor Radioiodine a SCINT 10 I-1 31 0.2 TBD individual inerted
-6 Gaseous e SCINT 10 Kr-85 0.2 TBD 816' Operating Floor Alpha -12 FSB 364A a SCINT 16 Pu-239 0.2 TBD A Mobile Monitor -
provided continuous monitoring of RSB Atmosphere RSB 79u' 355A Operating Floor Particulate e SCINT 10~ Cs-137 0.2 TBD A Mobile Monitor - can be used for post-
~6 Gneous e SCINT 10 Kr-85 0.2 TBD accident of contain-C ment atmosphere CC
-10 IB- 816' 262 Operating Floor Particulate 8 SCINT 10 Cs-137 0.2 TBD A Mobile Monitor - can r SGB be used for post-
-10
{". mM'Q Radiciodine r SCINT 10 I-i31 0.2 TBD accident monitoring r, of containment U '$ . Gaseous s SCINT 10
-6 Kr-85 0.2 I TBD Stmosphere
TABLE 12.3-6 (C0tlTINUED) LOCATIO'4 ARFA A.D/OR N DETECTOR .
" I PROCESS MONITOR TYPE 3ENSITIVITY BACKGROUND SETPT.(s) OUTPUT REMARKS MONITCRED TYPE pCi/cc BLDG. ELEV. CELL ( Ci/cm) (mR/hr) (note 2)
-10 Mobile Monitor - Pro-CB 316' -
Control Room Particulate 8 SCINT 10 Cs-137 - TBD A vides continuous Radiciodine a SCINT 10-10 1 -131 - TBD monitoring of con-
-6 Kr-85 TBD Gaseous 6 SCINT 10 -
U u LEGEND NOTE: Monitor lutput I
$ RCB - Reacter Containment Bldg. A. Monitor housing includes: loss-of-signal, high radiation, high-high radiation indicator lights, sample flow reading, ccur.trate meter RSB - Reactor Service Building (per detecticn channel), multipoint strip-chart recorder, audible alarms.
IB-SGB - Intermediate Bay-Steam Generator Building B. Radiaticn Monitoring Panel (Control Room): loss-of-signal, high CB - Control Building radiation, high radiation indicatnr lights, countrate meter (per detection channel). Alarm annunciation (loss-of-signal, high SGB - Steam Generator Building radiation, high-high radiation) providad by Plant Control (System 90). Permanent recording by PDH & DS (System 91). SCINT- Scintillation 14 9 C. Same provisions as B. N c:, -~. E 2 r N, J cg 3g M O O O
CHAPTER 13.0 CONDUCT OF OPERATIONS TABLE OF CONTENTS PAGE 13.0 CC ~'CT OF OPERATIONF 13.1 ORGlJ'7ATIONAL S'RUCTUP.E OF g'" CANT 13.1-1 13.1.1 Project Organi'ation 13.1-1 13.1.1.1 Functions, Responsibilities, and Authorities 13.1-1 of Project Participants 13.1.1.2 Applicants' In-House Organization 13.1-1 ,25 13.1.1.3 Inter-relationships with Contractors and 13.1-1 Suppliers I 13.1.1.4 Department of Energy (Duc) Participation 13.1-1 l25 Sol 13.1.1.5 Technical Support for Operations 13.1-1 13.1.1.5.1 TVA's Technical Staff 13.1-2 13.1.1.5.1.1 Division of Power Production 13.1-2 13.1.1.5.1.2 Other TVA Organizations 13.1-4 34 13.1-4 13.1.1.5.2 Project Technical Support 13.1.2 Operating Organization 13.1-4 13.1.2.1 Plant Organization 13.1-4 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 Maintenance Section 13.1-6 13.1.2.1.4 Health Physics Unit 13.1-6 13.1.2.1.5 Administrative Group 13.1-7 50 13.1.2.1.6 Division of Power Systems Operation (PS0) 13.1-7 Engineering Unit 13.1.2.1.7 Nuclear Plant Quality Assurance Staff 13.1-7 13.1.2.2 Personne~. Functions, Responsibilities and 2b18.157)., @ Authorities Amend. 50 13-1. June 1979
TABLE OF CONTENTS (Continued) PAGE 13.1-a gg l 13.1.2.2.1 Plant Manager 13.1.2.2.2 Assistant Plant Manager 13.1-8 Plant Operations Supervisor 13.1-8 13.1.2.2.3 13.1.2.2.4 Assistant Plant Operations Superviscr 13.1-8 13.1-9 13.1.2.2.5 Plant Engineering Supervisor 13.1 9 59 13.1.2.2.6 Plant Maintenance Supervisor Assistant Plant Maintenance Supe-cisors 13.1-9 13.1.2.2.7 13.1-9 13.1.2.2.8 Health Ph'sicist 13.1 10 13.1.2.2.9 Supervisor, Nuclear Plant Quality Assurance Staff 13.1.2.2.10 Safety Engineer 13.1-10 13.1.2.3 Shift Crew Composition 13.1-10 50 13.1.3 Qualification Requirements for Nuclear Plant Personnel 13.1-11 13.1.3.1 Minimum Qualification Requirements 13.1-12 1 3.1-16 13.1.3.2 Qualifications of Plant Personnel 13.2 TRAINING PROGRAM 13.2-1 13.2.1 Program Description 13.2-1 13.2.1.1 Program Content 13.2-1 13.2.1.2 Coordinction with Pre-operational Tests and 13.2-2 Fuel Loao.ng 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
}{g 7, O/ Amend. 50 0
.,u' June 1979 13-11..
TABLE OF CONTENTS (Contirued) PAGE 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 Respon,ible Individual 13.2-5 13.2.2 Retraining Program 13.2-6 13.2.3 Replacement Training 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 13.3 EMERGENCY PLANNING 13.3-1 13.3.1 General 13.3-1 13.3.2 Emergency Organization 13.3-2 13.3.3 Cooruination with Offsite Groups 13.3-4 50l 13.3.4 Protective Action Levels 13.3-5 13.3.5 Protective Measures 13.3-5 13.3.6 Review and Updating 13.3-5 50l 13.3.7 Medical Support 13.3-6 13.3.8 Drills 13.3-6 13.3.9 Training 13.3-6 13.3.10 Recovery and Re-entry 13.3-7 50 13.3.11 implementation 13.3 7 Amend. 30 13-111 Jung 1929 _,
/ g tj g(j,
TABLE OF CONTENTS (Continued) PAGE 13.3.12 Radiological Analysis 13.3-8 11 3.12.1 Projected Ground Level Doses 13.3-8 13.3.12.2 Accident Assessrrent, Warning and Evacuation Times 13.3-8 13.3.12.2.1 Assessment 13.3-8 13.2.12.2.2 Warning 13.3-8 S0 13.3.12.2.3 Evacuation 13.3-9 13.4 REVIEW AND AUDIT 13.4-1 13.4.1 Review and Audit - Constreetion 13.4-1 O Amend. 50 June 1979 13-iiia
TABLE OF CONTENTS (Continued) PAGE 13.4.2 Review and Audit - Test and Operation 13.4-1 13.5 PLAflT PROCEDURES 13.5-1 13.5.1 General i3.5-1 13.5.2 flormal Operating Instructions 13.5-1 13.5.3 Abnormal Operating 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 Technical Instructions 13.5-4 13.5.8 Section Instruction Letters 13.5-4 13.5.9 Site Erergency Plans 13.5-4 50 13.5.10 Radiation Control Instructions 13.5-4 13.6 PLANT RECORDS 13.6-1 13.6.1 Plant History 13.6-1 13.6.2 Operating Records 13.6-1 13.6.3 Event Records 13.6-1 13.7 INDUSTRIAL TECURITY 13.7-1 13.7.1 Organization and Personnel 13.7-1 ! 13.7.1.1 Division of Property and Services 13.7-1 50 : 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 i 13.7.1.5 Industrial Security Traiaing 13.7-4 i i 13.7.2 Plant Design 13.7-4 @ 50 13.7.2.1 Degn Features g7 c) 13.7-5 l1 Amend. 50 13-1y June 1979
1ABLE OF C0flTEflTS (Continued) PAGE 13.7.2.2 Physical Arrangemer.ts 13.7-6 13.7.2.3 Owner-Controlled 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 Plan 13.7-8 13.7.3.1 Access Control 13.7 3 13.7.3.2 Control of Equipment by Categories 13.7-9 13.7.3.3 Access _ontrol During Emergencies 13.7-10 13.7.3.4 Surveillance of Vital Eq.1pment and Material Access Areas 13.7-10 13.7.3.5 Potential Security Threats 13. 7-10 a 13.7.3.6 Administrative Procedures 13.7-10a
< 13.7. 3.7 Test and Inspection 13. 7 -1 J b 1 50 13-v Amend. 50 June 1979
,'j g '_; ,is
CHAPTER 13 C0flDUCT OF OPERATI0 tis LIST E TABLES TABLE NUMBER TITLE PAGE 13.1-1 Technical Support Sunwry 13.1-17 13.3-1 Participants in CRBRP Radiological Emergency P1an 13.3-11 13.3-2 Summary of Data Utilized for Source Term Radiological Analysis 13.2-12 13.3-3 Projected Maximum Resident + Transient Population Distribution Within 5 Miles of the Demonstration Plant for Census Year 1980 13.3-14 13.3-4 Projected Maximum Resident and Transient Population
- in Evacuation Sectors Within 50 5 Miles of CRBPP 268 3;;
Amend. 50 13-vi June 1979
CHAPTER 13 CONDUCT OF OPERATIONS LIST OF FIGJRES FIGURE NO. TITLE PAGE 13.1-1 CRBRP Organization Chart 1 3.1- 21 50 13.2-1 Proposed Training Schedule 13.2 ' 13.3-1 Elapsed Exposure Time to Reach Specific Bone Dose Versus Downwind Distance (Based on Site Suitability Source Term) 13.3-16 13.3-2 Elapsed Exposure Time to Reach Specific Lung Dose Versus Downwind Distance (Based on Site Suitability Source Term) 13.3-17 13.3-3 Elapsed Exposure Time to Reach Specific Thyroid Dose Versus Downwind Distance (Based on Site Suitability Source Term) 13.3-18 13.3-4 Elapsed Exposure Time to Reach Whole Body Dose er;us 00wnwind Distance (Sased on Site Seitability Source Term) 13.3-19 13.3-5 Project Area - North Hal f 13.3-20 50 13.3-6 Project Area - South Hai f 13.3-21 13.5-1 Plant Procedures 13.5-5 13.7-1 Security Organization 13.7-11 13.7-2 Site Plan 13.7-12 13.7-3 Security Genaral Arrangements -- t *oprietary) 13.7-4 sailding General Arrangements -- (Proprietary) 9q Amend. 50 O 13-vi i c . June 1979
CHAPTER 13.0 CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE OF THE APPLICANT Contract AT (49-18)-12 has been established to design, construct and operate a Liquid Metal Fast Breedec Reactor demonstration plant. The i
*., parties to the contract are the Department of Energy (DOE), the Tennessee l25 Valley Authority (TVA), the Commonwealth Edition Company (CE), and the Project Management Corporation (PMC). The organizational structure 5(j of the applicant (DOE, PMC, and TVA) is covered in Section 1.4. 25 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 Project Organizatic, 13.1.1.1 Functions, Responsibilities, and Authorities of Project Pa rticipan ts The functions, responsibilities, 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. 25 13.1.1.2 Applicants' In-House Organization This material is covered in Section 1.4.2. 13.1.1.3 Interrelationships with Contractors and Supplierr This material is covered in Section 1.4.3. 13.1.1.4 Department of Energy (DOE) Participation 5'0 The participation of DOE in the CRBRP Project is described in Section 1.4. In addition, DOE participates in R&D in support of the CRBRP Project through its LMFBR bace technology programs described in Section 1.5. 25 13.1.1.5 Technical Support for Operations TVA's Office of Power will be responsible for carrying out the operator role for the Clinch River Breeder Reactor Plant. Within the 50l TVA Office of Power, the Division of Power Production (P PROD) will be responsible for the operation and maintenance of the CRBRP. The TVA
, cechnical staff supporting the operation of the CRBRP will consist of 5(1 P PR00's home office staff in Chattanooga and also support from other divisions and offices within TVA (Section 13.1.1.5.1). In addition, SC' technical support will be supplied by the technical staffs of PMC, DOE, WARD. and Burns and Roe (Section 13.1.1.5.2). 25
@ 268 313 13,1-1 Amend. 50 June 1979
13.1.1.5.1 TVA's Technical Staff h 13.1.1.5.1.1 Division of Power Production _ The Division of Fower Production is responsible for the operation and maintenance of all TVA power plants and will have this responsibili;y 50 for the CRBRP. Included within P PROD for technical support are the fluclear Generation Branch, the Plant Engineering Branch, and the Power Plant Maintenance Branch. Table 13.1-1 provides a summary of the number of personnel in each of these branches as well as their educational background and tech lical experience. The fluclear Generation Branch will direct the operation and maintenance of the CRBRP; rcview and analyze operating and engineering data, regular and special reports, test results, and other information pertaining to the operation and maintenance of the CRBRP; and review and coordinate operating maintenance, and surveillance procedures to ensure that the CRBRP is operated to provide maximum safety along with achieving project operating and demonstration goals. The fluclear Generation Branch is responsible for providing input to the Office Power for operational aspects of the plant design. During initial oporation of the Plant, the Branch coordinates 50l and activities through acceptance teststhe withplar,. PMC, manager includir.g DOE, fluclehr preoperational, Regulatory startup, Commission (NRC), reactor manufacturer, and the equipment suppliers. The Branch will ' coordinate the activities of the CRBRP with other branches and divisions within TVA ii such areas as onsite fuel management and waste disposal. 30] The staff of this branch meets the " staff specialist" defini-tion of ANSI fil8.1-1971, and consists of the following: Chief, Nuclear Generation Branch Assistant Chief, Nuclear Generation Branch Supervisor, Reactor Engineering Staff Supervisor, fluclear Operations Staff 50 Supervisor, Praoperational Staff The Nuclear Generation Branch will analyze reports of abno. mal equipment operation or faults to determine the initiating cause so that remedial measures can be taken. The Branch also assists in training the plant staff and evaluating training programs. Amend. 50 13.1-2 June 1979 mm
The Chief, fluclear Generation Branch, meets both the definition and qualification of " Engineer in Charge" as set forth in " Standards for Selection and Training of Personnel for tiuclear Power Plants", ANSI til8.1-1971. The Plant Engineerino Branch provides a variety of engineering services for all TVA generating plants. For t'9 nuclear plants, it provides technical assistance in the area of nuclear engiaeering, chemical engineering, instrument erigineering, and testing. The staff of toe Plant Engineering Branch meets the " staff specialist" definition of AtiSI til8.1-1971. These staff specialists are: Chief, Plant Engineering Lranch Staff Environmental Engineer Staff Mechanical Engineer, Equipment and Structures Staf f Mechanical Engineer, Testing Staff Chemical Engineer Staff Mechanical Engineer, Thermal Cycles 50 Staff Instrument Engineer The Power Plant Maintenance Branch provides services from the central office staff on electrical and mechanical maintenance problems and major shop services from its central service shops. The staff of this branch meets the " staff specialist" defini-tion of AtiSI til8.1-1971. Those staff spe ialists are: Chief, Power Plant Maintenance Branch Staff Nuclear Engineer Staff Electrical Maintenance Enginee Staff Turbine Maintenance Engineer Staff Boiler and Reactor Maintenance Engineer Staff Technical Support and Planning Engineer 50 Superintendent, Power Service Shops Amend. 50 13.1-3 June 1979 268 3iL
Personnel within the organization of the Plant Engineering Branch and the Power Plant Maintenance Branch provide technical support to nuclear facilities when called upon to do so. 13.1.1.5.1.2 Other TVA Organizations Other organizations within TVA which supplement the CRBRP and Division of Power Production Staff are as follows: Division of Power Systen. Operations (PS0) Division of Transmission Planning and Engineering (TP&E) Division of Power Pesource Planning (PRP) Division of Engineering Design (EN DES) Division of Construction (CONST) Division of Chemical Development (CHEM D) Division of Medical Services (MED SV) Division of Property and Services (P & SVS) 50 Division of Environmental Planning (ENV PL) Office of PowerM anagement Services Staff Office of Power Quality Assurance and Audit Staff f 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 Support Project technical support for the operation of the CRBRP will Snl and Burns and Roe.be Areasprovided of supporttowill thebeDivision of Power in accordance withProduction the in TVA by PMC responsibilities described in Section 1.4. 13.1.2 Operating Organization 13.1.2.1 Plant Organization 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 Amend. 50 13.1-4 June 1979
Plant Engineering Section, and the Plant Maintenance Section. Sta f f services are provided by an administrative staff, a Quality Assurance staff, and the Health Physics Unit of the Radiclogical Hygiene Branch. The latter is under the administrative supervision of the Division of Environmental Planning. The CRBRP organization follows the pattern developed through experience and used at all TVA fossi? and nuclear generating plants. Plant employees are selected primarily from existir:g TVA con-ventional and nuclear plant staffs and DPP's central office. Personnel qualifications shall meet tne criteria set forth in the AN"I N18.1-1971. 13.1.2.1.1 Plant Operations Section The Plant Operations Section is responsible for all plant operations. It provides operating personnel for the preoperational 50l testing, fuel loading, startup testing, startup, and plant operation. It is responsible for coordinating and scheduling the training program for alI operations oersonnel. It provides the nucleus of emergency teams such as the plant rescue and fire-fighting organizations. The Plant Operations Section is under the direction of the Operations Supervisor who holds a valid NRC Senior Ructor Operator (SRC) license. He is assisted by an inline Assistunt Operations 50 Supervisor who also holds a valid NRC SR0 license. Within the Plart Operations Section are five shift crews. The minimum shif t crew will consist of the Shift Engineer who holds 50 an NRC SRO license, one Assistant Shift Engineer who holds an NRC SR0 license, one Unit Operator who holds an NRC Reactor Operator (RO) license, and three Assistant Unit Operators. One Health Physics Technician will also be assigned to each shift. Additional operators are assigned as necessary. Plant management and techiical support Mill be present or on call at all times, 13.1.2.1.2 Plant Engineering Sectica The Plant Engineering Section is under the direction of the Engineering 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, mechanical, instrumentation, and chemical engineering. Responsibilities of this section include plant and equipment performance tests, inplant fuel malagement, waste management, chemistry control, and instrumentation maintenance. 13.1-5 Amend. 50 June 1979
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O The Plant Engineering Section carries out a comprehensive program of plant tests, studies, aad investigations for the purpose of m nitoring the reactor, engineered safeguards, and plant operating conditions to assure cc.npliance with the operating license and technical specifications and to improve the efficiency of the plant. This includes the coordination of the surveillance test program with other plant sections. 13.1.2.1.3 Plant Maintenance Section The Plant Maintenanc; Section is under the direction of the Maintenance Supervisor. He is assisted by two inline Assistant Maintenaace Supervisors. The Plant Maintenance Section is responsible for mechanical and electrical maintenance work and inspections in the plant. This includes scheduling and conducting periodic inspections and tests on tfe systems assigned to this section associated with the reactor 50l and engineered safeguards, as required by the technical specifications and operating license. This section develops and carries out a preven-tive maintenance program that assures that the repair and replacement of parts are consistent witn 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 main-tenance .eports. 13.1.2.1.4 Health Physics Unit The Health Physics Unit of the Padiological 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, pro':1 ding spe-cialized training in radiation protection. It conducts comprehensive onsite environmental radiation monitoring before, dring, and af ter plant startup and provides radiological health coverage for all ope-rations 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 radiation, and contamination levels. This unit is under the administrative supervision of the Chief, Radiological Hygiene Branch in the Division of Environmental 50 Planning, aM under the functianal supervision of the Plant Manager. 13.1-5 Amend. 50 June 1979
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w 13.1.2.1.5 Ad inistrative Group The administrative staf f, under the supervision of the Super-visor, Administrative Services, performs management service functions and clerical services for the plant. 13.1.2.1.6 Division of Power System Oyerations [PS0) Engineering 50 Unit Si The P50 Engineering Unit, of the Division of Power System Operations, is responsible for the maintenance and testing of the relaying associated with the transmission system. They are also responsible for maintenance of all external communications systems at the plant (with the exception of the Bell Systems Equipment). They are responsible for maintenance of portions of the onsite distribution and bus protection relaying. This un# t is under the administrative supervision of the Chattanooga Area Superintendent in the Division of Power System Operations and under the functional supervision of the Plant Manager. 13.1.2.1.7 ((uclear Pl ant Quality Assurance Staf f The nuclear p' ant quality assurance staff is under the direction of the Quality Assurance Staff Suoervisor. The nuclear plant quality assurance staff is responsible for develnping, planning, initiating, and directing a comprehens've nuclear plant quality assurance / quality control program in the olant. Respon-sibilities include informing and advising other plant sections of the applicability, requirements, and implementation of the quality assurance program. The nuclear plant quality assurance staff is rt_ ponsible for coordinating, scheduling, and verifying surveiliance monitoring and 50 'nspections of safety-related structures, systems, and components. 13.1.2.2 Personnel Functions, Responsibilities, and Authorities During normal plant operations, the plant manager is responsi-ble for all plant activi ties. In the event of absences, incapacitation of per:onnel, or other emergencies, the following persons will be res-ponsible in the order listed for al~ plant activities: Plant Manager Assistant Plant Manager Amend. 50 @ 13.1-7 June 1979 268 319
Plant Operations Supervisor Plant Engineering Supervisor Shift Engineer 13.1.2.2.1 Plant Manager The Plant Manager has direct responsibility for all plant activities. He is responsible for safeguarding the general public and plant employees from hazards associated with the operation of the CRBRP through implementation of the TVA hazard control standards and require-ments, applicable DOE and NRC rules and regulations, and plant procedures, 50 and for adherence to all requirements of the operating license and technical specifications. He receives direction and supervision from the Chief, iluclear Generation Branch, and staff assistance from the Division of Power Production Central Office. 13.1.2.2.? 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. 13.1.2.2.3 Plant Operations Supervisor The Plant Operations Supervisor is responsible for the safe and efficient operation of the plant in accordance with the operating license, technical specifications, and aoproved procedures and TVA hazard control standards and requirements. He is responsible for the preparation and maintenance of up-to-date operating procedures and the preparation of operating records. He is also responsible for operator training programs and operating personnel schedul x e d 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 Operations Supervisor The Assistant Plant Operations Supervisor assists the Plant Operations Supervisor in reviewing, coordinating, and planning the activities of the plant Operations Section. In the absence of the Plant Operations Supervisor, he assumes the responsib. :ities of that position. 13.1-8 Amend. 50 June 1979 268 320
13.1.2.2.5 Plant Engineering Supervisor The Plant Engineering Supervisor serves as supervisor of the Plant Engineering Section and as a staff 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, safeguards, and licensing require-ments and a knowledge of the control systems used in a nuclear plant. He is re consible for analysis of the performance of the reactor and turbine cycle and associated equipment during the test, startup, and operation of the plant. 13.1.2.2.6 Pl_a re "aintenance Supervisor The Plant Maintenance Suporvisor ?s responsible for all mechani-cal and elec* rical mainter,ance work and inwections in the plant. He is responsible for maintaining safe working conditions for his employees and for their adherence to saf; working practices. He is assisted in his work by two Assistant Supervisors with experience in mechanical and electrical maintenar.ce. He is also assisted by foremen of the various crafts within the organization and engineers who will be assigned to the plant as the workload demands. The Plant Maintenance Supervisor must have a thorough knowledge of the operation and maintenance of all plant mechanical and electrical equipment. 13.1.2.2.7 Assistant Plant Maintenance Supervisors The two Assistar.t Plant Maintenance Supervisors--one a mechani-Cal specialist, the other an electrical specialist--assist the Supervisor in planning, coordinating, and directing the maintenance work and inspection in the plant. 13.1.2.?.8 Health Physicist The Health Physicist is the onsite supervisor of the Health Physics Unit of the Radiological Hygiene Branch and is responsible for direction of an ddequate program of health physics surveillance for all plant operations involving potential radiation hazards. He keeps the Plant Manager informed, at a'l times, of radiological conditions related to personnel exposu m and potential contamination of site and environs. His duties include t.aining and supervising health physics technicians; planning and scheduling monitoring and surveillance services; maintaining current data files on radiation and contamination levels; personnel 13.1-9 Amend. 50
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exposures, and work restrictions; and ensuring that operations are c .rried out within the provisions of developed radiological hygiene and pro-cedures. He provides monitoring assistance and technical advice tr> plant operations and provides assistance to the medical staff in emergencies where radiation and contamination hazards are involved. 13.1.2.2.9 Supervisor. Nuclear Plant Quality Assurance Staff The Nuclear Plant Quality Assurance Staff Supervisor serves as supervisor of the nuclear plant quality assurance staff and as a staff advisor to the Plant 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 pla' proceduras and instructions. He also advises the Plant Manager of failures of plant equipment to meet technical speci-fication requirements and other nonconforming aspects of operations. He is responsible for the inplant quality assurance / quality control training programs. 13.1 ?.2.10 Safety Engineer The Plant Safety Engineer provides conwltation to plant management on all fire safety matters; coordinates and evaluates testing, maintenance, and repair of all fire-related equipment and systems; con-ducts 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 emergenci7s as applicable. He reviews pre-fire plan and erg ncy pl nning documents and coordinates fire safety matters as 50 required with Safety Engineering Services at the Central Office. s 13.1.2.3 Shift Crew Composition g Normal Operations The Shif t Engineer on duty is in direct charge of the plant including startup, operation, and shutdown cf the reactor and turbo-generators. He may institute immediate acti)n in any given situation to eliminate difficulties or remove equipmert 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 Shif t Engineer is under the inmediate super-vision of the Shift Engineer. He follows established procedures in doing his work. However, if a particular situat'on is not covered by a procedure, he may seek advice from the Shif t Encinere; or, if the situation is critical, he may use his own judgment w prevent damage Amend. 50 13.1-10 ,a M
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to equipment, injury to personnel, or undue radiation exposure of perscnnel. He performs operatioris in the electrical switchyard, diesel generator building, and other areas inside and outside the main power-house structure. 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 from those ir.structions except as directed. He performs assigned routine 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 shift, the Radiochemical Analyst is under the functional supervision of the Shift Engineer. These dt. ties 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 shift, the Health Physics Technician is under the functional surt Msion of the Shif t Engineer. He performs routine radiation sur', m , personnel monitoring activities, and other assigned duties. He keeps the Shif t Engineer informed of radiation hazards and performs special surveys as requested. During the five-yea" demonstration test period, a Technical Engineer will be assigned to shift who is under the supervision of a Shift Engineer. However, he eceives his technical guidance from the Supervisor of the Plant Engineering Section. He has the responsibility of providing technical assistance to the Shift Engineer in performing the demonstration and test portions of the program. 13.1.3 Qualification Requirements for Naclear Plant Personnel All personnel at the CRBRP will be required to obtain and maintain qualification standards equal to or better than those specified in ANSI N18.1-197i. The personnel selection and training program that assures fulfillment of these qualification requirements also satisfies 50l NRC Regulatory Guide 1.8. Specific minimum qualifications for all those personnel discussed in Section 13.1.2 are given below. 13.1-11 Amend. 50 June 1979 0 m,1
13.1.3.1 M nimum Qualification Requirements Plant Manager At th time of initial core loading or appointment to the active position in the licensed plant, the Plant Manager shall have 10 years of resporisible power plant experience of which a minimum of three years shall be nuclear experience. A maximum of four years cf the remaining seven years of experience may be fulfilled by academic training on a one-for-one time basis. This academic training shall be in an engineeririg or scientific field generally associated with the production of power. The Plant Manager shall have acquired the experience and
. training normally required for examination by NRC for a SR0 license 50l whether or not the examination is taken.
If the Assistant Plant Manager meets the nuclear plant 50l experience and NRC examination requirenents established for the Plant Manager, the requirements of the Manager may be reduced so that only one of his 10 years of experience need be nuclear plant experience, 50l and he need not be eligible for NRC examination. The Plant Manager or the Assistant Plant Manager should have a recognized baccalaureate or higher degree in an engineering or scientific field generally associated with power production. Assistant Plant Manager At the time of initial core loading or appointnent to the activo position in the licensed plant, the Assistant Plant Manager shai have a minimum of eight years of rcspo nsible power plant experience of which a minimum of three years shall be nuclear plant experience. A maximum of four years of the renaining five years of the power plant experience may be fulfilled by satisfactorily completing academic or related technical training on a on-for-one time basis. A degree in science or engineering is desirable. He or the Plant Manager shall 50l be capab' > of fulfilling the requirements of an NRC SR0 license whether or not tne 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 Operations Supervisor At the time of initial core loading or appointment to the active position in the licensed plant, the Plant Operations Supervisor i shall hold an NRC SRO license and shall have a minimum of 8 years of 501 responsible power plant experience, of which a minimum of 3 years shall be nuclear plant experience. A maximum of 2 years of the remaining 13.1-12 Amend. 50 June 1979 n? 2(38 ,303
5 years of power plant experience may be fulfilled by satisfactory 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 Engineering Supervisor 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 8 years of responsible power plant experience or applicable industrial experience of which 2 years shall be nuclear plant experience. He should have an engineering or science degree. Plant Maintenance Supervisor 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 7 years of responsible power plant experience or applicable industrial experience, including at least one year of nuclear plant experience. A maximum of 2 years of the remaining 6 years of power plant or industrial experience may be fulfilled by satisfactory completion of academic or related technical training on a one-for-one time basis. He further should have familiarity with nondestructivc testing and maintenance of sodium containing components, craf t knowledge, and an understanding of electrical, pressure vessel, and piping codes. Supervisor, fluclear Plant Quality Assurance Staff At the time of initial corc loading or appointment to the active position in the licensed plant, the Sapervisor of the fluclear Plant Quality Assurance Staff shall have 7 years of responsible power plant experience or applicable Quality Assurance experience of which a minimum 2 years shall be nuclear vower plant experience. He shall be a graduate with a degree in er , aeering. A maximum of 2 years of the remaining 5 years of power pi nt 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 Power Quality Asturance Program as defined by the Office of Power Quality Assurance Manual. Amend. 50 @ 13.1-13 June 1979
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The Safety Engineer The Plant Safety Engineer shall have a sound understanding and thorough techniccl knowledge of safety and fire protection practices, procedures, standards, and other codes relating to electrical utility operations. He shall: be able to read and understand engineering drawings; possess an analytical ability for problem solving and data analysis; be able to communicate well both crally ana in writina; be able to write investigative reports and prepare written procedures; have the ability to secure the cooperation of management, employees, and groups in the implementation of safety programs; and be able to conduct safety presentations for supervisors and employees. He shall have experiencc in safety engineering work at this level or have ; years experience in safety and/or fire protection engineering. It is desirable that the incumbent be a graduate of e accredited college or university with a degree in industrial, mechanical, electrical, or 50 safety engineering or fire protection engineering. Health Physicist The plant Health Physicist chal! meet the qualifications as 50, specified in NRC Regulatory Guide 1.8. Plant Operations Section Employees At the time of initial core loading or appointment to the active position in the licensed plant, the Assistant Plant Operations Supervisor shall have a minimum of 6 years of responsible power plant experience, of which a minimum of one year shall be nuclear plant experience. A maximum of 3 years of the remaining 5 yees o' power plant experience may t'e fulfilled by satisfactory completion of academit 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 the requirements of TVA's Training Plan for Operators and have a high school diploma or equivalent and six years of responsible 50l nuclear power plant experience, plant exparience, andofhewhich a minimum shall hold of one an NRC SRO year shall be li ense. At t. e time of initial core loading or appointment to the active position in the licensed plant, the Assistant Shif t Engineers shall have fulfilled the requirements of TVA's Training Plan for Operators and have a minimum of a high school diploma or equivalent and five years of responsible power planc experience, of which a minimum of one year shall be nuclear plant experience, and he shall 50 hold an NRC SRO license. 13.1-14 Amend. 50 June 1979 2b0 32U
At the time of initial core loading or anpointment to the active posit. ion in the licensed plant, the Unit Operators shall have fulfilled tne requirements of TVA's Training Plan for Op_erators a high school diploma or equivalent and two years of power plant experience, of whie.h a minimun of one year shall be nuclear plant experience. The latter, for operators with no previous nuclear experience, will consist of a bc sic nuclear course, plant technolog, course, plant systems and operators course, and control board experience. He shall have an fiRC EO R0 license. At the time of initial core loading or appointment to the active 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 requirements of TVA's Training Plan for Operators, completed a basic nuclear course and plant systems course, and had several months of onsite plant familiarization. This position 50 does not require an tiRC R0 license. Plant Engineering Section Employees At the time of initial core loading or appointment to the active position in th licensed plant, the Instrument Engineer shall have a bachelor's degree in science or engineering and minimum of one year's experience in the field of instrumentation. Six months of this experience shall be in nuclear instr:rentation and control. At the time of initial core loading or appointment to the active position in the licensed plant, the Chemical Engineer shall have a bachelor's degree in science or engineering and a minimum of one year's experience in radiochemistry. At the time of initial core loading or appointment to the active positiori 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 measurements, core heat transfer, and core physics testing program. The Radiochemical Analysts and other engineering aides shall be high school graduates with a minimum of two years' experience in their respective fields. Each Instrument Mechanic shall have a minimum of three years' experi2 nce in his craft and shall be a skilled journeyman. Plant Maintenance Section Employees At the time of initial core loading or appointment to the active position in the licensed plant, the Assistant i'aintenance Super-visor shall have a minimum of five years of responsible power plant 13.1-15 2~ b 8 .- J ,/ ,.- Amend. 50 June 1979
experience or applicable industrial experience, including at least one year of nuclear plant experience. The position requires familiarity with nondestructive testing, craf t knowledge, and an understanding of electrical, pressure vessel, and piping codes. Each TVA craftsman shall be a skilled journeyman. These experienced journeymen will predominantly be transferees from other TVA genarating plants and installations. Craftsmen shall have a minimum of ' ree 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-mangement council, nonnally requires in excess of four years for completion. The program requires assignments designed so that he will develop skills equal to the recognized jourrieyman standard.
Related classroom and correspondence lesson assignments provide the technical information needed in the actual work being done on the job. Preoperational testing will be carried out for all safety-related fire protection systems in accordance with the requirements of nationally recognized fire codes and standards. The test director will be an experi 2nced system-oriented test engineer assisted by one or more engineers from the reactor manufacturer, the architect-engineer, and/or the constructor. The engineer responsible for eveloping and assisting in implementing the fire protection program shall have a B.S. degree in engineering and shall have 3 years or more of experience working on mechanical engineering projects. He must have acceptance testing experience in fixed fire protection systems and have a general knowledge of nationally recognized fire codes and standards. 13.1.3.2 Qualifications of Plant Personnel The positions listed in 13.1.3.1 have not yet been filled. These positions will be filled as indicated in Figure 13.2-1. A.nen-. 50 June 1979 13.1-16 7m
G O TABLE 13.1-1 TECHNICAL SUPPORT
SUMMARY
MANYEARS OF EXPERIEtiCE G ^ U R TOTAL UTILITY EXPERIEtiCE NON-UTILITY EXPERIENCE PERSONNEL tiUCLEAR POWER OTHER UTILITY NUCLEAR POWER OTHER Ef4GINEERING FIELD EXPERIENCE FIELC FIELDS Reactor Engineering Staff 91.25 6.0 41 7 Nuclear Generation branch - 23 Preoperational Test Staff 180.25 130.25 43.75 95 Nuclear Generation Branch - 40
.C Nuclear Operations y Staff 24 25 0 0 C Nuclear Generation Branch - 3 Chemical Section 63.25 82.50 37.50 44.00 Plant Engineering Branch - 12 Mechanical Section 31.50 161. 5 5 53 Plant Engineering Branch - 12 cg Structural Section 40.75 148.25 6.50 25.00 Plant Engineering lg p Branch - 11 ]
g m$ Instrument and , Controls Section r _, 106.85 1 21.5 42.5 17'. 5 Plant Engineering so 50 Branch - 61
TABLE 13.1-1 continued I MANYEARS OF EXPERIENCE . A ~ ~ 7 TlifAL ITTlLI'TTEXPERIENCE tion-UTILITY EXPE'PIIENCE R PERSONNEL t;UCLEAR POWER t;UCLEAR POWER FIELD lOTHERUTILliY EXPERIENCE FIELD l0THERENGINEERING FIELDS l Special Projects Section 0 61 0 49 Plant Engineering Branch - 11 Test Section 29 76.25 1 69.5 Plant Engineering Branch - 22 h Economy & Statistical Section 0 13 0 0 h Plant Engineering Branch - 8 TOTALS FOR 211 TECHNICAL SUPPORT PERSONNEL 566,85 825.25 177.25 514.0 50 N Cs CD D ma o. ah L I D, C. 2 wo 9 9 9
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13.2 TRAINING PROGRAM 11.2.1 P rog ram _De s c_r ip t i on The 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 ano ef ficient nanner,
- b. To assure that the CRBRP is operated in accordance with 50l NRC regulatory requirenents and guidelines.
- c. To assure that all training is formally documented.
50l d. io w et er exceen NRC licensing requirements. In achievir.g 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 Chief, Nuclear Generating Branch, af ter being approvm1 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. The ef fectiveness of the training program is s valuated by the 50 l performance of employees on TVA and NRC examinations in c arrying out their assigned duties. In addition, periodit audi ts of .he training program are performed by designees within the Office of 'ower , but outside the Division of Power Production. 13.2.1.1 Progran Content At the time of manning the CRBRP, TVA should have hichly 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. l Those positions at the CRBRP which require an NRC licensed 50 SRO shall be filled with personnel who have or are eligible to sit for an NRC SRO license on a comercial size light-water reactor. Other positions shall be filled with personnel within the TVA organization as available and selected from competent applicants from outside. All CRBRP personnel will be given comprehensive training to produce pcr-sonnel who have that combination of education, experience, and skills commensurate with their level of responsibility which provides reasonable assurance that decisions and actions during all normal and off-nornal conditions will be such that the plant is operatec in a safe and efficient manner. Amend. 50 aune 1979 13.2-1 Ml
The TVA L:udent operator training program and replacernent training at operating TVA nuclear plants shall ensure no loss , operator etfichocy at those plants because of transfer of personnel to the CRBRP Individual training needs shall be established by caref ully exanining the individual's experience and previous training and comparing these with the job requirenents. The f onnal program to be 50l pr ovided f or candida tes seeking NRC RO license er NRC 5RO license as well at the length of each aspect of the proqcam is aiscusv.d in the f ollowing paragraph and depicted in Fiqure 13.2-1. The training program shall consist of the following phases'. Basir Nuclear Courses Plant Technology and Specialist Training Peactor Operations (LWR and FBR) Actual Fast Reactor Training Onsi te Work-Study Program Nuclear System Special Training to include Sodium Technology A final training plan based on the needs of the individual staf t memtwrs will be prepared af ter personnel selection and shall be included in the Final Safety Analysis Report. 13.2.1.2 Con ed_i n_a_t i on_ w i t_h P reope ra t i ona l iests and Fuel Loading figure 13.2-) presents a proposed training schedule fo, the CRBRP which u tisfies the requi r en ents of ANSI NIH. i-1971. It is planned that the following personnel shall be licensed in accordance with the requirerents of 10CFR55 before initial fuel loading: Operations Super-visor, at least five Shi f t Enqineers, and at least five Assistant
ift Engineer' The Plant Manager or the Assistant Plant Manager sn<ll obtain the traininq required for an SRO license. It is planned 50 to 1btain R0 licenses for at least five Unit Operators during atartup testing of the plant. The various phases of training available are outlined below, along with descriptions of personnel participation in each phase 11 13.2.1.3 P r a c_t i_c a_1 _Re a c t o r Op e_ra t i_o n_
Practical training a t TVA's Browns Ferry, Sequoyah, and Wa tts Bar and at an operating sodium-cooled fast reactor is anticipated in order to p ovide the experience required for applicants for cold licenses. e pmans who will inidalh oMain W licenses shall participate 30linthistraining Trainir.g requirements shal' be individually deter-mined and training will be supplied to fill these needs. Amend. 50
~<n 'J ne 1979 13.2-2 /00 lb
The program involving the actual participation in the operation of a sodium-cooled fast reactor to gain experience in the areas of liquid metal systems and the characteristics and performance of fast reactors shall be integrated into this phase of the total training program for licensable personnel and management personnel, 13.2.1.4 Reactor Simulation Training A simulator for the CRBRP is not planned, therefore, simulator 5d training for candidates seeking NRC R0 licenses or NRC SR0 licenses is nmt included as a part of the nuclear training program. 13.2.1.5 Previous Nuclear Training Figure 13.2-l presents the tentative training schedule showing the relation of the training to the plant schedule for construction, testing, and operation. The actual training schedule will be dependent upon the bac' ground and expecience of the inGividuals 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 Training Bajic Nuclear Cour;es 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 instru-mentation, and reactor materials, with special reference to fast reactors; reactor core thermal-hydraulic characteristics, suen 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, will receive more abbreviated instruction in nuclear fundamentals as part of their on-site specialist ' raining. The prereqJisite qualifications for participation in the basic nuclear courses and succeeding phases of the operator training program are:
- a. High school education or equ1 valent,
- b. Knowledge of mathematics through high school algebra.
- c. Ability to use a slide rule.
- d. gemonstration during pr evious experience of a desire to learn and produce qual . cy work.
Amend. 50 13.2-3 June 1979 n .7
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- e. Demonstration by past perfonnance of maturity, good judgment, and i 'gh moral character.
- f. Satisfactory completion of medical examination.
- g. Satisfactory performance of tne work in his present classification.
Pl_ ant Technolony and Specialist Training A design lecture series covering the function, design and opera-tion of nuclear systems and components shall be conducted at the plant SOj si te. The persons requiring NRC SR0 licenses shall participate in this plant technology training. In addition, the plant management and engineers shall also participate. Various specialirt training, consisting of work-study assign-ments, shall bt conducted for plant engineers, technicians, and mai ntenance personr.el . This training shall be specifically tailored to the individual's needs. It is planned to use the Browns Ferry, Sequoyah, and Watts Bar facilities for as much of this training as feasible. Specialist training in LMFBR technology is planned for the Supervisors, the Nuclear Engineer, and certain central office staff engineers. Specialist training is planned to varying degrees in instrumentation and controls including process computer progranning and maintenance for the Instrument Engineer, several Instrument Mechanics, Plant Engineering Supervisor, and certain Central Office staff engineers. Training assignments at TVA nuclear plants of varying lengths are planned for plant staff personnel such as the Instrument Engineer, P2 actor Engineer, Mechanical Engineer, and Chemical Engineer. After completion of this specialist training, the appropriate personnel shall organize and conduct necessary specialist training onsite for the Chemical Technicians, Maintenance personnel, Instrument Mechanics, and others. Onsi te Work-Study Program This phase, which begins before fuel loading, integrates personnel into their plant assignments. It is conducted under the direction of TVA supervisory personnel. During this period, plant personnel participate in the preparatic of procedures and manuals, preoperational testing, preoperational checkout of the operating procedures, initial fuel loading, and initial startup program. The 5d applicants for NRC R0 and NRC SRO licenses will participate in further training and examination preparation related to obtaining SQ the required NRC licanse. All plant personnel will participate in a plant indoctril.. tion and radiation-protection course. 13.2-4 Amend. 50 June 1979 io , , .
. I
Fire Brioade 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, a'sistant shif t engineer, shi f t engineer, assistant operations supersisor, and optrations supervisor training. The objective of this t.aining is to ensure that plant fire brigade members, leaders, and other plant employees performing fire-related functions receive comprehensive first aid firefighting instructions and application tllat wlll be instrumental in the prevention, control, and suppression of plant fires. This training will be updated or revised 50 as necessary to ensure tha t current, ;cceptable practices are included. 13.2.1.7 Training Programs f or Non-Licensed Personnel TVA, on a continuing basis, plans and administers training programs for the professional and managerial development of its employeos. Relationships are maintained with both local and state educational institut'ans as well as with the vendors of various items of equipment. Advantage is .aken of appropriate seminars, specialized courses, and training activities of fered by these groups to keep employees abreast of new developments in po,ver production and safety. 13.2.1.8 Gene ra l Employee Training Personnel with specific duties and responsibilities in the plant shall receive instruction in the performance of these duties and responsibilities. All persons having unescort ?d access to the plant areas shall have completed eithe~ (1) intensive nuclear training, which will include radiation protection techniques and the site em gency plan, or (2) a brief plant indoctrination and radiation protection course which will include discussion of plant organization and layout, controlled zones, radiation and contanination hazards, exposure limits and controls, elementary health physics, and pertinent sections of the site emergency plan. When oersons who have not cornpleted either (1) or (2) above enter the plant areas, they will be escorted by an employee who has rc:eived training in radiation protection and plant emergency procedures. A. permanent plant personnel shall receive training periodi-Cdlly in the plant's fire protection polic:es. Tempo o j personnel 50 m msponWlities in fire protection will also oe included. Training and indoctrination relating to quality assurance will be provided to all employees as applicaDlc in conformaace to RDT-F2-2, Section 7.3.2.
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13.2-5 ' Amend. 50 June I C 9
Periodic retraining of plant personnel regarding radiation hazards with emphasis on individual actions, will be conducted at monthly meetings attended by all available plant personnel, and at short weekly meetings held within the various plant groups. The emergency plan will be discussed at least once annually in these mee ti ngs . 13.2.1.9 . Responsible Individual The individual responsible for conducting and adninistration of the nuclear power plant training program is the Assistant Plant Manager. The J.lant Quality Assurance Staf f Supervisor shall be responsible for developing and directing the Nuclear Plant f Jal i ty Assurance Progr'm which complies with RDT-F2-2. 13.2.2 Retraining Program This information will be included in the FSAR. 13.2.3 Rjplacement Traininc This information will be included in the FSAR. 13.2.4 Records _ 13.2.4.1 TVA Of ficial records of employee qualitications experience, training, and retraining of each member of the plant organization are maintained in the of ficial TVA Personal Hb +,rv Record (PHR) by the Division of Personnel. The PHR provides n :ndardized arrangement, the information officially recognized in retJrding and supporting ernployee s taus. The PHR is inaintained in current and accurate status and is contri . led as to availability. The material admitted to this record is restricted to items for which authenticity has been cc. firmed through established procedures; e.g. , official TVA forms, cigned statements from the employee, management representatives, etc. 13.2.4.2 Plant 50 Records supporting requests for NRC SRO and NRC R0 licenses are maintained in the plant master file. These records include training courses attended, retraining classes, number of reactor startups, dnd other information necessary to ensure that training requirements have been met. Some of these records are duplicated in the PHR. A training file for each member of the plant organization is maintained in the plant master file. Information regarding partici-pation 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. Amend. 50 3 -6 June 1979 2, 6 8 3D
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13.3 EMERGENCY PLs.iNING 13.3.1 General TVA's emerger.cy plans contain the precautionary planning, delegation of authority and responsibility, and plans of action to protect the public, plant employees, and equipment in case of unusual incidents. As specified in 10CFR50, Appendix E, these plans are for use at the local level for the control of general emergencies such as fire, personnel injury, tornadoes and high winds, and incidents that could result in tne rt; ease of a significant amount of radioactivity. The TVA Radiciogical Emergency Plan (REP), for the CRBRP will contain the overall TVA REP, the Nuclear Emergency Medical Assistance Plan, and the CRBRP Annex. The CRBRP Annex will contain four documents. They are the (a) Division of Power I'roduction REP, (b) Site REP, (c) En-rivons Emergency Plan (EEP), and (d) State of Tennessee REP. These documents are briefly described below. lne actual TVA REP for the CRBRP will be submitted as a separate document along with the Final Safety Analysis Report.
- a. The TVA REP is designed to handle all radiological emergencies which might occur within TVA. During a nuclear emergency at a plant site, the Central Emer-gency Control Center (CECC) staff will function to provide assistance as necessary to the site and division emergency organizaticis and will pr] vide all information requested by outside agucies.
- b. The Nuclear Emergency Medical Assistance Plan will outline all arrangements which have been made for medical services which may be required for the CRBRP employers or others affected by the emergency.
- c. The CRBRP Annex will contain the four following documents:
50 l 1. The Division of Power Production (P Prod) REP requires automatic staff actions to prov' e required assistance #or the site by aler .ng support facilities, concluding arrangements with civillan support facilities, and providing any support requested by the plant. The major assistance provided by the division errergency staff will be to the plant itself although the staff will provide personnel service s as required by state and local agencies. The division emergency staff will also coordinate the efforts of other divisions within TVA. @ g 7fC - Amend. 50 J ',r June 1979 13.3-1
- 2. The Site REP will deal with control of the emergency within the site boundaries.
- 3. The Environs Emergency Plan (EEP) will deal with the emergency beyond the sita boundary.
- 4. The State of Termessee REP will provide the support of state organizations in the event of a nuclear emergency and is principally concerned with the well being of area citizens. This plan will work hand in hand with th CRBRP EEP.
13.3.2 Emergency Oraanization The normal shif t operating crew provides the nucleus of the plant's emergency organization. The shift crew has an adequate number of personnel with the authority to take required immediate action in any emergency. The plant emergency organization is headed by an Emergency Director. The Shif t Engineer is responsible for declaring an emergency and acting as Emergency Director until relieved by the Plant Manager or a designated alternate from the plant staff. After relief, the Shift Engineer remains in charge of detailed inplant opcrations. The shif t organization is supplemented by predesignated individuals f rom the remaindte of the plant staff af ter notification by telephone or messenger. The plant emergency organization has pre-assigned duties and responsibilities and is trained to perform all actions that may be necessary to cope wi h the emergency and to implemert the emergency plan. In addition to the plant emergency organization, the unaf fected plcnt staff could provide additional personnel tu assist as necessary. In the event of an emergency involving the possibility of danger to the public or the offsite environment, the plant Emergency Director notifies TVA's operating duty specialist who notifies the Central Emergency Control Center (CECC) Director. The CECC organization consists of TVA management personnel from various TVA divisions and offices and is located in Chattanooga. The CECC has the authority to make arrange:.ents and expend funds as necessary to protect tM environ-ment from the adverse effects of an emergency. They coordinate TVA offsite activities and work with various other Governmental emergency groups. The members of the CECC staff are predesignated, aware of their responsibilities, and conduct periodic drills to maintain a high degree of readiness. 50 1 The P PROD emergency organization is also notified by the Plant Emergency Director and provides additional manpower as required to augment the plant organization. The personnel may come from other 50 lTVA nuclear plants, the P oROD Central Office, or the P PROD Service Shops Section, depending on the nature of the emergency and the E3l disciplines required. The P PROD emergency organization will also provide technical support groups for emergency planning and recovery operations. Amend. 50 13.3-2 2 b O 3 [ }. June 1979
As required, an environs emergency team is dispatched to the vicinity of the emergency to conduct TVA's offsite moni+nring activities and to work with monitoring groups from other agencies. Conmunication networks are adequete to handle predicted emer-gencies. These co amunications facilit',es include:
- a. Public Address Intra-Plant Connunication (PA-IC)
This system provides primary comunication throughout the plant. It also provides a warning throughout the plant of fire. high radiation, and building eu :ating chrough the use of a mul ti-tone genera tor,
- b. Private Automatic Exchange (PAX)
This system provides conmunication re chs through tie lines and interfacing circuits of the folltwing:
- 1) Microwave comunication to the TVA-wide direct-dial system
- 2) The page channel of the PA-IC syste
- 3) The Powerline carrier communication system.
It also incorporates a dial-in executive right of way.
- c. Manual Telephone Switchboard at the electrical control desk which is connected through tie-lines and interfacing circuits to the following:
- 1) Microwave connections to the TVA-wide direct-dial system
- 2) Pooerline carrier communication system
- 1) The PAX switching equipment which provides full bene-fits of the PAX system
- d. Bell System commercial te'.ephone service - this system is not connected to any of the communications system within the plant and is not a part of the TVA-wide communications system. However, this service is available at key locations in the plant.
- e. Maintenance Communicatior s Jacking System (MCJ) - this system consists of sound powered headsets / microphones and jack stations provided to facilitate the testing and calibration in the maintenance discipline.
50 Amend. 50 13.3-3 ]g ,, June 1979
- f. Radio equipment including a VHF base station, portable radios (walkie-talkies), vehicle-mounted radios (security officers) and radios located at environmental sampling stations, 13.3.3 Coordination with Offsite Groups TVA has agreements with other Federal agencies through the Interagency Radiation Assistance Plan to assist in the evaluation and 50lcontrol et any radiological emergency. These agencies include DOE's Savannah River and Ock Ridge Operacions Offices. The CECC staff will request assistence from these outside agencies as required. The CECC 50lstaff is also responsible for notification of NRC's regional office of the Division of Compliance.
Agreements have been made with the Tennessee Departments of Public Health, Civil Defense, Agriculture, Public Safety, and Conser-vation to provide planning for emergencies at TVA nuclear facilities. This planning includes evacuation arrangements, traffic control, and support from civil defense agencies. The Tennessee Department of Public Health will be notified anc will coordinate assistance from other state agencies as required. TVA maintains liaison with tarious agenc.es of county and municipal governments, particularly with respect to the availability of emergency services. The State Department of Public Health infornis these agencies of actions to be taken under their respective statutory authority and assists them in developing emergency procedures. TVA may call upon these agencies directly for # ire and police protection. TVA will meet with representatives of the various county and municipal o vernments to discuss their involvement in the Radiological Emergency Plan. TVA will provide training for local fire departments in radio-logical hygiene practices and recognition of radiological hazards. The attached Table 13.3-1 lists the organizations that will be participating in the CRBRP Emergency Plan. Arrangements will be made with a local private ambulance service to provide emergency service as required to the plant and aff(cted areas in the event that such service it required. Agreement will also be culminated between TVA and a local hospital to provide emerge.'cy treatment of irradiated or contaminated patients as requireo. TVA will assist in training ambulance attendants and hospital pesonnel in this type of treatment and will ensure that adequate equipment is made available. Agreement has also been made with the Oak Ridge Associated Universities Radiation Emergency 50 Assistance Center / Training Site #or emergency treatment of severely contaminated or irradiated personnel. Amend. 50 13.3-4 June 1979 0 ?) I$ 'I
13.3.4 Protective Action Levels Protective action levels are established oependinl on the nature of the emergency and the value of continuously monitored variables. These p;otectivc action IcVels defined in the REP include ttose which may affect only a local area of the plant or a small number of employees as well as these which could possibly involve the public in unrestricted areas. The protective action levels are ba ed on control room indications of continuous ly monitoring variables such .at the operator can quickly determine the nature of the emergency. Local emergencies may also be detected by .he shif t crew during routine olant tours and inspections. 13.3.5 Protective Measures For each protective action level, definite protective measures or actions are specified in the emergency plan. The Site REP contains such cctiors which include mustering personnel in preassigned assembly 'reas, informing personnel of the nature of the emergency, confirming the in'.; cation of the emergency, notifying the CECC Director, mustering the plant emergency organization in the Emergency Control Center, con-ducting radiation surveys, locating missing personnel, evacuating non-essential personrel from the site, accounting for any visitors, and limiting access to the plant areas. Personnel safety is the prime consideration in all protective measures. Protective measures become more detailed and extensive with the increasing protective action levels. The emergency plar, contains site drawings designating assembly areas and evacuation routes. Protective measures, including evacuation, for persons living outside the plant boundary are included in the EEP and State REP and are expected to be required only a#ter evaluation of plant conditions and effluent release rates. However, immediate protecti'.e measures will be specified based on previously oetermined dose rates, population distributions, meteorological conditions, and plant conditions that could cause site boundary conditions requiring action. These protective measures include preplanned e/acuation procedures and routes, reassembly points, traffic control, and public announcement. 13.3.6 Review and Updating The Plant Operations Review Committee will periodically review and update the CRBRP Site REP. All holders of these plans will acknowledge in writing, receipt of all changes. Amend. 50 13.3-5 June 1979 268 .5.o.
13.3.7 Medical Supp_ ort The emergency plan includes a description of the "edicai faci-lities at the plant and the arrangements made with other facilities to provide additional suppcrt. The plant medical facilities include a treatment area consisting of an emergency room, trcatment room, bedroon, physiotherapy room, and waiting room. A full-time nurse in on duty during the day shi ft. A complete stock of medical supplies and first-aid equipment is available. One ambulance is maintained at the site. Medical consultation is available from TVA doctors in Ch3ttanooga and other areas. Members of the plant energency team are trained in first aid. Arrangements will be made with a 1ocal hospita1 and with attending ohysicians for the emergency treatment of contaminated, injured, and exposed individuals. The Oak P,idge Associated Universi-50 ltics Radiation Emergency Assistance Center / Training Site has agreed to provide treatment to severely ontaminated or exposed individuals. Arrangements will be made with a local private ambulance service to provide emergency service as requit ed te the plant and affected areas in the event that more than one ambulance is required. 13.3.8 prills Periodic drills will be conducted by the olant staff on the
' ant emergency plans. Personnel will assemble, b.: accounted for, and 6 repare to assume their preassigned duties. Contact will be made with concerned persons outside the plant organization to confirn the adequacy of communications facili ties.
On an annual basis, a TVA-wide drill on the Radiological Emergency Plan is conducted. The various members of the emergency staff will assemble, assure their duties, and get in touch with the various outside organizations, identifying the action as a drill. 13.3.9 Training Each person having unescorted access to the plant will have received either intensive nuclear tri :ning which included the energency plans action or a brief plant indoctrination and health physics course which includes pertinent sections of the em< raency plan. Specific training will be conducted for individuals amigned to the plant emergency organization. This will include first-aid training, radio-logical hygiene training, decontamination training, and training in the emergency proceGu es. Amend. 50 J"" l979 13.3-6 268 K6
TVA wiil assist in .~'viding training in decontamination and treatment of contaminated pouents to the staff of the local hospital and the conviercial ambulance service. 13.3.10 Recovery and Reentry The emergency plans provide for the development and implemen-tation of detailed recovery and reentry plant based on evaluation of conditions existing at the time. Recovery and reentry will be a deliberate, thoroughly planned evolution ard will be reviewed by the Plant Operations Review Comnittee depending on the natuce of the emergency. 13.3.11 Implementation Operating instructions, promulgated in the plant operating manual, are used to control plant operations during normal operating
'onditions. Abnormal operating instructions and emergency operating instructions are used to specify the manipulation of controls of the plant during conditions requiring protective measures to be taken to place the plant in a safe condition. The abnormal and emergency in-structions contain assignments of responsibility for the performance of specific tasks not ntherwise established by plant practices and instructions.
Plant instrumentation indicatiens requiring implementation of emergency and abnormal operating instructions are specified in these instructions. Protective action levels, also based on plant instrumen-tation indication, requiring implementation of the Radiological Emergen-cy Plan for protection of personnel and the environment are specified in the emergency plan. Specific actions required of offsite support groups are delineated in the TVA-wide Radiological Emergency Plan and in a Division of Power Production Emergency Plan. Instructions for medical treatment and handling of contami-nated and exposed individuals are contained in the Plant Emergency 50 Plans manual and a TVA Nuclear Emergency Medical Assistance Plan. Equipment requirements, inlcuding communications equ;pment, for impiementation of the plant emergency plans c a contained in these plans. Stor >qe and calibration requirements are also specified. Alarm signals are de3c-ihed in the respective emercency plans. Insu. uctions for restoring the emergency situation to normal, from the standpoint of hazard to personnel, plant safety, and the environment, are contained in the emergency olans and the emergency and aonormal operatinc instructions. Instrv cior.s for repair of plant equipment or structures wili be prepared af ter evaluation of the damage or malfunction invol ed. Amend. 50 June 1979 1 . -7
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13.3.12 Radiological Analysis A radiological analysis of the facility design features, site layo , and site location with respect to considerations of surroundings in con pliance to 10 CFR 50, Appendix E has been conducted. The findings of this analysis are listed in this section. 13.3.12.1 Projected Ground Level Doses Plots showing projected ground level doses, for both whole body and thyroid, resulting from the most serious design basis accident analysis is depicted in Figures 13.3.1 through 13.3.4. These provide, respectively, the elapsed exposure times to reach specific bone, lung, thyroid, and wnole body doses as a function of downwind distance based on exposures resulting from the Site Suitability Source Term (SSST). The use of the SSST is conservative since it envelopes the most serious design basis accident analyzed in the PSAR. Table 13.3-2 summarizes the data and parameters used in the analysis. The information piovided in the Table fully describes the basis of the analysis. 13.3.12.2 Accident Assessment, Warning and Evacuation Times 13.3.12.2.1 Assessment The time requireu for ^e initial accident assessment of the most serious design basis accident may required 15 minutes. This time is an esti. rate based on the operation of the reactor instrumentation used to follow the course of accidents. Based or. TVA's experience, the time requir ed to perform an initial dose projection and notify of t-site authorities can be acccmolished in 15 minutes. For most serious design basis accident, the projected two-hour doses at the exclusion area boundary do not reach the protective action guide level for evacuation, i.e., 5 rem whole body dose and 10 rem child's thyroid dose. 13.3.12.2.2 Warning The notification of persons within the potential evacuation sector (see Figures 13.3-5 and 13.3-6) can be accomplished in less than one hour. This time frame has been established by conversations with DOE-ORO, State Health Department, and local livil Defense officials who will be involved in evacuation efforts and who will establish the 50 detailed means used to warn all affected resident and transient persons. Amend. 50 13.3-8 Ji , 1979 m~ > : ,.
' Specific discussions regarding emergency planning for the facilities located within the LPZ boundary have been held with DOE-0RO. the R7ane and Loudon County, Tennessee Civil Defense Directors, and the Nuclear and Inductrial Safety manager of U. S. Nuclear, Inc. Also, with the reduction of the LPZ beundary from the original 5.0 miles (letter, S:L:997, P. S. Van Nort te R. S. Boyd, " Amendment No. 18 to the PSAR for CRBRP", dated April 30,1976) only U. S. Nuclear, Inc. and approxi-nately / of the recreational areas identified in Table 2.1-14 will be located within the LPZ. Consistent with the guidance in the Standard Format and Content of Safety Analysis Reports of Nuclear Power Plants, however, the emergency plans extend outward encompassing a five-mile area.
13.3.12.2.3 Evarna u on Control within the exclusion area, excluding the waterway, will be by the applicant arid can be evacuated in approximately 45 minutes. Due to the short distance involved, the applicant (by means of loud speakers used along the patrol road) will be able to warn persons using the waterway of adverse conditions. Control of the waterway will be coordinated with the regional office of the Tennessee Wildlife Resources Agency, through the Tennessee Office of Civil Defense and Emergency Planning. Upon request, the Agency will block the flow of river traffic, evacuate water craf t, and evacuate persons along the shoreline. The Agency estimates that within one hour of notification, they can have craf t on the river for traffic control at the plant exclusion area. The estimated time to accomplish nou fication and evacuation of any sector of the environs (see Figures 13.3-5 and 13.3-6) is approximately two hours. The basis for the environs time frame is derived from discussion with DOE and local civil defense directors who will be responsible for evacuation efforts. The times are estimates ano .etailed evacuation procedures will be developed for the CRBRP-Radiological Emergency Plan (REP). Buses may be used to evacuate the Edgewood School. The problem has been discussed with the Roane County Defense Director and will be specifically addressed in the CRBRP-REP. Ta t' e 13.3-3 shows the combined resident and maxirium transient nopu-lat un within the five-mile zone for 1980 (projected approximate level at the time the plant is scheduled te commence operation). Figures 13.3-5 and 13.3-6 outline the proposed evacuation sectors and Table 13.3-4 gives the resident and transient population data for each evacuation sector for 2010 (projected peak level during ex-pected life of the plant). Table 2.1-14 estimates the average peak hour use of recreational areas within 10 miles including Melton Hill Dam 50 (Site Number 12) which are included in the analysis.
%end. 50 13.3-9 June 1979 bb Jd,
Figures 13.3-5 anc i3.3-6 show the roads available for evacua-tion of the plant environs out to 10 miles from the site. All road types surface characteristics, and road widths are identified by the legend. Road encumbrances not evident are identified in the CRBRP Map Informa-tion block. Discussicos have been held with the U. S. Nuclear, Inc., the only location within the LPZ Boundary where people will nomally populate on a pre-arranged schedule, c.'ncerning the CRBRP emergency plans. Safe-guard measures for the continued protection of SNM at the U. S. Nuclear, Inc. , facility and an adequate emergency plan insuring protection of U. S. Nuclear employees is possible either through measures taken to protect persons required to remain at the plant and the evacuation of all non-essential personnel. The CRBRP - Radiological Emergency Plan will contein the specific emergency plans regarding the U. S. Nuclear, Inc., facility. w' Regarding the ORGDP id ORNL, both of these facilities are outside the LPZ, but since the ire withir a five-mile radius of the CRBRP, they will be addressed in the emergency plans. Discussions have been held with DOE concerning these two facilities. Should the need be identified, all non-esse..tial personnel can be evacuated in approximately two hours. However, process requirements and security requirements may prevent the complete evaccation of these DOE facilities. The dose levels to these few remaining oersonnel would be acceptable and consis-tent with appropria guidelines. Special procedures, protection or 50 equipment could be .j, as necessary to provide reduction of the dose. Amend. 50 13.3-10 2bb J d
TABLE i3.3-1 PARTICIPANTS IN CRBRP RADIOLOGICAL EMERGEf1CY PLAN Tennessee Department of Public Health Tennessee Department of Civil Defense Tennessee Department of Agriculture Tennessee Department of Public Welfare Tennessee Department of Safety Tenr.essee Department of Conservation Tennessee National Guard Tennessee Game and Fish Commission Tennessee Department of Transportation f l Tennessee Wildli fe Resources Agency City and County Officials of Roane and Anderson Counties Sheriffs' Departments of Roane and Anderson Counties Civil Defense Coordinators of Roane and Anderson Counties Local Police Departments Local Ambulance Service Local Fire Department Oak Ridge Associated Universities Radiation Emergency Assistance Center / Training Site Department of Energy Savannah River Plant Operations Office (IRAP) 50 Department of Energy Oak Ridge Operatio7s Office (IRAP) Environmental Protection Agency, Region IV, Atlanta Eastern Envircnmental Radiation Facility, Montgomery, Alabama Amend. 50 June 1979 13.3-11 JJ1
TABLE 13.3-2
SUMMARY
OF DATA UTILIZED FOR SOURCE TERM RADIOLOGICAL ANAL' ISIS Source Term 100> Noble Gases 25% Halogens (50% release to containments,1/2 if which [25% total] is airborne and available for release) 17 Solid Fission Products 1% Plutonium Released instantly to and uniformly distributed in Reactor Containment Building. Meteorology _ Atmospheric dispersion parameters (x/Q's) are the ninety-fif th percentile values (see Section 2.3). Consistent with Regulatory Standard Review Plan, Section 2.3.4, the 0-2 hour exposure intervals were evaluated based on the single-hour 95% x/Q value. Plume front transit times to downwind positions are based on a wind speed of 1 mile / hour. Containment Modeling The following parameters are used to evaluate Source Term releases from containment: RCB Leakage te Annulus 0.1% Volume / Day (Direct to Annulus Filter Intake) Annulus Flow Rates Filtered Exhaust 3000 CFM Fil terer Recirculation 3500 CFM per 1000 CFM Exhausted Time Delay from Source Term Release No Delay to Initiation of Annulus Filtratior. Time Delay from Source Term Release <10 Secondr to Initiation of Annulus Recirculation Total Bypass Leakage 0.001% Volume / Day (17 of RCB Leakage) Bypass Leakage Direct to Environment 0.0006% Volume / Day (60% of Total Bypass) Bypass Leakage to the RSB 0.0C04% Volume / Day (401 of Total Bypass) 50 Amerd. 50 June 1979 13.3-12 9 , ,,
- i. a V : ,.
TABLE 13.3-2 (Continued) Sources of Bypass Leakage 96.4? Personnel and Equipment to the RSB Airlock 3.6% All Other Sources Gamma Shielding 1.5" Steel (RCB) Plus 4' Concrete Filter Efficiencies Iodine 9 5 Particulate 99 ? Noble Gas Radiological Parameters Inhalation dose factors are per Regulatory Guide 1.109 for a standard adult. Time dependent breathing rates are per Regulatory Guide 1.4. External ganma whole body exposure is based on a semi-infinite cloud per Regulatory Guide 1.4 for the released material and includes direct exposur' com the material within the 2eactor Containment Building. Radioactive decay of nuclides during downwind transit of the plume is conservatively neglected. While conservative, this assumption has minimal inpact on the resu! Ls, since off-jite exposures are 50 controlled by relatively long-lived nuclides. 3r ,-
/Od Jja Amend. 50 June 1979 13.3-13
TABLE 13. 3-3 PROJECTED MAXIMUM RESIDENT + TRANSIENT POPULATION DISTRIBUTION WITHIN 5 MILES OF THE DEMONSTRATION PLANT FOR CENSUS YEAR 1980 Radial Interval (miles) Sector Desi gna ti on 0-1 1-2 2-3 3-4 4-5 N O 95 0 0 0 NNE O O O O O NE O O O O O ENE 5 5 0 0 6520 E 10 5 30 125 1122 ESE 5 5 6015 65 115 SE O 15 125 95 115 SSE O 15 20 120 125
$ 60 305 20 75 95 SSW 0 35 5 70 65 SW 5 25 au 100 75 WSW 0 30 70 315 345 W 0 55 165 105 150 WNW 0 155 100 30 55 NW 0 20 20 0 45 NNW 0 20 6000 0 75 Sum for Radial Interval 85 785 12600 1100 8902 Accumulative Total up 50 to Radius Indicated 85 870 13470 14570 23472 Amend. 50 June 1979 13.3-14 bb 3b/
T ABLE 10. 3-4 PROJECTED MAXIMUM RESIDENT AND TRANSIENT POPULATION
- IN EVACUATION SECTORS WITHIN 5 MILES OF CRBRP Sector 1980 2010 A 6545 6520 B 7497 9162 C 885 885 D 960 1055 E 1365 1955 F 5220 6295
- Transient population is based on current available information 50 +See F igures 13. 3-5 and 13.3-6 2b'd b b; ')
Amend. 50 13.3-15 June 1979
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instructions will contain inf ormation describing the incident or conditions, penbable indications, automatic actions which occur, immediate operator action, and any subsequent operator action necessary to correct or control the situation. The primary responsibility for initiating the co:rettive action will rest upon t he operator who first becomes aware of the situation. He will notify his supervisor of the existing condition and the action he has taken. Al' operat i ng personnel through training and experience will have learned to recognize and evaluate impending failures or malfunctions and to 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. Equipyen t_ Clea rance_ Ins t ruc tion _s The clearance instruction is the method used 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 appli-cable clearance instruction. The Shift En ineer will be responsible for the tagging of equipment within his area o 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 nanes appear on official clearance lists. A clearance will t e established by the use of colored protc tive cards placed to indicate the boundary of isolation or special carating limitations. 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 violate the areas and conditions outlined. 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 condi tion. Maintenance work will be initiated through work requests and/or by the preventati ve maintenance program. Safe working conditions will be assured by the use of TVA's hold order, clearance, and special work permit bh . ,i. . 13.5-3
instructions. Complex and critical maintenance operations which require step-by-step performance will be detailed in written instructions. These instructions covering nechanical, electrical, and instrumentation maintenance will provido information to assure proper coordination of operating and maintenan:e employees as well as step-by-step proceduros to be followed by the craf tsnien doing the work 13.5.6 Surveillance Irstructions Instructions will be prepared covering the conduct of all surveillance tests and inspections designated in the plant technical specifi-cations. These instructions will specify prerequisites, precautions, references, acceptance criteria, necessary step-by-step actions for conduct of the tests and return to normal, data sheets, and sionatures of those conducting and reviewing the tests or inspections. Detailed test schedules and records will be maintained to assure that all surveillance requirements are conducted in a timely nanner and the resul ts a re properly documented. 13.5.7 Technical Instructions Instructions covering routine technical operations will be prepared as t equi red. Examples of these opera + ions are chemical sampling and analysis, chemistry cor. trol, and calibration of vital instrumentation. Fuel accountability instructions delineating the requirements, responsibilities, and meti,ods of nuclear raterial control from the time new fuel is received until it is Aipped from tne plant as spent fuel will be utilized. They will provide detailed steps for physical safeguards, invencory, accounting, and for preparing records and reports. 13.5.8 Section Instruction Letters Each section supervisor will, as the need arises, prepare numbered instruction letters pertaining to administrative routines, responsibilities, and trethods to be followed by members of his section. 13.5.9 Site Em.rnency Plans These plans are discussed in Section 13.3. 13.5.10 Radiation Control Instructions Radiation control instructions are written and made available to all plant personnel. These instructions include pemissible personnel exposures corsistent with 10 CFR Part 29 and other requirements and guidelines to minimize radiation exposures. All plant personnel will b" roq.jired to follow these procedures. Amend. 50 g oune 1979 13.5-4 b0 b : <.
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HEALTH NYSICS MANUAL ADMINISTRATIVE RELEASE MANUAL DIVISION PROCEDURES MANUAL OPERATIONAL QA MANUAL TV A i;AZA RD CONTROL MANU AL NUCLEAR MATERIALS MANAGEMENT GUIDE RADIOLOGICAL EMERGENCY PLAN STANDARD PRACTICES C m o, GENERAL EMERGENCY S ECT IO N N OPERATING N OR M A L MAINTENANCE c.r' INSTRUCTIONS OPER\ TING UPf R A TI NG INSTRUCTIONS I N STRUC T I ON i' INSTRUCTIONS I NS TR U CT I O NS LETTERS CD I ABNORMAL SITE R AD I OLOGI CA L TECHNICAL OPERATING SURVEILLANCE EMERGENCY CONTROL I NS TRUCT I MS INSTRUCTIONS INSTRUCTIONS PLA NS INSTRUCTIONS o> g8 Figure 13.5-1 Plant Procedures
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13.7 INDUSTRIAL SECURITY The requirements of 10 CFR 73.55 and NRC Regulatory Guide 1.17 will be met for the CRBRP. This section discusses in general how the CRBRP will meet these requirements. The CRBRP Physical Security Plan
, shall provide specifics including the contingency plan and qualification and training plan for security personnel and wili be submitted as a separate proprietary document at the FSAR stage. It will provide specific details as required by 10 CFR 50.34(c).
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 organizacion chart shown in Figure 13.7-1 delineates this responsibility which is explain'd in the following Se, Lion 13.7.1.1 and 13.7.1.2. 13.7.1.1 Division of Proper _ty and Services The Division of Property and Ser/ ices (P&SVS), with the assistance of other TVA organizations, develops guides and standards on property protection, reviews protection plans for compliance with these guides and standards, and advises in their development and appli-cation. P&SVS provides police and fire protection service on properties foc which it is responsible and furnishes these services to other organi-zations in accordance with their protection plans. Tne 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 ~VA Emergency Coordinator in carrying out P&SVS's security respon-sibilities with other TVA organizations and providing liaison with federal, state, and local agencies on security and emergency preparedness matters. His organization prevides supervision f r the PSS. The supervisor of PSS, located in the Public Safety Service Branch office, has several PSS area chiefs, each of which are over several PSS supervisors of units at power generating plants in their area. The PS Security unit at the CRBRP is under direct supervision of the Public Safety Service Branch, but functions as an onsite armed security force for the Plant Manager according to existing plans and 50 the Manager's requirements. Amend. 50 13.7-1 June 1979 263-6C}l-
The Public Safety Service Branch is responsible for recruiting, training, and assigning security force personnel to the CRBRP unit as required. 13.7.1.2 Office of Power Tne Office of Power is responsible for protection of power properties. It develops detailed plans and applies specific measures, with the advice of P&SVS. The Chief, Management Services S'caff of the Power Manager's Office, represents the Power Manager on all security matters. The Power Security Section which reports to the Chief of Management Services Staff coordinates planning a".d admin stration of industrial security i 50 measures with all correrned. Each Power division is responsible for security and fire protection and prevention af Power facilities under its control in dCCordance with general enlicies and general instructions from the Power Manager's Office. The Division of Power Production is responsible for securi ty of nuclear ]lants. At the CRBRP, the Plant Manager and in his absence the Senior lant Supervisor on duty is responsible for security of the plant 'n accordance with general policies, plans, and instructions received through administrative channels. 13.7.1.3 Employee Selection As discus ,ed in Section 13.1, TVA will operate the plant, and accordingly, will provide all operating and security personnel who will be regular TVA employees. TVA appoints, promotes, transfers, and retains employees on the basis of merit and efficiency, 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, reliability, and stability to vacant positions in preference to hiring persons from outside , e organization. This .is of ten accomplished by upgrading employees through internal training programs. Specific instructions pertaining to personnel matters are contained in Section III of the TVA Administrative Release Manual. These instructions are observed by all plant supervisors, especially as they apply to appoiritment, transfer, promotion, and retention of employees. Selection for a position is supportable by records of education, training, and experience, and by records of judgements which have been made regarding work performance, ability, and condition 1 of health. Amend. 50 13.7-2 7a nn- June 1979 tU/ UVs
In selecting for placement or retention in positions, covered by agreements negotiated Setween TVA and the employee organizations, the provisions of such agreaments are observed. Because of TVA's conformance to the Veteran's Preference Act, when employing 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. Each new annual TVA employee is given a physical examination and a national 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
, there must be satisfactory results from his security check and 50 emotional stability check.
PS officer selection procedures include a preemployment interview by the PSS area chief and one or more PSS unit supervis rs in addition to the steps previously mentioned. Upon acceptance, the candidate's first six months of - 'acent are 3robationary. Appointment as a PS officer is dependent upt i satisfactory service during this period and satisfactory completion of training and qualification as provided for in the CRBRP Training and Qualification Plan to be 50 submitted with the CRBRP Physical Security Plan. 13.7.1.4 Employee Evaluation Because of the general policy of promoting present employees rather than appointing candidates from outside TVA, most employees at the CRBRP will be known from their previcus encioyment 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 regular part of day-to-day continunus 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. In addition to this kind of review, the perfonnance of employees in management and salar r olicy positions are reviewed formally and the results reported in order (1) to further aid in maintaining a high level of employee performarice and the maximum util-ization of employee abilities; (2) to provide recorded evidence of 1 13.7-3 Aand. 50 June 1979 2 6 :' 0 0 '<
employee performance for use in making judgements concerning transfer, demotion, promotion, and terminations; (3) to assure that employees are adequately and systematically infomed of the effectiveness of their service; and (4) to further facilitate the maintenance of a high standard of supervision in TVA. All employees' services are reviewed formally at the time of status changes and at such other times as may be required to achieve the above purposes. A service review shall precede each recommendation for operator liccnsing or renewal of an operator license. 13.7.1.5 irdus trial Securi tLTraining All erployees 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 con-50 ducted by a menber of the Plant Security Force under the direction of the Plant Manager 13.7.2 Plant Desion The physical plant design has been developed so as to accom 5(j date the necessary security provisions. TVA, along with DOE and PMC and its architect-engineer, Burns and Roe, will provide a continuing review of the olant design, as well as the detailed security provisions. Burns and Roe, as the architect-engineer for tne Project, has been delegated the responsibility for detailing the security provisions. 31 The design criteria used at the CRBRP will assure that the physical security facilities and the plant layout are developed so as to thwart 5danyattemptedsabotage. The physical sectri ty design will (1) Control entry to the plant site and portions of the plant; (2) Deter or discourage penetration by unauthorized persons; (3) Detect such penetrations in the event they occur', a id (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 reasures to prevent unauthorized persons from enterin') the protected area. Should such persons succeed in entering the protected area, special access control measures will prevent them from entering vital equipment areas and the SW1 material access area. ) l 31 Amend. 50 June 1979
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13.7.2.1 Design Features
'he dtsign features and other physical security measures that will protect against or limit the effects of possible sabetage efforts include:
50 a. A security barrier with dual intrusion detection system around the perimeter of the plant, with gates that are kept closed and locked except during times 47 of authorized use.
- b. Emplo) ?e 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 void of obtrusive structures and plant growth. In addition, a cleared zone will be raintained outside the isolation 'one to facilitate observation of persons approaching the isolation zone.
- d. A perimeter patrol road extending completely around tne
,lant inside the security barrier.
- e. A renote-controlled, outdoor, closed circuit television (CCTV) system to permit observation of the plant peri-meter, isolation zone, cleared zone, protected area, and approach roads.
- f. An outdoor lighting system to provide illumination to the protected area and isolation zone at a level com-patible for both visual and CCTV observation.
- g. A minimum number f exterior plant doors leading to vital areas, all of which shall be hardened against 31 penetration and kept locked or otForwise secured when not in use. 3)
- h. A cardkey electronic access control system to control personnel access to vital areas in conformance with each employee's level of authorization.
- i. An alarm system to indicate status of hatches, emergency exits and seldom used equipment or personnel access doors providing access to vital areas not cardreader equipped.
- j. An access control building (gatehouse) to control personnel access to the protected area and containing equipment to search personnel for weapons, explosives and special nu lear material. 1 Amend. 50 13.7-5 2b9 bbb d"""
- k. A communication system which will allow cor'inuous communications between PSS officers and the central alarm s ta ti on. Also, redundant communications links will be maintained between the plant and the local law enforce-ment agency.
- 1. An electric power system to provide emergency power to the security and light'ng loads during periods of
" blackout" or loss of nori.31 power.
- m. A force of trained, uniformed, and armed PS officers used on a three-shif t basis to police the property, prmide access cnntrol, respond to alarms, evaluate 50 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 effects of postulated major equipment failures, natural disasters, and operator errors which would alsc serve to minimi7e the effects of industrial sabotage.
13.7.2.2 Physical Arrangements The CRBRP site is in a remote location. It is unlikely 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 13.6 near the center of a 1364-acre tract owned by and in the custody of the United States Government (see Figure 13.7.2). 13.7.2.3 Owner-Controlled Area U1 timately, a permanent access road to the plant will lead into the plant. During construction, a temporary construction road will lead into the 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 shall 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 shall be made at least once each evening and night shift. While construction is in progress, the tem-porary constru tion road will be the only route of access to the Project. ) 13.7-6 Amend. 50 June 1979 269 007
A continuous access control guard post will be maintained on this road for the duration of construction activities. Figure 13.7-2 shows the reservation boundary of the owner-controlled area. 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 mluding the mair. plant buildings and associated outdoor facilities. 1 isolation zone shall be maintained at least 20 feet outside and E0 feet inside the security barrier. A perimeter patrol road will be located inside this barrier. A sectionalized intrusion detection system designed to be self-checking and tamper-indicating will be located along the barrier with sensors located on or between it and the patrol road. A closed-circuit television (CCTV) system using low-47 light level cameras with zoom lens and remote pan and tilt controls as required will be used to provide a means of promptly viewing the sector or general area involved. Proprietary Figure 13.7-3 indicates compliance with ANSI N18.18-1973, Section 3-3. 13.7.2.5 Vital Equipment and Vital Areas All vital equipment and material access areas shall be located within a vital area or building which, in turn, shall be located within 47l a protected area. Doors and gates to vital areas and to other selectcd sensitive areas shall be kept closed and locked at all times when the areas are not occupied. Proprietary Fi-gure 13.7-4 indicates compliance with ANSI N18.17-1973, Section 3.4, and other applicable guides and regulations. The material access area is located within the Reactor Ser ' ice t!uilding. Na activities other than those which require access to secial nuclear mdterial (SNM) or equipment employed in the process, use, or storage of SNM will be conducted in the material access area. 31 As construction nears completion and the equipment made operational, the doors and gates to vital areas shall oe ident fied by sigrs which state that entry through them shall be witl the perris > ion of the shif t 7ngineer on a need basis. Upon completion of construct. ion, 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 mergency 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 1 13.7-7 Amend. 50
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June 1979
system will be prograt med individually and can be programmed out at any time if lost. The cards will be issued and returned daily to insure that they do not leave the site. Also, since the card will be required 50 in the performance of the employee's duties, this will serve as a con-tinuing availability check of issued cards. 13.7.2.6 Alarm Station All intrusion alarms will terminate on a graphic security intrusion display panel and annunciate in a continuously manned central alarm station located within the protected area and in at least one other continuously manned station (see Proprietary Figure 13.7-4). 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 50 facilities and the CCTV system will be tested at the beginning of each PSS work shift. This onsite alarm station shall be considered a vital area. All intrusion alarms, emergency exit alarms, and other alarms will be required when purchased to meet the level of performance and reliability specified by Interim Federz ' Speci fications W-A-00450B, GSA-FSS, dated February 6,1973. 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 st ands of barbed wire on angle brackets) or other equivalent arrangement such as masonry walls. Other fencing located within the protected area nay be only 6 feet high without barbed wire. The alignment of the new security 50 barrier will have a minimum number of angles and curves to facili-tate ef fective observation and maximum length sectors of the intrusion detection system. 13.7.3 Security Plan T e Plant Physical Security Plan shall describe security measures used to minimize the potential for industrial 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 encloses all vital areas. The plant shall have two portals for normal access: 1 l3.7-8 MW'd.50 QuDe,1979
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(1) a personnel portal and (2) a nearby vehicle gate. General public visitors shall not permitted inside the security barrier. Employees and special visitor's parking areas shall be located outside the security barrier, Vehicle access shall be limited 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 47 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 doors shall be kept locked or secured when not in use. Also, a number f interior doors shall be cardkey controlled to prevent unauthorized 47l' access to certain more important areas. All persons authorized to enter the protected area unescorted shall have had a satisfactory securit3 check and emotional stability check and sha'.1 have completed as a m .imum a brief plant indoctrina-tion and radiation protection course which describes plant organization and layout, controlled zones, radiation and contamination hazards, exposure limits and controls, elementary health physics, and pertinent sections of the site emergency plan. Even those persons who are authorized unescorted access shall have their movement limited by physical barriers, such as locked doors, to precent them from entering areas containing vital equipment or areas of high radiation levels. Only those who need access to these areas shall be prnvided means of entering. When special visitors and other persons who have not completed SC this training enter the ".otected area, they shall be escorted by an employee trained in radiation protection ano plant emergency pro-cedures. The escort shall be responsible for the safety and action of the people in his charge until he checks them out of the access portal. 5L 13.7.3.2 Control of Personnel by Categories 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 recognization, TVA identification card, or other available identification media. They will be issued a white numbered visitor's badge to be worn on their outei garment while 1 within the plant prntected area. @ Amend. 50 13.7-9
/69 0]O June 1979
During construction, temporary fencing and other physical barriers will separate the operating unit and associated equipment from construction activity to prevent uncontrolled access by construction workers into operating unit areas. The only unescorted construction workers who may ba 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 50 , and relevant sections of the site emergency plan. Contractor personnel, manufacturers' representatives, and other special visitors who require access to the olant shall be logged in and badged by the PS officer on duty. The officer shall then call the appropriate plant supervisor and arrange for an escort. Acms Conuol During Bergencies 50 Upon hearing of an emergency, the PS officer on duty at the access portal shall lock all doors to ensure controlled entry and exit. Special visitors who are onsite shall be escorted to the access portal. Plant employees shall report the predesignated stations from which th 'y will be dispatched as needed to combat the emergency. All access control procedures will be compatible with the CRBRP Radiological Emergency 50 and Contingency Plans. 13.7.3.4 Surveillance of Vital Equipment and Material Access Areas Unit operators shall continuously monitor the status of plant systems and equipment by means of annunciators, indicating lights, 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 perfomance. Most equipment will be in continuour operation and any change will immediately be detected by the operator. Standby and emergency equipment shall ')e periodically tested on a routine basis as required by the technical specifications. Assistant unit operators shall inspect equipment and spaces at least once each shif t. In addition, the assistant shif t engineers and other supervisory personnel knowledgeable in plant ] conditions shall make frequent unsecheduled inspection tours through the plant. Procedures shall be employed to control access to the vicinity of the material access area. In addition, activities in the vicinity of the material access area will be monitored. The combina- 31 tion of these efforts should provide reasonable assurance that unauthorized physical changes in the status of components of equipment l will not be undetected for long periods. 13.7-10 Amnd. 50 June 1979 9 mw
Key operating log sheets and selected recorder tracings shall be reviewed daily except for veekends by the Plant Engineering Section. Abnormal changes observed shall be cailed to the attention of the Plant Manager and the appropriate supervisors for investigation and corrective action, if required This operational audit shall serve
;o assure earl; ction of physical changes which would have a significant bearing on plant performance.
13.7.3.5 Potential Security Threats Should an unauthorized person succeed in entering the protected area, the access control measures in use would not allow him access to vital equipment. Operating personnel trained to be alert for unauthorized persons would recognize him as an intruder and arrange for his appre-hension by a PS officer. Plans shall be prepared to cover actions in the event of civil disturbance, emergencies, and bomb threats. Detailed emergency pro-cedures shall be provided plant employees so that they may cope witt these and other events in the optimum manner possible. If there appears to be a real threat of civil disorder or another type of serious security threat to the plant or radiological emergency, all off-duty PS officers shall be recalled, and additianal PS officers in the area shall be called in. Local and State law 50 enforcement authorities shall be contacted for assistance. Arrangements with federal, State, and local law enforcement agencies will be addressed in the CRBRP Radiological Emergency Plan. In the plant, precautions shall be taken to protect vital areas from threats of fire or other damage. When appropriate, the Plant Manager will make a written report to NRC. Bomb threats normally come by telephone. Employees who might receive such a call shall be trained to extract as much information as possible from the caller. Based on this and other information, action would be taken to search for the bomb, evacuate areas, shut-down the reactor, or tako any other actions deemed necessary to ni otect the plant and personnel. More detailed descriptions of decisions / actions regarding potential security threats shall be included in the CRBRP Contingency 50 Plan. 13.7.3.6 faministrative Procedures In the event of an incident of suspected sabotage or condition which threatens the security of the plant, the Public Safety Service shall immediately notify the Plant Manager and initiate a thorough 13.P M 0: 2 Amend. 50 June 1979
investigation. A report shall be prepared which includes as a minimum the cause of the even , extent of damage, if at y, and action taken to prevent recurrence of similar event. Copies of the report shall be sent t( the Plant Manager; Chief, Nuclear Generation Branch; Power Security Section; Chief, Publ c Safety Service Branch, P&SVS; and the Division of Law. When appropriate, the Plant Manager shall also report the situation tr NRC. Representatives of the Power Security section and Public Safety Service Branch, P&SVS, shall make an annual audit of the CRBRP Physical Security Plan for adequar/ of content a.id performance. Based on their audit, they will make recommendations for revising and updating the plan and related plant procedures. 13.7.3.7 Test and Inspections This information will be supplied in the CRBRP Physical , 1 Security Plan. O Amend. 50 June 1979
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TABLE 15.5.2.4-1 0FF-SITE DOSES FROM COVER GAS RELEASE DURING REFUELING Dose (REM)* Site Boundary LPZ [2__h_r. -0. 42 mi . ) (30 days-2.5 mi.) D,(Skin) e 4.0 x 10 -3 1.1 x 10 -3 49 D., (Whole Body) 4.4 x 10 -3 1.2 x 10 -3 0 49 l
- Integrated exposure based on puff release.
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O 50 l This page intentionally blank. 15.5-21 Amend. 50 June 1979 269 0: 3
15.7.2.3 Generator Breaker Failure to Open at Turbine Trip 15.7.2.3.1 Identification of Causes and Accident Description In the event of a turbine trip, the generator load break switch is altamatically opened by a signal from the turbine trip logic. The turbine trip logic simultaneously causes the generator field breaker to open regardless of whether or not the generator load break switch opens. A generator load break switch failure can occur from electrical or machanical failure of the tripping mechanism. 15.7.2.3.2 Analys_is of Ef fects and Consequences If the generator load break switch fails to open af ter a turbine trip, d PldN L EUwl 3Uppl, luCkout is initiatcd. The lockout initiates the disconnec-tion of tne Plant Power Supply by tripp'ng the appropriate 161 KV circuit breaker in the Generating Yard. This causes loss of the Preterred AC Power Supply as described in Section 8.2.1.1. Upoa loss of power from the Preferred AC Power Supply, the Normal AC Distiibution System and the Safety-Related AC Distribution System, are automatically transferred to one of the Reserve Transformers as described in Section 8.3.1.1.4. The reactor can be shut down with no adverse consequence, as described in Section 15.3.1.5, which evaluated the effects of a turbine trip. 15.7.2.3.3 Conclusions The consequences of a turbine trip with subsequent failure of the generator load break switch to open is negligible, since one offsite power supply is still available to the AC Power Distribution System. uO/') ' ap) ( , -, , 9 15.7-11
15.7.2.4 Rupture in RAPS Cryostill 15.7.2.4.1 Identification of Causes and Accident Description The RAPS cold box contains the cryogenic still in which Krypton and xenon are extracted from the reactor cover gas stream. Durino normal operation, this stream is collected in the surge vessel in th : RCB, flows at a controlled rate of 10.0 scfm into the cold box, which is in the '<SB, and then through the cryostill. The krypton and xenon isotopes. botl stable and radioactive, are condensed in the liquid argon still bottoms. The argon is condensed to liquid, in turn, by coiled tubing through which liquid nitrogen is passed. This coil penetrates the cryostill wall at two locations, sealed by welding. A complex rupture at either of these locations could vent both the still interior and the liquid-nitrogen line to the cold box, and hence to the cell atmosphere. Although such a complex rupture is not expected to occur, it would result in both a significant activity release and a significant increase in cell-atmosphere pressure. This postulated act Jent determines the RAPS cold box cell leak-tightness requirements. It is presented in order to report the maximum crediblo resultant doses to unrestricted areas, and the indicated cell leaka se specification. 15.7.2.4.2 Analysis of Effects and ConseqJences For the purpose of the accident analysis, it is conservatively assumed that the reactor has been operating sufficiently long, with gaseous fission products from li' failed fuel, for steady-state isctopic composi-tion to exist in the cover gas system. It is assumed, also conservatively, that the cryostill has not been off-load a to the noble gas storage vessel for one year (maximum period) and the.efore contains a maximum inventory of radioactivity. The accident is the rupture of a liquid nitrogen line at the cryostill wall in such a manner as to breach the wall also. This rup-ture would release liquid nitroger, liquid argen, and reactor cover gas flowing from the surge vessel into the cold box. The cold box vents to the cell under a slight pressure differential. The volume of nitrogen released to the cell corresponds to 10 min-utes of maximum nitrogen flow; after this time,the nitrogen is automatically valved off by a cell pressure signal. 7300 scf of nitrogen are thus esti-mated to be released into the cell at the initiation of the incident. Also released at this time is the liquid still bottoms,1.5 cu f t. , which corresonds to 1275 scf of argon. Redundant radiation monitors, located in the RAPS cold box cell, will 49 sense the presence of radioactivity, sound an alarm, and initiate a signal which will close cell isolation valves. The signal will not close the valve which allows the cell to vent to LAPS since it is a normally open valve. Therefore, the cell will normally continue venting to CAPS af ter the accident. However, for this analysis, the valve is assuned to be closed in this accident scenario requiring the cell to have a tighter 16 50 leakage specification. 15.7-12 Amend. 50 June 1979 h q r) nl' U1 z' C / d
The most critica: shutoff valve in this system is located on the inlet side of the cold box. Fce purposes of this analysis it is assumed that in addition to the c Tve incident, this valve fails to close. As a result of the alarm, the operator has an assumed 30 min. to take alte native corrective action. One such action is to close this line by resetting the flow control valves, located between the surge vessel and the cold box, to zero flow. During the maximum operator response period of 30 min. , radioactive argon will continue to flow at the normal rate of 1G.0 Ec fm from the surae vessel to the cold box and into the cell thiouyn tna break. This will result in an additional 300 scf of gas being released into che cell, The assumed initial candition, tnen is that the gases from all three sources (liquid nitrogen, liquid argon and gaseous argon from the surge vessel) come instantly to standard temperature, but elevated pressure. No allowance is taken for radioc:tive decay during the pressure-rise period. The total amount of gas released into the cell (whose net volume s 8603 i acf) is the total of the above or 17480 scf The resultant calculated initial pressure is 15.1 psig. T1e inti"al adioactivity inventory is shown on Table 15.7.2.4-1. The design basis for the leakage sp. cification is that the in-tegrated 2-h site boundary dose (e a) be limited to a value below 10 : of the 10 CFR 100 value following the rupture incident in the cold box. 15.7.2.4.3 Conclusions The postulated RAPS cryostill rupture ir cident requires that the cold box be located in a controlled- eakage enclosJre. with a permissible leakage rate such that the site bouricary dose is below the CRBRP guide-line value of 2.5 rem. The technical specification and testing provisions are discussed in PSAR Section 16.4.8. The analysis of the scenario des-cribed in Section i5.7.2.4.2 shows thet a cell leakale specification limit of 29 of the cell volume per day at 13.1 psid will prevent the site boundary dose from exceeding 2.5 rem, in the very unlikely event of this accident. With this leakage specification, the calcuiated 2-h radioactive gas release to the environment is shown in Table 15.7.2.4-2. This cell will require testing, prior to plant startup and demonstration that the 40' leakage is within the specificatior. Aoiitional testir g of the cell will be required, if the cell is accessed. 36 269 010 Amend so June 1979 15.7-13
TABLE 16.2-1 PLAf4T PROTECTION SY:TEf1 PROTECTIVE FUtiCTI0 tis _ Primary Shutdown System Secondary Shutdown System o Flux-Delayed Flux e Modified Nuclear Rate e Flux-v/ Pressure e Flux-Total Flow e High Flux e Startup tiuclear e Primary to Intermediate o Primary to Intermediate Flow Speed Ratio Ratio e Primary Pump Electrics e Steam Drum Level e Reactor Vessel Level e Evaporator Outlet Sodium Temperature e Steam-Feedwater Flow Mismatch 50 e IHX Primary Outlet Temperature e Sodium Water Reaction 0i9 g 269 Amend. 50 June 1979 16.2-2
Amendment 50 List of Responses to NRC Questions
Reference:
NRC Letter Dated August 17, 1976 NRC Ques. No. 111.23
'/.69 020 Q-i Amend. 50 June 197
Question 001.301 (Chs. 3, 5, 9,11) Resolve the numerous inconsistencies in the ASME Code Classes de-signed for identical systems and components in Chapters 3, 5, 9 and 11 of the PSAR.
Response
With the exception of the two items listed below, no inconsistencies have been identified in PSAR designations of ASME code classes for the PHTS, IHTS, SGS, SGAHRS, the Nuclear Island Heating, Ventilation, Cooling and Air Conditioning System, Recirculating Gas Cooling System, Chilled Water Systems, or Nuclear Islar.d Treated Water Systems. Two inconsistencies have been resolved as noted below: a) Table 5.5-7 has been revised to delete the alternative code classification and provide consistency with Table 3.2- 5 b) Table 11.2-5 has been revised to indicate " Manufacturers Standards" for pumps per Regulatory Guide 1.26, Rev. 2, June 1975. As indicated in the response to Question 001.274, the CRBRP Inert Gas Impuring Monitoring, EVST Cooling and Auxiliary Liquid Metal Systems requirements are revised to comply with the NRC position on Safety Classes. 50 269 02; Amend. 50 Q001.301-1 June 1979
1.0 Introduction Sodium Leaks from the Primary Heat Teansport System (PHTS) piping into the Reactor Cavity and the PHTS Cells and from piping within the Overflow and Primary Sodium Storage Tank Cells have been analyzed. These cells constitute the major cells within the Reactor Containment Building (RCB) both with respect to the size of the cells and the volume of sodium contained in the equipment within the cells. All of these cells operate with an inerted atmosphere (nitrogen) with a maximum of two volume percent oxygen. A spect. rum of leaks has been considered for the analysis. Leaks as small as 100 gm/hr can be detected by the sodium leak detection system in 250 hours and are indicative only of the initiation of a potential breach of the piping. Such leaks are so small, however, that no significant sodium is lost from the affected system and the impact on the cell temperature and pressure and the cell liners is neoliaible. The smallest leak rates considered in the analysis presented herein corresponds to the piping Design Basis Leaks (DCL) that have been established for CRBRP. The cefinition of these leaks is discussed in PSAR Section 3.8-B and represents a conservatively established limit for the leak rates that may be expected to occur in the piping systems within the RCB that contain primary coolant sodium. As shown in Table QC01.581-1 for the three cells considered here. the DBL leak rates are 3 gpm or less. In order to assess the sensitivity of cell design parameters ( pressures temperatures, etc.) to leak rate and quantity, other leak rates were examined. A Moderate Energy Fluid Systems (MEFS) Leak has been defined which is equivalent to the leak size established in the NRC Standard Review Plar - Section 3.6.2. The leak size is defined as a circular opening with area equivalent to a rectangle which has dimensions of one-half the pipe diameter and one-half the pipe thickness. In addition, a still larger pipe break has been consideret that results in a leak which approximates the maximum flow through the piping systems in the affected cells. This leak has been termed the Evaluation Basis Leak (EBL). In developing the mpectrum of leaks considered, the operating characteristics of the plant were considered. The Design Basis Leak assumes that the leak detection system is functional and that operator actions will be taken to reduce and finally eliminate the source of the leak. The MEFS Leak scenario assumes that leaks in the primary piping which are large enough to reouce the sodium level in the reactor will activate the Plant Protection System (PPS) with resulting pump trip. However, the tirre of the trip is arbitrarily adjusted to maximize the resulting cell gas temp-erature and pressure. For the PHTS piping EBL leak it is conservatively cssoned that the sodium leak rate is equal to the design sodium flow rate in the piping and that this flow rate is maintair.ed until the maximum evcil-able system inventory is discharged through the break. Nc purrp trip is therefore assumed. This is, of course, extremely conservative since the actual plant flow characteristics would tend to reduce the discharge rate
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?Uia Gau 0001.581-3 Amend. 50 June 1979
and the quantity of sodium released. Actually, for a leak of this magni-tude, the protection system would detect the i'ailure end shut down the pump within a few seconds. For leaks in the Overflow and Primary Sodium Storage Tank Cells, leak detection systems are assumed operable and operator action is assumed. Other conservative assumptions used in the analyses concerning leak rate and bak volume are noted below. It should be noted that the location of all i HTS leaks in the Reactor Cavity is assumed to be the cold leg. The cold leg leak was selected because cold leg pressure is significantly higher than that of the hot leg (%100 psig vs %C psig). Since the pipe leak geometrics are very close (an area within a factor of 2.5) for the hot and cold legs, leaks in the more highly pressurized cold leg result in significantly higher leak rates and thus higher cell atmosphere temperatures and pressures, even though the cold 50 leg sodium temperature is lower. Also, the duration of the EBL leaks are slightly different in the RC than in the PHTS Cell. The reason for this difference is that slightly more sodium can be discharged into the RC than the PHTS Cells because of the piping elevations, but the leak rate is assumed to be the same for both. The EBL leak rate is not as high as mis... be postulated for a double ended rupture of the PHTS piping. However, as shown in the response to NRC Question 001.700, additional flow, if it were available, would not 50 le d to higher cell pressurization. In all cases, the elevation of the pipe break was chosen such that the @ maximum cur.ulative sodium release was considered. The maximem spray fire consequences were included by assumirig that the er. tire discharge was completely converted into spray. This is, of course, unrealistically conservative for the larger leak sizes. The effect of various geometric pipe break configurations has been included in the analyses which have been performed. The design basis leak considers a longitudinal crack in the pipe. The MEFS Leak assurres a circular hole in the pipe which has the same flow characteristics es a sharp-edged orifice. The EBL assumes that the break size has the same area as the original pipe. In all cases, it was assumed that the total discharg was converted into spray and the resulting droplets traveled through the entire height of the cell. The analyses presented herein were perforred to determine the transient temperatures and pressures imposed upon a PHTS cell, the Reactor Cavity and Cell 102A (In-Containment Primary Sodium Storage Tank Cell) for a spectrum of pipe leak sizes assuming the cell liners remain intact. To assess the consequences of a failed liner, a parametric study was also performed for the PHTS cell which assumed that portions of the liner system had failed. (The PHTS cell was chosen as a prototypic cell in which to denonstrate that the inerted lined cells in the containment have margin Q001.581-9 ')" $s 9 0 Amend. 50 June 1979
to accommodate substantial failure of the liners.) A failure mode analysis of the liner system presented in the response to NRC Question 130.89 indicated that in the unlikely event of a liner failure, the extent of the failure would be very limited. Section 4 of this response presents an assessment of the consequences of a liner failure which includes the effects of sodium-concrete interaction. The analyses were performed using three computer codes, SPRAY, SOFIRE and CACEC9. For the conditions in which the liner was assumed to remair, intact, SPRAY and SOFIRE were used to evaluate cell transients. For the evaluation assuming liner failure, all three codes were employed with the sodium-concrete-water reactions included in the CACEC0 analysis. The basic material properties included in the SPRAY and SOFIRE analyses are listed in Table Q001.581-2. The material properties for the CACECO code used in analyscs of failed liners are presented in Table Q001.581-5. Summary results of all the analyse are presented in Section 2. Detailed discussions of the analyses, on a cell-by-cell basis, are presented in Sections 3 and 4. 2.0 Sumrary 2.1 Summary Results - Cell Liner Design Conditions Table Q001.581-4 presents the summary results of the analyses for the design conditions where cell liner integrity is maintained. For each cell and leak evaluated, the following peak transient values are itemized: gas pressure, gas temperature, liner temperature, floor str-tural concrete temperature and the non-wetted wall struc h ral concrate tec ature. The concrete temperatures provided in the tab.e represent the temperatures in the first one-half inch of structural concrete behind the floor gravel aggregate and the wall insulating concrete. Note that the peak transient temperatures for the non-wetted wall structural concrete are not specified in absolute terms but rather as a temperature that this concrete will not exceed. This is necessary because it is not feasible to continue the SOFIRE analyses for the time the wall concrete passes through a peak temperature, as is the case for the floor concrete trar.sient. However, by by examining those transients in the insulating concrete which have passed through their peak temperature and are decreasing, it is possible to estimate a peak temperature value that the wall structural concrete will asymptotically approach and thus to specify a temperature it will not exceed over the entire course of the transient. This method was used to specify the bounding temperatures in Table Fl.581-4 for the wall structural concrete. It should also be noted that the pressures in the reactor cavity quoted for the MEFS and the EBL will be limited by the pressure venting system that is being installed to accommodate the Thermal Margin Beyond the Design SC Base events. While provi sions have been included for a venting system, the rupture disk burst pressure has not been selectad. It is expected to Q001.581-10 Amend. 50 June 1979 2rj p g c9y,
be in the range of 8-10 psig. The primary purpose of this study was to determine the potential pressures that may exist in these cells for a spectrum of postulated leaks. F;r this reason, use of the Reactor Cavity vent system or consideration of venting the PHTS cells or Cell 102A was not included in the evaluation. It should be noted that the maximum pressures reached are less than the design pressures for the cells. O Amend. 50 Q001.581-10a
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Three general observations based on the data summarized in Table Q001.581-4 follow: e The transients associated with any of the DBLs are minor, l'ath in terms of cell gas pressurization and in terms of structural concrete transients. Specifically, none of the DBLs result in a cell gas pressurization in excess of 1.1 psig. The structural concrete temperatures (floor and wall) do not exceed 120 F for any of the DBLs . e for each of the cells considered, the MEFS Leaks result in peak gas pressures and temperatures approximately a factor of 2 less than the corresponding EBLs. e For the PHTS Cell and the Reactor Cavity the structurcl concrete transients are essentially equivalent for either the MEFS or EBL. However, for Cell 102A, the EBL concrete structural transients are slightly (30cF-80 F) more severe than the MEFS concrete transients. 2.2 Summary Results - Evaluation with Cell Liner Failure Cell Liners will prevent the interaction of sodium with the concrete cell structures in the event of all sodium spills. However, in response to NRC Question 130.89, a failure mode assessment was performed to determine a scenario that could lead to liner failure. Such a scenario was developed by assuming a series of very pessimistic conditions. The Failure Mode Evaluation results in the determination that the worst failure that can be postulated is a small crack in the liner floor, assumed to occur at or e near a weld joint, which extends the entire length of the cell. The crack opening would not exceed approximately 0.2E inches in the PHTS Cell ini-tially and would not appear until the pool of sodium began to cool down. For the conditions assumed in the response to NRC Question 130,89, the crack would not open for over an hour after the spill. During this time delay period, much of the water in the upper portion of the structural concrete will be released and vented through the liner venting system, thus reducing the amount of water available for reaction with the sodium pool following crack opening. In order to accurately evaluate the effects of a liner failure, a very complex analysis would be required and probably a series of tests conducted to confirm the results. The sodium which leaks through the assumed liner crack would be contained in th:. aggregate. Since the aggregate is chemically inert to tne sodium, it provides two major benefits. First, the aggregate will occupy approximately one-half the volume between the liner and the structural concrete. This will limit the amount of sodium tnat is available to react with the water released from the concrete. Since the liner failure is very small compared to the cell floor area, there will be little or no opportunity for fresh sodium to replace reacted sodium under the liner. Secondly, the aggregate will tend to hold the reaction pro-ducts on the surface of the concrete, forming a partial barrier to further Q001. 581 -l l Amena. 40
")g g 23,uly 1977
cell atmosphere with sodium discharged after the peak is reached, but neglecting this and assuming only pool-oxygen reaction is conservative with regard to long-term structeral transients and insignificant with regard to the cell atmosphere transients since the SPRAY analysis shows these transients to have already peaked. Figure Q001.581-4 presents the results of the spray-phase transient analysis. As indicated, the peak cell pressure of 11 psig occurs at 3.5 minutes, corresponding to peak cell atmosphere temperature of 5700F. Figures Q001.581-5 through-S present the longer-term cell transients based on the SOFIRE analysis. Evaluation Basis Leak (EBL) The PHTR Cell EBL is a spill of tne total spillable volume in a loop (20,000 gallons). The spill rate is conservatively taken to be the normal lcop flow rate of 33,500 gpm which yields a spill duration of 0.6 minutes. The SPRAY analysis for this leak assumed that the cntire discharge was in the form of 0.18" droplets at 1015 F and in a manner such that the spray occupied one-third of the cell volume. Figure Q001.581-9 pre-sents the results of the spray nhase transient analysis. The peak cell atmosphere pressure and temperature of 23 psig and 10300F, occur near the end of the sodium discharge and begin to de' ase after the sodium dis-charge is complete. The entire PHTS Cell ox,;en content is depleted by the end of the sodium discharge. As discussed for the MEFS leak, the longer-term transients were evaluated with SOFIRE, with initial conditions corresponding to the peak spray-phase transient conditions. In this case, however, the oxygen concentration used for SOFIRE was zero, since all the oxygen is consumed during the spray. Figures Q001.581-10 through-13 present the longer-term cell transients based on the SOFIRE analysis. The Evaluation Basis Leak (EBL) of the IHTS piping within a PHTS cell results in a maximum spill rate of 29900 gpm of 9360F sodium for a ceriod of 30 seconds. When compared to the primary system EBL of 33,500 gpm of 10150 sodium for 35.5 seconds, it is noted that each of the above para-meters for the primary system leak envelopes comparable parameters for the intermediate system. The analysis results presented for the primary system EBL therefore conservatively envelopes the intermediate system EBL 50 in the PHTS cell.
;a on
/ N) t. I Q001. 581 -14 Amend. 50 June 1979
3.2 Reactor Cavi ty Design Basis Leak The Reactor Cavity piping DBL is described in Table Q001.581-1. The total duration of ti.e DEL is 410 minutes and the total sodium injected into the cavity is 390 galloos; 521' of this sodium is injected during the first 40 minutes of the transient af ter which operator ;rtion can be assumed to have tripped the reactor and pumps. The basic method used to evaluate the *ansients resulting from this leak is identical to that described for the PilTS DBL. The only prin-cipal difference is that the Reactor Cavity DBL origirates from a cold-leg piping fault so that the temperature of the injected sodium is taken 9 Amend. 50 June 1979 Q001.581-14a
, {s o/ nn" L'_ ULJ
as 750 F, the peak cold-leg temperature. Figure Q001.581-14 presents the Reactor Cavity atmosphere pressure and temperature transients. The structural transients resulting form the DBL are presented in Figures Q001.531-15 and -16. Mcderate Energy Fluid System Leak The total duration of the Reactor Cavity MEFS Leak is 150 minutes and the total sodium injected into the cavity is approximately 20,000 gallons; 201 of this sodium is injected in the first 5 minutes after which it can be conservatively assumed that the Plant Protection System has shut-dcwn the plant. The Reactor Cavity MEFS Leak originates from a cold-leg piping fault and the temperature of the sodium injected is specified as 750 F, the peak cold-leg sodium tempertture. The same analysis procedure as described for the PHIS Cell MEFS Leak was u.,ed to evaluate this~ leak. Figure Q001.531-17 presents the results of the spray-phase transient analysis. The peak Reactor Cavity atmosphere pressure and temperature, 9.8 psig and 560 F, occur at approximately 2.5 minutes. The pressure and temperature decline gradually out to 4 minutes, the end of the maximum MEFS Leak rate (880 gpm), and begin to decrease sharply at 5 minutes when the MEFS Leak rate decreases to 120 gpm. Figures QO01.5Cl-18 through-21 present the long-term Reactor Cavity transients based on the SOF:RE analysis. Evaluation Basis Leak The Reactor Cavity EBL is a spill of the total spillable sodium in a icop (20,000 gallons). The spill rate is conservatively taken to be the normal lcap flow rate of 33,500 gpm, which yields a spill curation of 0.6 minutes. The Reactor Cavity ECL originates from a hot-leg piping failure and the temperature of the sodium injected is taken as 10150F, the maximum hot-leg sodium temperature. The same analysis procedure as described for the PHTS Cell EBL was used to evaluate this leak. Figure Q001.581-22 presents the results of the spray-phase transicnt analysis. The peak Reactor Cavity atmosphere pressure and temperaturc, 21 psig and 1020 F, occur at approximately 15 scconds. The Reactor Cavity oxygen is also depleted at roughly 15 seconds. For the EBLs, which are essentially identical for both the Reactor Cavity and PHTS Cell, oxygen depletion occurs more rapidly in the Reactor Cavity principally because the Reactor Cavity free volume is only 50% as large as that of the PHTS Cell, so that oxygen cvailable in the Reactor Cavity is roughly one-half that available in the PHTS Cell. The longer-term Reactor Cavity transients, based on SOFTPE analysis, are provided in Figures QC01.581-23 through-26. 3.3 Overflow and Prinary Sodium Storage Tank Cell Design Basis Leak 7'69 02'9 Amend. 40 0001.581-15 July 1977
Question 011.23 (11.3.2.1) In Subsection 11.3.2.1, you discuss the procedure for periodic t;ottling of Ar-39 and Kr-85 from the RAPS cryogenic still. Discuss the procedure in greater detail; provide bottle storage pressu e; discuss procedures and the means for monitoring leakage of ga , from the storage bottles; provide the anticipated onsite storage time; describe the shipping container to be used for transport of the storage bottles to a licensed burial site; and discuss the acceptability of bottled radioactive cases at the licensed burial sites. Justify your conclusion that bottling, shipping and ultimate storage of the long-lived gaseous radioisotope s (Kr-85 and Ar-39) represents a lower risk to public health and safety than releasing these isotopes under controlled and fa/orable conditions to the environment. Include your consideration of keeping occupational exposures as low as practicable.
Response
The procedure for disposing of the RAPS cryostill bottoms is discussed in revised PSAR Sections 11.3.2.1 and 11.3.4. The procedure involves controlled gradual release of the noble gases through CAPS during normal creration. Considerations of keeping occupational exposures as low as reasonably achievable supported the change to the method for disposal of the RAPS cryostill bottoms as described in PSAR Sections 11.3.2.1 and 11.3.4 269 030 Amend. 50 0011.23-1 June 1979
Ouestion 222.77 (7.8.1, 7.5-l) (RSP) Provide a Table similar to Table 7.5-1 listing all instrumentation intended for accident ard post-accident monitoring of plant parameters necessary to follow the course of an accident and to achieve a safe shutdown. Branch Technical Position EICSB 23 of Appendix 7A of the Standard Review Plan pro-vides guidance and requirements for the qualification of this type of in-strumentation. We will require compliance to these requirements. Moreover, you should provide a discussion outlining and justifying your rationale for the selection of parameters for this group of monitoring instrumentation.
Response
Section 7.5.10 provides a discussion of instrumentation provided to enable the pl3nt operator to assure that the plant is maintained in a safe shutd]wn status. Table 7.5-4 lists the parameters monitored to perform this function. Compliance to the qualification requirements of EICSB 23 50 wi l be carried out as outlined in the response to Question 222.54.
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[v/ UJl Amend. 50 Q222.77-1 June 1979
Question 310.316.4) Describe the physical location of the outside air intakes for the corr trol room ventilation systen. Indicate the locations on plant layout drawings (plan and elevation views).
Response
The interim response to this question stated that the need for a second Control Room air intake would be evaluated based upon the radiological dose rates, Control Room leakage rate, plant effluent release point locaticns and site meteorological conditions. To insure Control Room habitability following extrerrely low probability accidents which are beyond the design basu, two widely separated intakes are provided. One Control Room air intake will be located at the SW corner of the Control Building roof 49 at approximately elevation 880' and the other one will be located at the NE corner of the Steam Generator Building Auxiliary Bay roof at approximately elevation 858'. The selected air intake locations are based on the following: (1) Control Room Filter Units (a) 500 CFM outside air intake through charcoal / HEPA filter train for 1/4 inch W.G. Control Room pressurization. (b) 8,000 CFM Control Room air recirculation through same charcoal /HEPA filter train, as (a) above. (c) Redundant charcoal /HEPA filter trains with 95" charcoal and 99.977 HEPA filter efficiencies. (2) Two door vestibules for all Control Room exits / entrances. (3) 3 CFM unfiltered air infiltration based on Item 2 above. The following new and revised sections, tables and figures indicate re-visions to the design basis of the Habitability System, the addition of redundant toxic chemical and smoke detectors in the Control Room air intake duct, the increase in size of the Control Room filter trains, the deletion of water sprays for the charcoal filter banks, and the conformance to Regulatory Position 4d of Regulatory Guide 1.52: (a) Revised Section 6.3.1.1 (b) Revised Section 6.3.1.2 (c) Revised Section 6.3.1.3 (d) Revised Section 6.3.1.5 (e) Revised Table 6.3-1 ,, , _ _ ,
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(f) Revised Section 9.6.1.2 22 Amend. 49 0310.3-1 April 1979
(g) Revised Section 9.6.1.3.1. (h) Revised Section 9.6.1.3.4. (i) Revised Table 9.6-1 (j ) Revised General Arrangement Drawing 1.2-72. 22 50 e
'{ t) 3 0Jb Amend. 50 Q310.3-2 June 1979
Ques _ tion 331.20 (12.3) Describe in detail procedural and other controls that would prevent the spread of radioactive contamination from the Reactor Service Build.ng to the cold laboratory and counting room (Figure 1.2-46), and the ]ffice areas. Justify placement of the hot laboratory adjacent to the ccunting room, and the existance of a direct corridor to the office area, apparently not requiring passage through the change area.
Response
Additional information responding to this question is provided in revised sections 12.1 and 12.3. 1 33 133 50 s 269 03; Q331.20-1 Amend. 50 June 1979
Question 422.1 (13.3) 10 CFR Part 50, Appendix E, requires that the Preliminary Safety Analysis Report contain sufficient information to assure the compatibility of pro-posed emergency plans with facility design features, site layout, and site location with respect to such considerations as access routes, surrounding population distributions, and land use. To this end, we request that you submit an analysis which includes information and findings which will be needed to assure adequacy of emergency planning with respect to the protec-tive measure of evacuatic 'f persons fron the exclusion area and from any potentially affected sector of the environs, as follows: (1) Plots showing projected ground-level doses, for both whole body and thyroid, resulting from the most serious design basis accident analvm in the Safety Analysis Report. These should be based on the same isotopic release rates to the atmosphere and the same (Gaussian i m-persion model as are acceptable for use in Chapter 15 of the PMP. for the purpose of showing conformance to the siting dose criteric af 10 CFR Part 100. Data inputs to the atmospheric dispersion model should be consistent with those used and acceptable for siting dose (10 CFR Part 100) calculations. Plume front transit times, radioactive decay in transit, and dose conversion calculations may be incorporated on a physically realistic basis or a conservatively simplified basis. The bases should be fully described. Note that the plots furnished in response to this item may require ro'ision if the core disruptive accident is, at some later date, coroidered a design basis accident (D3A) and its consequences are more severe thaa the most serious DBA analyzed in the current PSAR. Present the data in the following format: (a) Use a Mg-log scale sith time (hour) following onset of release as the Ordinate, and distance (miles) fror the release point as the abscissa. (b) Provide curves for whole body doses of 1, 5, and 25 rem, and thyroid doses of 5, 25, 150, and 300 rem. Each curve should represent the elapsed time to reach the specified dose level as a
# unction of distance from the release point.
(c) Extend each curve from an ordinate of 2.0 hours or from an abscissa equal to the exclusion radius, to an ordinate of 8.0 hours or to an abscissa equal to the LPZ outer boundary, which-ever results in the greatest range of coverage. (2) The expected accident assessment time which includes the time required to identify and characterize the accident, the time requirea to predict the projected doses resulting from the accident, the time to notify nffsite authorities, and the time required to determine the approprial.e protective measures from the affected areas. Include sufficient informa-tion to support your estimate. 2h9 b Q422.1-1 Amend. 40 July 1977
(3) An estimate of the time required to notify all persons within the O potential evacuation area determined in (4) belcw, and the means assumed for such estimate. (4) An estimate of the evacuation times, measured from the time of initial warning, to remove persons from the exclusion area, and from each
" sector", or increments thereof, of the environs out to a distance determined by the 8 hour terminus of the 5 rem whole body dose curve, the 25 rem thyroid curve, or the outer LPZ boundary, which ever is the greatest. From a practical viewpoint, the "scctors" chosen for analysis may be bounded by certain gographical or manmade features, If means for effecting the but sh^uld cover an arc of at least 45 .
phys',_ 4 evacuations other than the use of private automobiles are used in estimating any of the foregoing evacuation times, these should be specified. Population data should include both resident and transient persons, including those resulting from the facilities described in Chapter 2 of the PSAP, at levels projected at approximately the time the plant is scheduled to commence operation. Provide a figure showing the ccmbined maximum daily resident and transient population for each 22-1/2 sector in increments of one mile concentiic rings from the plant out to a distance as determined above. (5) Identify any special evacuation problems that arice from the nature of the activities associated with the operation of the Oak Ridge flational Laboratory and the Oak Ridge Gaseous Diffusion Plant. (6) Provide a map or figure showing all roads available for evacuation of the plant environs out to a distance of at least 2 miles beyond that determined in (4) above. Identify the character of each road shown with respect to size, number of lanes, surface characteristics, and other factors which may affect vehicular traffic capacities.
Response
(1) The information requested in Part 1 of this question is provided in Section 13.3.12.1. Ill (2-6) Parts 2 through 6 of this question were restated in Question 422.2 40 50 and are answered in i.he response to this question. 13 5 267 03o Amend. 50 Q422.1-2 June 1979
Question 422.2 (13.3) 10 CFR Part 50, Appendix E, requires that the Preliminary Safety Analysis Report contain sufficient infonnation to assure the compatibility of proposed emergency plans with facility design features, site layout, and site location with respect to such considerations as access routes, surrounding population distributions, and land use. To complete our evaluation of the feasibility of carrying out appropriate protective measures, an analysis is normally made to determine if there is reason-able assurance that evacuation of persons in the environn of the plant can be accomplished within a time frame such that the potential conse-quences resulting from the most serious design basis accident analyzed for siting purposes would not have an adverse effect upon the public health and safety. At this time, the above analysis cannot be finalized in the absence of a source tenn acceptable to the flRC staff. However, the portion of the analysis dealing with the evacuation aspects can be completed. The following is therefore requested: (1) Provide the expected assessment time for a serious design basis accident having offsite consequences. This should include the time r: quired to identify and characterize the accident, the time required to pred1ct the projected doses resulting from the acci-dent, and the time to notify offsite authorities. Include suffi-tient information to support your estimate. (2) Provide an estimate of the time required to warn all resident and transient persons within the LPZ , and the means to be used for such warning. (our response should take into account the following facilities within the LPZ: (a) Thirteen Recreation Areas as identified in Table 2.1-14, including the stockcar track. (b) U. S. Tiuclear, Inc. (c) fluclear Environner.tal Engineering, Inc. (d) fluclear Assurance Company (e) Oak Ridge Gaseous Diffusion Plant (f) Edgewood Elementary School 960 n7. (g) Gak Ridge flational Laboratory ' Ud' (h) tielton Hill Dam (3) Provide an evaluation of the evacuation times, measured from the time of initial warning, to remove resident and transient persons from the exclusion area, and from each " sector" of the LPZ. From Q422.2-1 Amend. 25 Aug. 1976
O a p.actical vie. oin+ , the " sectors" chosen for analysis may be bceded by :erta' ..tographjca: or manmade features, but should cow a' drc of c,. least 45 . 1' means for effecting the physical eva'.uations other than the use of private automobiles are used in estimating any of the foregoing evacuation times, these should be specified, e.g., methods for nass evacuation the Edgewood Elementary School . Population data should include both resident and transient persons at levels projected at approximately the t1m the plant is scheduled to commence eperation. Note tnat the peak hour transient population at the Melton Hill Dam was omitted frrm the PSAR. This number should be provided =nd included in the
;nalysis.
(4) Provide a map or figure showing all_ roads available for evacuation of the plant environs out to a distuce of 10 miles. Identify the character of each road shown with respect to size, number of lanes, surface characteristics, and other factors which may affect vehi-cular capacities. (5) Special proolems would be associated with implementing the protec-tive measure of evacuation for some of the facilities identified in (2) above. Discuss your planning for coping with en,ergencies, including acceptable alternatives to evacuation., in the light of the possibility that (a) safeguards measures for the continued protection of SM1 at Clinch River Consolidated Industrial Park facilities and at ERDA facilities may not be compatible with evacuation of such facilities, and (b) for other reasons certain facilities may not be lef t unattended.
Response
Responses to the five (5) parts of Question 422.2 are provided in the l following PSAR Sections. I 42 Part (1) Section 13.3.12.2.1 Part (2) Section 13.3.12.2.2 Part (3) Section 13.3.12.2.3 Part (4) Section 13.3.12.2.3 Part (5) Section 13.3.12.2.3
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50 Amend. 50 June 1979 Q422.2-2
Question 422.3 Figure Q422.2-1 provided in response to item 422.2(4) is not sufficiently legible and detailed to: (1) identify all roads available for evacuation of the plant environs, (2) identify road surface characteristics from the legend, and (3) identify road widths and number of lanes. Submit the figure and information previously requested in item 422.2(4).
Response
The subject figure has been revised and re-incorporated as Figures 13.3-5, 6. Full-size maps of this figure have been provided to NRC under separa te cover. 50 9 bh b)'/ Amend. 50 $ Q422.3-1 June 1979}}