ML19339B670
| ML19339B670 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 11/07/1980 |
| From: | ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | |
| Shared Package | |
| ML19339B669 | List: |
| References | |
| NUDOCS 8011070417 | |
| Download: ML19339B670 (200) | |
Text
{{#Wiki_filter:_ .__ - . _ . ____. . .. . - _ . _ _ _ _ _ _ _._ _ _ 00836 O Department of Energy Clinch River Breeder Reactor Plant Project Office P.O. Box U Oak Ridge Tennessee 37830 i Docket No. 50-537 November 7, 1980 Mr. Darrell G. Eisenhut, Director Division of Project Management ; i Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Eisenhut:
1 AMENDMENT N0. 57 TO THE PRELIMINARY SAFETY ANALYSIS REPORT FOR CLINCH ! RIVER BREEDER REACTOR PLANT The application for a Construction Permit and Class 104(b) Operating i License for the Clinch River Breeder Reactor Plant, docketed April 10, ! 1975, in NRC Docket No. 50-537, is hereby amended by the submission of Amendment No. 57 to the Preliminary Safety Analysis Report pursuant to l 50.34(a) of 10 CFR Part 50. This Amendment No. 57 includes: updates to Section 7, " Instrumentation and Controls"; Section 11.6, "Offsite Radiological Monitoring Program"; Section 15.A, "CRBRP Radiological Source Term for Assessment of Site Suitability"; and other updates and revisions. A Certificate of Service, confirming service of Amendment No. 57 to the PSAR upon designated local public officials and representatives of the EPA, will be filed with your office after service has been made. Three signed originals of this letter and 97 copies of this amendment, each with a copy of the submittal letter, are hereby submitted. Since ly, h1 - A nond Cope n PS:80:332 ting ssis ector for Public afety Enclosure cc: Service List SUBSCRIBED and SWORN to before me Standard Distribution this SJ _ of October,1980. Licensing Distribution ~-
~H/d N6 tar ublic l/$ hPI My Commission Expires April 28, 1984 l 8011070 N'
_ ._ . . . _ _ _ _ _ _ . _ _ _ . _ _ _ _ . . _~
1 I LICE!1SIrlG DISTRIBUTIO!i Mr. Hugh Parris Manager of Power Tennessee Valley Authority 500A CST 2 Chattanooga, Til 37401 Mr. R. M. Little Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303 Dr. Jeffrey H. Broido, Manager Analysis and Safety Department Gas Cooled Fast Reactor Program P. O. Box 81608 San Diego, CA 92138 O O 8/22/80
SERVICE LIST Atomic Safety & Licensing Board flatural Resources Defense Council U. S. fluclear Regulatory Commission 917 15th Street,f4W Washington, D. C. 20555 Uashington, DC 20036 i Atomic Safety & Licensing Board Panel Dr. Cadet H. Hand, Jr., Director U. S. fluclear Regulatory Commission Bodega Marine Laboratory Washington, D. C. 20555 University of California P. O. Box 247 Mr. Gerald Largen Bodega Bay, CA 94923 Office of the County Executive Roane County Courthouse Lewis E. Wallace, Esq. Kingston, Til 37763 Division of Law Tennessee Valley Authority Dr. Thomas Cochran Knoxville, Tri 37902 flatural Resources Defense Council, Inc. 917 15th Street,fiW 8th Floor Washington, DC 20005 Docketing & Service Station Office of the Secretary U. S. fluclear Regulatory Commission Washington, DC 20535 ( Counsel for flRC Staff U. S. Iluclear Regulatory Commission Washington, DC 20555 William B. Hubbard, Esq. Assistant Attorney General State of Tennessee Office of the Attorney General 422 Supreme Court Building tiashville, Til 37219 Mr. Gustave A. Linenberger Atomic Safety & Licensing Board U. S. fluclear Regulatory Commission Washington, DC 20555 Marshall E. Miller, Esq. Chairman Atomic Safety &. Licensing Board U. S. Fluclear Regulatory Commise'nn Washington, DC 20555 Luther M. Reed, Esq. Attorney for the City of Oak P,idge p3 253 Main Street, East Q) Oak Ridge, Til 37830
i i STANDARD DISTRIBUTION i , Mr. R. J. Beeley (2) Mr. W. W. Dewald, Project Manager (2) Program Manager, CRBRP CRBRP React'or Plant Atomics International Division Westinghouse Electric Corporation Rockwell International Advanced Reactors Division P. O. Box 309 P. O. Box 158 l Canoga Park, CA 91304 Madison, PA 15663 l 2 Mr. Michael C. Ascher (2) Project Manager, CRBRP Mr. H. R. Lane (1) Burns and Roe, Inc. Resident Manager, CRBRP. 700 Kinderkamack Road Burns and Roe, Inc. Oradell, NJ 07649 P. O. Box T , Oak Ridge, TN 37830 Mr. Lochlin W. Caffey (2) , Director Mr. George G. Glenn, Manager (2) Clinch River Breeder Reactor Plant Clinch River Project P. O. Box U General Electric Company Oak Ridge, TN 37830 P. O. Box 508 ' Mr. Dean Armstrong (2) Acting Project Manager, CRBRP Stone & Webster Engineering Corp. P. O. Box 811 O Oak Ridge, TN 37830 Mr. Harold H. Hoffman (1) l Site Representative
- U. S. Department of Energy l
Westinghouse Electric Corporation Advanced Reactors Division P. O. Box 158 '
. Madison, PA 15663 Mr. William J. Purcell (2)
Project Manager, CRBRP Westinghouse Electric Corporation Advanced Reactors Division P. O. Box W Oak Ridge, TN 37830 l l O j 10/30/80
. ~
O V PAGE REPLACEMENT GUIDE FOR AMENDMENT 57 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT (DOCKET N0. 50-537) Transmitted nerein is Amendment 57 to the Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537. Amendment 57 consists of new and replacement pages for the PSAR text. Vertical lines on the right hand side of the page are used to
) identify question response information and lines on the left hand side are used to identify new or changed design information.
The following attached sheets list Amendment 57 pages and in-structions for their incorporation into the Preliminary Safety Analysis Report. l C
l AMENDMENT 57 PAGE REPLACEMENT GUIDE REMOVE THES: PAGES INSERT THESE PAGES Chapter 1 1.5-22 thru 27 1.5-22 thru 27 1.5-461, 47 1.5-461, 47 1.A-i 1.A-i 1.A-1 thru 6 1.A-1 thru 6 1.A-7 thru 29 1.A-7 thru 39 Chapter 4 4.1-1, 2 4.1-1, 2 4.2-120, 121 4.2-120, 121 4.2-138, 139 4.2-138, 139 4.2-164 thru 169 4.2-164 thru 169 4.2-174 thru 177 4.2-174 thru 177 4.2-180, 181 4.2-180, 181, 181a O 4.2-316, 317 4.2-320, 321 4.2-420, 421 4.2-316, 317 4.2-320, 321 4.2-420, 421 4.2-528, 529 4.2-528, 529 4.2-536b 4.2-590 thru 593 4.2-590 thru 593 4.4-9, 10 4.4-9, 10 4.4-127, 128 4.4-127, 128 Chapter 5 5.2-4a, 4b 5.2-4a, 4b 5.2-6, 6a, 6b 5.2-6, 6a, 6b 5.2-9b, 9c 5.2-9b, 9c 5.2-11a, 11b, 12, 12a 5.2-11a, 11b, 12, 12a 5.2-14, 14a 5.2-14, 14a, 14b 5.2-17, 17a 5.2-17 5.2-21, 22 5.2-21, 22 5.3-1, 2, 3, 3a, 4 5.3-1, 2, 3, 3a, 4 5.3-7, 7a 5.3-7, 7a 5.3-23, 24 5.3-23, 24 5.3-33b, 34 d. -33b, 34 5.3-74 5.3-74 5.3-78, 79 5.3-78, 79 5.3-82 thru 85 5.3-82 thru 85 () 5.3-112 5.3-112 A
I3 REMOVE THESE PAGES INSERT THESE PAGES
\s' Chapter 7 7i, ii 7i, ii 7vi, via 7vi, via 7ix thru xii 7ix thru xii 7.1-1 thru 6, 6a 7.1-1 thru 6, 6a 7.1-7 thru 12 7.1-7 thru 12 7.2-1, la, 2, 2a 7.2-1, la, 2, 2a 7.2-3 thru 14, 14a 7.2-3 thru 14, 14a 7.2-15 thru 24 7.2-15 thru 24 7.2-69, 70, 70a, 70aa, 70b, 70c, 7.2-25 thru 38 70d, 71 thru 78 7.3-1 thru 7 7.3-1 thru 7 7.4-8a 7.5-1, 2, 3, 3a 7.5-1, 2, 3, 3a 7.5-4, 5, 6 7.5-4, 5, 6 7.5-13, 14 7.5-13, 14 7.5-33, 33a thru 33d 7.5-33, 33a thru 33d 7.5-34 thru 39 7.5-34 thru 39 7.5-43, 44 7.5-43, 44 7.7-4, 4a, 5 thru 8 7.7-4, 4a, 5 thru 8 7.7-22 thru 25 7.7-22 thru 25 7.8-1, 2 7.8-1, 2 7.9-1 thru 7, 7a 7.9-1 thru 7
(Ss,) 7.9-8, 9, 10 7.9-8, 9, 10, 10a 7.9-11, 12 7.9-11, 12 Chapter 11 11.6-1 thru 8 11.5-1 thru 10 Chapter 12 12.3-1 thru 4 12.3-1 thru 4 Chapter 13 13.3-15 thru 19 13.3-15 thru 19 Chapter 15 15.A-1, 2 15.A-1, 2 15.A-5, 6 15.A-5, 6 15.A-11 thru 18 15.A-11 thru 18 Chapter 16 ; 16.6-11, 12 16.6-11 + A) Q, 8 i i
1 l l O ! Amendment 57 Question / Response Supplement This Question / Response Supplement contains an Amendment 57 tab sheet to be inserted following Q-1 (Amendment 56, Aug. 1980) page. Page Q-i (Amendment 57, Nov.1980) is to follow the Amendment 57 tab.
, Replacement pages for the Question / Response Supplement are listed below.
Replacement Pages Remove These Pages Insert These Pages Q7-1 Q7-1 i.
'J C
l j O l 1.5.1.1.4. Criteria of Success 41 The latch Component Test in Sodium has been completed. The Inconel gripper /Inconel coupling head performed in accord with specifications for 4-times the required number of cycles. The components are considered act.eptable. The prototype system test will confirm the capability to function 41 reliably throughout its design life. I
- 1. 5.1.1. 5. Fallback Positions l 41 For the latch mechanism, other designs could be utilized as l a fallback position. During the SCRS concept selection phase a linkage-type latch was identified as a potential fallback latch.
! 39
- l 1.5.1.2 Direct Heat Removal 41 26 g 1.5.1.2.1 Purpose 41' The Direct Heat Removal (DHR) service provides a supplementary means for removing long term decay heat for the remote case when none of the steam generator decay heat removal paths are available. This system must be l able to transfer fission product decay heat and other residual heat from the
- ' reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant boundary are not exceeded. This
, supplementary decay heat removal is performed by a cooling system incorporated into the sodium make-up/ overflow system with plant conditions as specified in Chapter 5 (Section 5.6.2). The principal uncertainty of the make-up/ overflow cooling system is short circuiting of the make-up flow with the i reactor vessel. Short circuiting would occur if the inlet fluid flows
! directly to the overflow line without cooling the reactor core. Tests are needed to design the system to ensure short circuiting does not, compromise core cooling.
j 1.5.1.2.2 Program i This program is conducted by Hanford Engineering Development Laboratory at the Integral Reactor Flow Model Test Facility. A 1/21 scale outlet plenum model test was used initially to test promising OHR candidate designs for the outlet plenum. Of concern is the location of the make-up and overflow nozzles to reduce short circuiting of make-up flow. This test will conceptually determine overflow nozzle locations. Amend. 41
; 1.5-22 Oct. 1977
Confirmation testing of the selected make-up and overflow concept was 53 successfully completed in the Phase I testing of the Integral Reactor Flow Model.
- 1. 5.1.2 . 3 Schedule 41 I b
DU m
$EE 8
$ " $ T3 5 4l CY "t M 5 a +
74 75 76 77 78 79 80 81 Make-up/0verflow Location Tests in 1/21 Scale Outlet Plenum Model 4)l a conpleted Confirmation Tests and Evaluation of Ma ke-up/0verfl ow 4jl 53l Concept , c ompleted 4]l 1.5.1.2.4 Success Criteria The tests demonstrated that the distribut. ion of the nake-up flow 2C}53 in the outlet plenum was adequate to assure that the DHR service will function to remove decay heat following a reactor shutdown. This system must be capable of removing heat at a rate such that specified acceptable fuel design limits and design conditions of the reactor coolant boundary are not exceeded. 53l 1.5.1.2.5 _Resul ts of Tests 53] The 1/21 scale outlet plenum and the HEDL IRFM medel tests have been completed. The results show that short circuiting of make-up flow to the over-53] flow nozzle is limited to approximately 5'!. The test and results are discussed in more detail in Response 001.580. 57 l 1. 5. l .3 _ Blanket Failure Threshold 41 1.5.1.3.1 Purpose The CRBRP is being designed to operate ith a limited number o"f failed
- 4) fuel and blanket rods. Thi.s requires demonstratun that operation with failed 57l blanket rods exposed to sodium does not result in rod-to-rod failure propagation.
This progrc a investigates the potential of blanket material / sodium reaction to cause swelling, flow blockages, and rod-to-rod failure 57l propagation in blanket assemblies. Amend. 57 O 1.5-23 Nov. 1980
57l It needs to be further demonstrated that a relatively small local flow 57l blockage will not lead to failure in a substantial number of blanket rods. Tests performed for core fuel rod bundle geometries (Ref. 4), indicated p that such propagation is highly unlikely. However, the geometry, thermal \d 57) conditions and flow conditions in the blanket, assemblies are sufficiently different from that in core fuel assemblies to warrant an independent evalua-57 l tion of flow blockage effects. The variation in coolant flow rates to blanket assemblies cover a wide flow velocity range from laminar through transition to turbulent modes of flow. At low flow rates and with steep temperature gradients across assemblies, buoyancy effects could become a significant contributor to the temperature and flow distribution within the blanket assembly. 57 Efforts are therefore planned to: (1) evaluate the failed blanket rod performance; specifically to verify the performance of blanket rods with failed cladding and blanket matericl exposed to sodium, 57l (2) to verify the effect of the high oxygen-to-metal ratio and density on the 57l blockages, probabilities(3)oftoblanket material evaluate sodium the cooling reaction, rate behind swelling, a solid orand flowlocal porous flow blockage with tightly arranged rod bundles with pitch to diameter ratios <l.1. 41l 1.5.1.3.2 Program This program will be conducted by Westinghouse at its ARD facility.
- 1) Failed Blanket Rod Tests 53 l The scope of the blanket rod RBCB (Run Beyond Cladding Breach) portion Q
v cf the LMFBR Reference Fuel Irradiation Program Includes the design, irradiation, and examination of EBR-II tests and/or TREAT experiments. The scope includes the acquisition, evaluation, analysis and reporting of results to: o demonstrate performance capability of breached blanket rods at beginning, middle, and end-of-life, o test the capability to accommodate design transients at end-of-life, o provide insight into the effect of reactor shutdown on breached blanket rods, and o establish a theoretical understanding of post-breach bahavior through predictive iterations based on and supported by experiments. Infonration developed from the RBCB task will support the following specific areas:
- 1. Plutonium contamination of sodium.
'2. Allowable operating time after breach.
57 3. Operating procedures for reactor shutdown and restart. O V Amend. 57 Nov. 1980 1.5-24
- 4. Delayed neutron detector values for removal of breached rods.
- 5. Operational transient margins of breached rods.
- 2) 0/M Ratio and Density Effects on Blanket Fuel-Na Reaction These effects were evaluated as part of the test program on fuel-sodium reaction phenomena conducted by General Electric Company. The results of this program are given in Reference 12.
- 3) Blanket Assembly Local Flow Blockage Evaluation The effect of a local flow blockage in a blanket rod bundle have 57 been report.
evaluated with a water flow test and will be documented in a future 1.5.1.3.3 Schedule 41 T 8 saz 33 PE% 53l S c [" CY 79 80 81 82 83 84 85 86 53l 1. Failed Blanket Rod Tests ?
- 2. 0/M Ratio and Fuel Den- (Conplete) sity Effects on Blanket Fuel-Na Reaction
' 41 ' 3. Blanket Local Flow (Complete) 57 Blockage Evaluation l 1. 5 .1. 3 . 4 Criteria of Success 57l The program will demonstrate that operation with failed blanket rods exposed to sodium does not result in rod-to-rod propagation and that a relatively small local flow blockage will not i2ad f to failure
- 57) in a substantial number of blanket rods.
Amend. 57 1.5-25 Nov. 1980
41 l.5.1.3.5 Fallback Position In the event that operating with failed blanket assemblies cannot be 57lshown to be satisfactory from a public safety viewpoint, the reactor may be required to shutdown when the blanket material is exposed to the sodium. l.5.l.4 Sodium-i mter Reaction Pressure Relief Test 41 1.5.1.4.1 Purpose The principal concern associated with the large water to sodium leak in steam generators is potential system damage, principally to the IHX by propagation of transient pressure waves through the Intermediate Heat Transport System (IHTS). The objective of the Sodium Water Reaction Pressure Relief Subsystem (SWRPRS) is to relieve pressures from the IHTS and thereby protect the primary coolant boundary from damage in the region of the primary sodium to intermediate sodium heat transfer interface. The approach to design of CRBRP IHX is to assume a conservatively large design basis water to sodium leak and to use a validated calculational method to predict pressure loads on the IHX. It is a design requirement that the IHX be able to withstand the sodium-water reaction pressures without compromising the primary coolant boundary. A survey of available existing analytical methods was completed to select the best method for improvements consistent with CRBRP requirements. The TRANSWRAP computer program (Ref. 5) was selected for use in the CRBRP p) ( analysis. An improved version of this code was used to establish loads on the IHX for the reference design IHTS piping and component arrangement and the reference design SWRPRS. A design basis leak was assumed to consist of a Double-Ended Guillotine failure (DEG) of a steam generator tube followed ininediately by the equivalent of six additional DEG tube failures. The seven tube DEG failure is not intended to represent a realistic event, but rather it provides a basis for calculating conservatively large pressure loads for the design of IHX and the pressure relief system. Results of analyses using this basis are reported in Section 5.5.3.6. To increase confidence in assuring integrity of the primary coolant boundary even during a large sodium-water reaction, the development program will provide technical information which is not available for inclusion in the PSAR. The safety related objectives of the development program are: a) to validate the computer program used to predict pressures in the IHX during a postulated sodium water reaction, and b) to confirm that effects of the design basis leak assumed for determining pressures in the IHX are conservative. O Q Amend. 57 Nov. 1980 1.5-26
Program 41 l 1.5.1.4.2 As part of the Steam Generator Development Program, AI has constructed the large Leak Test Rig (LLTR). The test programs 4j included pulling apart a notched tubular specimer in the sodium filled test article to simulate a DEG failure. A steam / water mixture was forced through 44 41 l the burst tube into the sodium. For most tests, surrounding tubes contained 1 stagnant, pressurized steam / water mixtures. In general, the development 44l effort provided technical information regarding the design of pressure relief systems to handle unexpected water-to-sodium leaks. Measured values of pressure at various locations in the test rig are being compared with calculated pressures obtained using the modified TRAllSWRAP computer program to analyze the test rig and test article. It now appears that the computer code predicts values of pressure that are either in agreement with measurcments or are conservatively large relative to measured pressures for the test rig and test article. Thus, it appears that the analysis of CRBRP for 41l sodium-water reaction pressures using this code are being conservatively 44 accomplished. This conclusion is still under review and evaluation and there-fore subject to adjustment as thm remaining test data are examined. Examination of the test article following intentional bursting of 44 l a single tube gives some indication of the nature and extent of damage propagation to other tubes. It is expected that the tests will demonstrate 47 that the calculated loadings from sodium-water reactions are conservative.
- 1. 5.1.4 . 3 Schedule
- 4) l 44.41 CY 73 74 75 76 77 78 79 80 81 LLTR-Module Steam Generator (WSG) test 47l 4j l data available Modified TRANSWRAP validated by test
- 4) l resul ts g Extent of damage 1 44 41 l in MSG evaluated Amend. 47 Nov.1978 h
1.5-27
i l
- 1. E. 2.12. 3 Schedule !
41' CY I 75 1 76 1 7T7 73 79 _ 1 1 l A. In-Vessel Model 1 : ' A A A B. Nel t-Thru Model 4 5 6 A A A C. Experimental Support - C.1 Experir: ental Studies 7 8 ir. In-Vessel Core A A Material Transport C.2 CDA Dynamic Loads 9 10 Simulation in a a 3 Scale Model Miles tones : 1 = Preliminary Design 2 = Improved Model 3 = Final Model 4 = Preliminary Model ( 5 = Improved Model ( _) 6 = Final Fodel 7 = Development of a Flashing Source 8 = Flashing Source Test Report G = CDA Dynamic Loads Simulation in a Scale Model Phase One Tests Complete 10 = CDA Dynamic Loads Simulation Phase Two Tests Complete 19 1 . 5 . 2 .12.4 Criteria of Success The model development and its associated confirmation by test results must confirm the margins of conservatism in the present analysis of radiological source terms which do not take full credit for remov il mechanisms. 1.5.2.12.5 Fallback Positions If the development program does not provide specific confirmatory information on schedule, continued use will be made of generic information extrapolated to CRBRP conditions to provide conservative estimates of radiological 30q source terms.
/-~ 5 i
)
Amend. 41 1.5-46 1 Oct.1977
1.5.3 References
- 1. IIE DL-SA-7 71, " Fuel Pin Transient Behavior Technology Applied to Safety Analysis," Presentation to AEC Regulatory Staff, 4th Regulatory Briefing on Safety Tec:hnology, t'ovember 19-20, 1974.
- 2. Ilaugen, E. G. , "Probabilistic Approaches to Design," '. liley Book Co. ,
New York (1968).
- 3. Chang, E. C. C., and C. B. Brown, " Functional Reliability of Structures,"
Journal of the Franklin Institute, September,1973.
- 4. Fontana, M.11. , et al. , "Effect of Partial Blockages in Simulated LMfBR Fuel Assemblies," Proc. Fast Reactor Safety Meeting CONF-740401-P3, 1139 (1974).
- 5. Cell, C. R. "TRAflSWRAP - A Code for Analyzing the System Effects of Large-Leab Sodium-Water Peactions in LMFRR Steam Generators," Proc.
Fast Reactor Safety 'ieeting, C0ftF-740401-P1,124 (1574).
- 6. Divislon of Reactor Research and Development, USAEC, RDT Stcndard M3-7T, "Austenitic Stainless Steel Welded Pipe (ASME SA-358 with Additional Requirements)" November,1974.
- 7. Marr, W. W. , et al . , " Subassembly-to-Subassembly Failure Propagation:
Thermal Loading of Adjacent Subassembly," Proc. Fast Reactor Safety Meeting, CO.'iF-740401-P2, 598 (1974).
- 8. Van Erp, J. B. , et al . , "An Evaluation of Pin-to-Fin Failure Propagation in LMFBR Fuel Subassemblies," ibid, 615 (1974).
- 9. Erdman, C. A. , et al . . " Improvements in Modeling Fuel-Coolant Interactions and Interpretation of the S-ll TREAT Test," ibid, 955 (1974).
10 Letter from A. Buhl to R. Boyd, " Transmittal of Additional 34 Information on SRI Scale Model Tests", SE-1788, October 26, 1976. 41l11. Letter from R. P. Denise (NRC) to L. W. Caffey (ERDA) May 6,1976. 12. GEFR-00424, UC-79B, "A Physicochemical Model for Predicting Sodium 57 Reaction Swelling in Breached LMFBR Fuel and Breeder Elements", R. W. Caputi, M. G. Adamson, and S. K. Evans, March,1979. Amend. 57 1.5-47 Nov. 1980
CHAPTER 1 - APPENDIX A Flow Diagram Symbols 1 O G , I IABLE OF CONTENTS 2 Title Page L i ne Des i g na ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l . A-1
! 16l Line Cl a s s i fi ca t i ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-2 Valve Body Symbols............................................. 1.A-4
~
Valve Actuators................................................. 1.A-7 Abbreviations Associated with Va1ves............................ 1.A-9 h2
; Va l ve O ri en ta ti on Symb ol o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A- 10 S p ec i al ty Symb o l s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,1. A - 11 i
E q u i pme n t Symbol s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A- 15 Ins trument Desi gna ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-17 I nd i ca ti ng L i g ht s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-17 Rel ay and Convers i on Device . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-19 Annunciator.................................................... 1.A-20 Instrumenta tion Identi fica tion Tab 1e . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-21 One Line & El ementary Diagrams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-22 One Line & Elementary Diagrams (Transformer Connections) . . . . . . . 1. A-28 Elementary & One Line Diagrams (Devices)....................... 1.A-29 Elementary & One Line Diagrams (Switch Contact & Miscellaneous Symbo1s)......................................... 1.A-29 Power, Grounding & Lighting Plans.............................. 1.A-30 Power, Grounding & Lighting Plans (Grounding).................. 1.A-33 Control Device Contacts (Elementary)........................... 1.A-34 Devi ce Abbrevi ations ( El emen ta ry) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. A-35 Power, Grounding & Lighting Plans (Device Abbreviations). . . .. .. . 1. A-36 l Communications................................................. 1.A-37 j Telephone...................................................... 1.A-37 i I Radiation Monitors............................................. 1.A-3.8 i Underground Distribution Plans................................. 1.A-38 Hazard Monitors - Loop / Logic Use................................ 1.A-39 57 Hazard Monitors - System Arrangement Use........................ 1.A-39 v
) Amend. 57 1.A-1 Nov. 1980 lE 1 4
, , < - - -, ~
O' LINE DESIGNATIONS aisi sin in .=ci.e s l L*' ct"55 ca m LINE CONTINUES ON j svstiu v'.".in
, su es,s " ' " ""' "
100 [250 DWG INDICATED __ li M St&A
- GIPRB - 3BEB . 33
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ZONE g C4 [,*'d"",'!" IN ZONE INDICATED PRINCIPAL FLOW ROUTE
# b (f CUT PIPE HEHE 4 d TO REMOVE ALTERN ATE FLOW ROUTE C O PIPELINE ELECTRICALLY HEATED PIPE SPECIFICATION CHANGE g (FLOW SHEET ONLY)
~*
ELECTRICAL SIGN AL BURIED PIPE
=*
,# " " ^
ff ff PNEUMATIC SIGN AL <s PNEUMATIC SUPPLY X X X L ED SYSTEM b b EL HYDR AU LIC SIGNAL FW FIELD WELD RADI ATION OR SONIC
^
^
m SIGN AL WITHOUT WIRING ! OR TUBING FW FIELD WELD IN SINGLE i
; e ; LINE DR AWING Amend. 57 1.A-1 lov. 1989 O
l l l 4
, - - - e. -
LINE CLASSIFICATIONS
- 1. PIPING CLASSES ARE DESIGNATED BY A FOUR LETTEH CODE. THE FIRST LETTER INDICATES THE PRIMARY VALVE AND FL ANGE RATING; THE SECOND LETTER THE TYPE OF MATERIAL; THE THIRD THE LETTER SYSTEM FLUlOTHE CODE TO WHICH THE PIPING IS DESIGNED AND THE FOURTH LETTER INDICATES THE DESIGNATIONS ARE AS FOLLOWL FIRST LETTER: PRESSURE RATING A - 4500 ANSI
,7 J l1 B - 2500 ANSI C - 1500 ANSI D- 900 ANSI E - 600 ANSI F - 400 ANSI G- 300 ANSI H- 150 ANSI J - 125 ANSI B16.1 K - 175 WOC UNDERWRITERS LABORATORIES, INC. L - 250 ANSI B16.1 X - GRAVITY R ATING SECOND LETTER: MATERITL A - ALLOY B - CARBON STEEL C - ST AINLESS STE E L (TP 304) D- COPPE R E - ST AINLESS STE E L (TP 316Hi F - CARBON STEEL - COPPER BEARING G- CARBON STEEL - LINED H - CAST 1RON J - CONCRETE PIPE K - VITRIFIED CLAY PIPE L - CARGON STEEL - lMPACT it ;TE D M- DURIRON N - CARBON STEEL - G ALVANIZED P - CAST lRON - CEMEN T LINED 0 - ASBESTOS - CEMENT 571 U- PCV CHROME THIRD LE TTER: DESIGN CODE A - NUCLEAR POWER PLANT COMPONENTS. ASME B&PV CODE SEC. til C. ASS I B - NUCLEAR POWER PLANT COMPONENTS. ASME B&PV CODE SEC. lit CLASS 11 C - NUCLEAR POWER PL ANT COMPONENTS, ASME B&PV CODE SEC. lit CLASS Ill D- POWER PlPING CODE ANSI B31.1.0-1967 F - N ATIONAL FIRE PROTECTION ASSOCIATION CODE G - NATIONAL PLUMBING CODE H- POWER BOILERS, ASME B&PV CODE SEC I J - AMERICAN WATE R WOR KS STANDARDS 57 Amend. 57 Nov. 1980 1.A-2 0
O LINE CLASSIFICATIONS (Continued) FOURTH LETTER: SYSTEM FLUID A - SODIUM K - AIR 8 - STEAM L - LIUOID NITROGEN C- FEEDWATER M - MIXED CAPS GASES D- NAK N~ LIQUiO ARGON E - DOWTHERM P - ACID F - ARGON R- CAUSTIC G - NITROGEN (GAS) S - ALCOHOL H - CHILLED WATE R T - HYDROGEN J - WATER V - CARBON DIOXtDE W- OIL SYSTEM /SUBSYST EM/ PIPE IN DIC ATOR 1
- 1. FIRST TWO DIGITS IDENTIFY THE SYSTEM IN WHICH THE PIPE IS LOCATED
- 2. THE NEXT TWO ALPHAS IDENTIFY THE SUBSYS TEM '.
- 3. THE LETTER "D" INDICATES TH AT THE ITEM IDENTIFIED IS PIPING .
LINE NUMBER
- 1. THE LINE NUMBER IS DESIGNATED BY SEQUENTI ALLY ASSIGNED SERIAL NUMBERS FOR RUNS OF PIPE MITHIN THAT SYSTEM (THE MAXIMUM OF FOUR DIGlTS)
SPOOL NUMBER
- 1. A SEQUENTIALLY ASSIGNED SERI AL NUMBER DENOTING PIPE SPOOLS WITHIN THE PIPE LINE.
THE ASCENDING ORDER OF THE SEOUENCE IS IN THE DIRECTION OF THE FLUID FLOW. SPECI AL CASES
- 1. IN THE PIPELINE IDENTIFICATION SYSTEM 'HE TWO-DIGIT FIELD INDICATING NOMINAL PIPE DI AMETER MAY BE EXPANDED TO EXPRESS FRACTIONS,e.g.2M INCH PIPE IS POSTED AS 2.5.
- 2. THE P!PELINE IDENTIFICATION
. MAY BE ABBREVIATED ON DRAWINGS BY REDUCING THE THREE-DIGIT SEQUENTIAL SERIAL NUMBERS FROM 001,002,003, ETC.
TO 1,2,3. ETC.
- 3. PIPELINE NUMBERS ON PIPING DRAWINGS MAY BE FURTHER ABBREVlATED BY ADDING A
< CENERAL NOTE STATING THAT"ALL PIPELINE NUMBERS APPEARING ON THIS DRAWING INCORPORATE THE SYSTEM / SUBSYSTEM DESIGNATION XXXX UNLESS OTHERWISE NOTED". THE SYSTEM /SUB-SYSTEM CHARACTERS MAY THEN DE OMITTED FROM PIPELINE IDENTI-FICATION NUMBERS POSTED ON THE DRAWING.
- 4. THE PIPELINE IDENTIFICATION . FOR PIPING DRAWINGS WILL BE ABBREVIATED FOR PIPELINE LISTS BY ELIMINATING THE 2 DIGli FIELD INDICATING
'JOMINAL PIPE SIZE AND THE 4-DIGli FIELD INDICATING PIPELINE CLASSIFICATION.
THE NOMINAL PIPE SIZE AND THE PIPELINE CLASSIFCATION WILL BE IDENTIFIED AS DATA ELEMENTS IN THE PIPELINE LIST, BUT NOT AS PART OF THE PIPELINE NUMBER.
- 5. IDENTICA. NPE LINES IN DIFFERENT LOOPS WITHIN A SYSTEM ARE IDENTIFIED BY LETTER SUFFIX ( A, B, C, etc) .
Amend. 57 1.A-3 Nov.-1980 J 1 w- - -
l VALVE BODY SYMBOLS SYSTEM NUMBER SUBSYSTEM IDENTIFIER VALVE N.O. GLOBE j
- SE0llENTIAL SERIAL NUMBER l
51 IN V 101 N.C. N.O. GATE ~ N.O. STOP CHECK h N.C. N.C.
' CHECK ANGLE AUTOMATIC CHECK POSITIVE CLOSING l BUTTERFLY O
NEEDLE l BALL Ihl PLUG FO UR-WAY TH REE-WAY FOUR WAY INDICATING FAILURE PATH 57 . 1.A-4 Amend. 57 Nov. 1980 1 0
VALVE BODY SYMBOLS (Continued) THREE WAY-lNDICATING - SAFETY OR RELIEF FAILURE PATH DIAPHRAGM AUTOMATIC BALL DRIP CHECK (SAUNDERS TYPE)
"Y" PATTERN GLOBE (BLOWDOWN)
[ AIR VENT l MANUALLY OPERATED TEST 3 HOSE GLOBE O HOSE GATE BELLOWSSEAL HOSE ANGLE PINCH l h(F FREEZE SEAL ANGLE CHECK (STOP CHECK) 57 1.A-5 Amend. 57 Nov. 1980 O i 1 _ _ _ ' ~ '
VALVE BODY SYMBOLS (Continued) SLIDE OR BLAST GATE FIRE HOSE, EXPOSED [~23 FLO AT OPERATED POST INDICATOR M HYDRANT PREACTION VALVE DRAG DISC VALVE DELUGE VALVE O ALARM CHECK VALVE 37 1.A-6 Amend. 57 Nov. 1980 0
VALVF ACTUATORS M HAND OPERATED (Mounted MOTOR OPERATED at top side or bottom of valve assembly) H HYDRAUllC f-- -- E/H ELECTRO-HYDRAULIC T T 5--- S SOLENDID OPERATED P PNEUMATIC T T DIAPHRAGM: SPRING OPPOSED, WITHOUT POSITIONER OR DIAPHRAGM; PRESSURE-BALANCE 0 OTHER PILOT O ~ b
.,,, _ DIAPHRAGM: SPRING OPPOSE ,
OVERRIDING Pit CT VALVE THAT CIAPHRAGM; SPRING OPPOSED, PRESSURIZES DIAPHRAGM WHEN ASSEMBLED WITH PILOT, ONE
,, rI ACTUATED CONTROLLED INPUT SUPPLY
- j. ., _ 1,2,3
[ SINGLE ACTING CYLINDER: SINGLE ACTING CYLINDER CONVERTER. OVERRIDING PILOT f-s*- == VALVE THAT PRESSURIZES WITHOUT POSITIONER OR g - -- I OTHER PILDT gGNAL_' DIAPHRAGM WHEN ACTUATED SUPPLY ~
- f. 2,3 f--#-- DOUBLE ACTING CYLINDER WITH DOUBLE ACTING CYLINDER F POSITIONER, CONVERTER, OVEP- "=
RIDING PILOT VALVE -- WITHOUT POSITIONER OR OTHER PILOT It!" #, . SUPPLY - w 2.3 1. NORMALLY SHUT PORT IS FILLED IN. y SINGLE ACTING CYLINDER WITH 2. OTHER COMBINATIONS ARE POSSIBLE AND WHEN _ POSITIONER USED SHALL FOLLOW THE FORMAT ESTABLISHED BY { g, THERE EXAMPLES. { 57 SIPPLv 3. ITEMS NOT SHOWN ON PhlD ' i O Amend. 57 A-7 N v.1930 i l l _ -- __ ,_
VALVE ACTUATORS (Continued) PRESSURE AND VACUUM RELIEF j$ VALVE WITH INTEGRAL PILOT. y . NOTE: ACTUATOR SYMBOL: f Press. FOR PRESSURE RELIEF OR SAFETY
, I VALVES ONLY: DENOTES A SPRING 57 WElGHT OR INTEGR AL PILOT PRESSURE REDUCING PRESSURE REDUCING REG ULATOR SELF-CONTAINED - REGULATOR WITH EXTERNAL PRESSURE TAP
! PRESSURE REDUCING REGULATOR P
r--- S WITH INTEGRAL OUTLET PRESSURE' PRESSURE RELIEF OR SAFETY RELIEF VALVE AND OPTIONAL : VALVE, ANGLE PATTERN, f PRESSURE INDICATOR (TYPICAL PS + TRIPPED BY INTEGRAL SOLEN 0ID AIR SET) ,T , DIFFE RENTIAL PRESSURE REDUCING REGULATOR BACK PRESSURE REGULATOR WITH INTERNAL AND EXTERNAL SELF CONTAINED PRESSURE TAPS h O BACK PRESSURE REGULATOR WITH EXTERNAL PRESSURE QUICK OPENER t- b& # S SULENDID RESET (OPTIONAL) l Amend. 57 1.A-8 "UV'
8 ABBREVIATIONS ASSOCIATED WITH VALVES i i N 0. NORMALLY OPEN N C. NORMALLY CLOSED LO LOCKED 0 PEN LC LOCKED CLOSED F0 Fall OPEN FC Fall CLOSED FL Fall LOCKED FI Fall INTERME DI ATE FAI Fall AS IS A0 AIR OPENS AC AIR CLOSES SOV SOLENDID OPERATED VALVE AOV AIR OPERATED VALVE MOV MOTOR OPERATED VALVE NRV NON RETURN VALVE SPECIAL CASES p 1. THE VALVE IDENTIFICATION SYSTEM MAY BE EX?ANDED TO DESIGNATE VARIOUS () TYPES OF CONTROL VALVES WITH A 4-DIGIT CODE DEFINING THE VALVE FUNCTION, THE VALVE FUNCTION CODE APPEARS IN THE " INSTRUMENTATION IDENTIFICATION TABLE" l'1 COLUMNS HEADED " CONTROL VALVE" AND SELF " ACTUATED VALVE", e.g. FCV SELF ACTUATED FLOW CONTROL VALVE FFV FLOW RATIO CONTROL VALVE LCV SELF ACTUATED LEVEL CONTROL VALVE PDV PRESSURE DIFFERENTIAL CONTROL VI.'.VE TDCV SELF ACTUATED TEMPERATURE DIFFERENTIAL CONTROL VALVE
- 2. THE THREE-DIGIT SEQUENTIAL SERIAL NUMBER IN THE VALVE IDENTIFICATION SYSTEM POSTED AS 001, 002, 003, ETC. ON VALVE LISTS MAY BE SIMPLIFIED TO READ 1,2, 3, ETC.
1 BY ADDING A GENERAL NOTE TO DRAWINGS STATING THAT ALL VALVE IDENTIFICATION NUMBERS (' TEARING UN THE DRAWING "ARE PREFIXED BY SYSTEM /SUB-SYSTEM XXXX UNLESS OTHERWISE NOTED", THE VALVE IDENTIFICATION SYSTEM ~ MAY BE FURTHER SIMPLIFIED ON PIPING DRAWlhGS BY ELIMINATION OF THE FOUR-DIGIT SYSTEM /SUB-SYSTEM CHARACTERS.
- 4. ON DRAWINGS INTENDED FOR FABRICATION / ERECTION HEVERSE VALVE INSTALLATIONS SHALL BE CONSPICUOUSLY IDENTIFIED ON THF DRAWING BY A SPECIAL NOTE,e.g. " INSTALL BACKWARDS",
INDICATING THAT THE VALVE SHALL BE POSITIONED WITH THE FLOW ARROW ON THE VALVE BODY 57 DIRECTED AGAINST NORMAL SYSTEM FLOW. (Al Amend. 57 1.A-9 tiov. 1980 L
VALVE ORIENTATION SYMBOLOGY O' NAMEPLATE SEAT - BRIDGE MARK - THE VALVE ORIENTATION SYMBOL FOR CONVEN-4 TIONAL "Y"-PATTERN GLOBE VALVES (W) DENOTES FLOW UNDER THE PLUG WHEN THE PROCESS FLUID IS MOVING FROM LEFT TO RIGHT AND OVER - - THE PLUG WHEN FLOW IS IN THE REVERSE DIREC- g , TION.THE SEAT BRIDGE MARK CLEARLY DENOTES T . VALVE STEM SEAT ORIENTATION. R ELATIVE TO SYSTEM FLOW FOR ALL GLOBE VALVE. A s CONVENTIONAL GLOBE VALVE NAMEPLATE FLOW DIRECTION ARROW _4_ ALL OTHER VALVES EXCFPT THE VALCOR LIQUID METAL VALVES SHALL UTILIZE A FLOW .
+
DIRECTION ARROW ( ~ )TO DENOTE THE ORIENTATION OF THE VALVE IN THE LINE.
~
THE ORIENTATION SYMBOL WILL BE LOCATED , BELOWTHE VALVE SYMBOL ON THE PIPING & INSTRUMENT 01AGRAMS AND INCLUDED ON THE
+ s VALVE NAMFPLATES AT THE TIME OF MANUFACTURE.
CONVENTIONAL GATE VALVE NAME N-[ PLATE STRAIGHT PATTERN LIQUID METAL VALVE (LMS) A B THE VALCOR LIQUID METAL VALVE SHALL [ "l IDENTIFY THE INLET PORT AS PORT "A" AND THE OUTLET PORT AS PORT"B". THE DESIGNATION
- ~ k SHALL BE ON THE VALVE BODY & VALVE A B NAMEPLATE.
- A " B t
l STR AIG N' PATTER N LIQUID METAL VALVE (LMS) 1.A-10 Amend. 57 Nov. 1980
O SPECIALTY SYMBO LS i I f 4 IN LINE SELFCLEANING ! u STRAINER bl Y TYPE STR AINER
@ SIMPLEX STRAINER DUPLEX STRAINER FILTER
]'[J2 =- TEMPORARY STRAINER 4
BLIND FLANGE PLUG OR CAP
~
O
-0 "x'^ s'o 'a' '
- SeECTaCLe LANoE
' ORIFICE PLATE IN QUICK i RESTRICTING ORIFICE ll CHANGE FITTING l
) m III (1) = INDOO RS VENT VENTWITH FLAME ARRESTER (0) = OUTDOORS TAPERED EXPANSION ! JOINT i l 57 i O 1.A-11 Amend. 57 flov. 1980 l
SPECI ALTY SYMBOLS (Continued) I DRIP PAN ELE 0W N - U TRAP (FOR STEAM RELIEF VALVE) LOOP SF AL ggg g SPRAY N0ZZLES
- IN LINE SIGHT FLOW FLEXIBLE CONNECTION O GLASS l
FLOW N0ZZLE QUICK DISCONNECT (TUBE OR VENTURI) 9 M BEARING REDUCER OR INCREASER i FLOOR DRAIN, HUB OR Y TRENCH % CHEMICAL SEAL l CIRCULAR OR HAMMER BLIND VAPOR TR AP, DRY l Amend. 57
- 1. A-1,c.
flov. 1980
. - . .- . - . . ~ - - , ..
O V SPECIALTY SYMBOLS (Continued) I 1 VAPOR TRAP, WET FUNNEL DRAIN 4 EXHAUST HEAD " MIXING TEE
' MAGNETIC FLOWMETER
- FLOW STRAIGHTENING VANES E.M. = ELECTROMAGNETIC P.M. = PERMANENT MAGNET IF SERVICE CONNECTION II II INSULATING FLANGE lt FL ANGE CONNECTION 11 X SPOOL PIECE I
l
+ PITOT OR PITOT ANCHOREO CONTAINMENT VENTURITUBE PENETRATION j
i BELLOWS SEAL CONTAINMENT PENETRATION
# LINED CELL PENETRATION I
i l O Amend. 57 l
\
I 1.A-13 Nov. 1980
i l SPECI ALTY SYMBO LS (Continued) N ORIFICE PLATE WITH VEN A CONTRACTA RADIUS OR ATTEMPERATOR PIPE TAPS SAFETY HEAD -- SPRINKLER N0ZZLE U d RUPTURE DISK VACUUM N/ R /\ RUPTURE DISC PRESS REllEF I i g PIPE SUPPO RT PX PIPE SUPPORT GAS SEPARATOR H I IDENTIFICATION NO.
,u , PIPE SUPPORT FOR SINGLE LINE DRAWING NE ACO USTIC (N OISE) DETECTOR
- VARIABLE SPEED COUPLING l
l l
] e CLUTCH & BRAKE I
57 Amend. 57 l Nov. 1980 i 1.A-14
- -- .~ . . . -- .- - -. . . -- . . . - - - _ - -
i t I
- O EQUlPMENT SYMBOLS 1
j J DIESEL DRIVE EDUCTOR OR EJECTOR (
+ DESUPERHEATER -
TURBINE DRIVE ELECTRIC MOTOR DRIVE CENTRIFUGAL PUMP f POSITIVE DISPLACEMENT I i O PUMP
=
_ CENTRIFUGAL 9
~
Q LIQUID CHILLER PLUGGING METER 57 c GENERAL PURPOSE
- GENERAL PURPOSE VERTICAL PUMP HORIZONTAL PUMP sa I
i ROTARY PUMP j - PISTON PUMP i O- Amend. 57 Nov. 1980 1.A-15 _,.-e - - - -w - w . , - - - , ,.---...~----v---. e----,en-. - - -,.,,---~,.rm-. v -e a -- .- -.,n - , , - - -
l l EQUIPMENT SYMBOLS (Continued) O 1 HEAT EXCHANGER _r fgg p ELECTROMAGNETIC PUMP (L10UID LIQUID)
} n
.3 l HEAT EXCHANGER (LIQUID . G AS)
[ WATER TRAP l - FLOW SENSOR SHOWER HEAD EVE WASH FLOW DALANClhG VALVE O h VERTICAL PUMP MIXER Jay
- WATER HAMMER ARRESTER BACAFLOW PREVENTER m 12C ROTOMETER _gNLINE PUMP 57 l
Amend. 57 flov. 1930 1.A-16
O INSTRUMENT DESIGNATIONS (S/161N.DI A) RELAY OR LOCAL INSTRUMENT LOCAL INSTRUMENT FOR TWO INCLUDING TR ANSMITTER ME ASURED VARI ABLES OR FOR SINGLE MEASURED MORE THAN ONE FUNCTION VARIABLE PANEL MOUNTED INSTRUMENT PANEL MOUNTED INSTRUMENT FOR SINGLE MEASURED FOR TWO MEASURED VARIABLES VARIABLE OR MORE THAN ONE FUNCTION FACE BACK VARI ABLE INTO DATA SYSTEM WHERE: (1/2 x 7/8) A = TYPE OF VARIABLE OR CONTROL SIGNAL FROM MEASUREMENT . D ATA SYSTEM _ ANNNB NNN = SERIAL NUMBEH (LOOP CD OR CHANNEL NUMBER WHENEVER POSSIBLE) B = TYPE OF INPUT SIGNAL CD = OPTIONAL PARALLEL OR REDUNDANT MEASURE-O
- WHERE T' MENTS AND PLANT PF' ~ N SIGNALS NA L "B" IS 9/16 IN. DI A.
WALL y INSTRUMENT BALLOON WITH
] [ INSTRUMENT NUMBER (WALL PDIT 101 AD F BAuO N MAY BE RUMURED E -EVENT (CONTACTSENSE)
P - PULSE (CONTACT INTERRUPT) NUMBER). ARROW INDICATES DIRECTI0fJ IN g WHICH RELAY RESPONOS TO A FAULT. , ARROW UP = FORWARD LOOKING 57 ARROW 00WN 8 REVERSE LOOKING 1 INDICATING LIGHTS 0 - OPEN COLORS 0 OPEN W WHITE 7/16 C CLOSED G-GREEN H - HIGH R RED L-LOW A AMBER C - CLOSED O G Amend. 57 f! V. 1980 1.A-17
INSTRUMENT DESIGNATIONS (CONTINUED) INSTRUMENT IDENTIFIEllIS AS FOLLOWS: NMABCDEFGXYZHJ WHERE: NM ARE TWO NUMBERS IDENTIFYING THE SYSTEM DESIGN DESCRIPTION (SD D) AB ARE TWO LETTERS REPRESENTING THE SDD SUBSYSTEM CDEFG ARE FIVE LETTERS REPRESENTING THE FUNCTION OF THE INSTRUMENTATl0N CONSISTENT WITH ISA SS.1 XYZ ARE THREE NUMBERS TO REPRESENT LOOP OR CHANNEL NUMBER (ASSIGNED BY THE COGNIZANT ENGinlEER HJ ARE TWO OPTIONAL LETTER (S) TO INDICATE REDUND ANT OR PARALLEL MEASUREMENTS WITHIN A LOOP OR CHANNEL. ALL LETTERS A THROUGH Z CAN BE USED WITH "P" AND "S" RESERVED FOR PL ANT PROTEllTIGN SYSTEM. NOTE: THE SDD AND THE SUBSYSTEM IDENTIFYER ARE NOT PLACED INSIDE THE INSTRUMENT BALLOON ON A DRAWING BUT ARE IDENTIFIED BY A NOTE ON THE DRAWING. IF MORE THAN ONE SYSTEM IS REPRESENTED ON A DRAWING, THE CONVENTION OF ISA-SS.1 SHALL APPLY. (INSTRUMENT BALLOONSIN EXAMPLE ARE ENLARGED RATHER THAN RUPTURED T0 ACCOMMODATE INSTRUMENT NUMBER) FOR EXAMPLE: TE TT TIR 105 105
~ ~ ]- ~ 56~PR g
105 I d l i l U 90 AA i 237 1 57 l Amend. 57 l Nov. 1980 I 1-A-18 ' l 1 1
l 0 RELAY AND CONVERSION DEVICE 1/P CURRENT TO PNEUMATIC E/P VOLTAGE TO PNEUMATIC E/l VOLTAGE TO CURRENT P/l PNEUMATIC TO CURRENT R/l RESISTANCE TO CURRENT E/H VOLTAGE TO HYDRAULIC P/E PNEUMATIC TO VOLTAGE b OlFFERENCE FREQUENCY METER GE GENERATOR EXCITER RELAY DESIGNATIONS N-- - - - -- ----m-- - - -
- - = . 8EF ANSI C37.2) 21 TIMER (TO BE USED WITH 21Z2) ,
21Z1 PHASE DISTANCE RELAY, ZONE 1 21Z2 FOR ZONE 2 21Z3 FOR ZONE 3 ETC. 27 UNDER VOLTAGE RELAY 40 FIELD CURRENT RELAY 42 CONTRACTOR 48 PHASE BALANCE CURRENT RELAY 49 THERMAL RELAY O 50 INSTANTANEOUS OVERCURRENT RELAY 50BF BREAKER FAILURE CURRENT DETECTOR RELAY 50CM CURRENT MONITORING RELAY 50FD PHASE FAULT DETECTOR 50G INSTANTANEOUS GROUND OVERCURRENT RELAY 51 TIME OVERCURRENT RELAY SIG TIME GROUND OVERCURRENT RELAY 51N NEUTRAL INDUCTION TIME OVERCURRENT 51V GENERATOR INVERSE TIME OVERCURRENT WITH VOLTAGE RESTRAIN RELAY 58 OVERVOLTAGE RELAY 80 FUSE FAtLURE RELAY 82 TIME DELAY RELAY 63 FAULT PRESSURE RELAY 87 DUAL POLARIZED DIRECTIONAL GROUND RELAY l 74 ALARM RELAY 79 RECLOSER 81 FREQUENCY RELAY 83 DROPOUT RELAY 85 CARRIER AUXILIARY RELAYS 85R CARRIER RECElVER RELAY FOR LINE RELAY CHANNEL 85T CARRIER TRANSMITTER RELAY FOR LINE RELAY CHANNEL 85TTS CARRIER TRANSFER TRIP RELAY SEND 88 HAND RESET LOCKOUT RELAY 87 DIFFERENTIAL RELAY D4 HIGH SPEED T91PPING RELAY OSC OSCILLOGRAPH ELEMENT TT TRANSFER TRIP 57 BFBU BREAKER FAILURE BACMUP i f% (' Amend. 57 Nov. 1980 1.A-19 1
O RELAY AND CONVERSION DEVICE (Continued)
+ BIAS -
AVG AVERAGE AVG R AVERAGE REJF.CT 1:1 B00ST
% 0 R 1:3 GAIN OR ATTENU ATE X MULTIPLY y OlVIDE t(x) FUNCTION GENERATOR ANALOG TO DIGITAL OR REV REVERSING A/O OR D/A DIGITAL TO ANALOG f INTEGRATE [ SUMMER D OR dNi DERIVATIVE OR RATE 4 SELECT LOWER
> SELECT HIGHER CP COMPUTER LIM LIMITER VOT 2 0UT OF 3 VOTER AUTOMATICALLY CONNECT, DIS-1-0 CONNECT OR TRAMSFER ONE OR V'" SELECTOR eg 1/4-10UT OF 4 MORE CIRCulTS I
PANEL MOUNTED PATCHBOARD (1/4j INTERLOCK OR MATRIX CONNECTION ANNUNCIATOR y PANEL NO. HIGH C13 _p 01 ANNUNCIATOR POINT AUDIBLE ALARM WINDOW NO. (1/16 ) L-LOW
- t. mend. 57 Nov. 1930 1.A-20
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,y 4 WHE N Q .' USE0 4S A >E.ON0 08 30CL,5EO NG L E E H .4 0EN01ES AN INIE GRAllNG fr sti fG #P sa se M30iFIER F OR EI AMPLE FO t AN 6N0'C s'NG tlCW 6NIE GH A f 0R .OR TOT AlllER)
N)'E ' m A aE :N EGR A'ING D UNs':04 5HaL L *E : HUWN W81H SEPAR ATE IDENTIFI-LA' IUN F OR EX AMrLE FQi.<8 Rs 07 F R/' 0:5 GT hit S. STARTUP AND SHUT 00WN DEVICES ARE USU ALLY SLINO, BUT MAY BE INDICATING OR RECORDING. IF S0, A00 T OR 'R' AF TER ME ASURf D VARIABLE. FOR EXAMPLE; FtS, it ist i, ,, TRS SWITCH FUNCTIONS SHALL BE FURTHER M00lFIE0 SY 'L' FOR LOW AND 'H' FOR HIGH. JI JII sy ga J8
- 6. THE DESIGN ATION 'AJ V MAY DENOTE A SC ANNING ANALYZER INDICATOR, RECORDER, AT sit av si TRANSMITTE R, ETC. BY USING THE DESIGN All0N Ajl, AJR, AJT, ETC. RESPECTIVELY, Lf Lii (G LP tY La LB
- 1. TW' OENOTES AN EMPTY THERM 0WELL.'TE'OENOTED A THERM 0nELL WITH THERMO-my yij ,, , , COUPLE OR RTO 1*d A THERM 0WELL OR A SURF ACE MOUNTE0 SHE ATHED THERMO-COUPLE.
8 FOR DEVICES OTHER THAN CONTROL VALVES,SUCH AS HYORAULIC COUPLING,VARl-of oy 0 A8LE SPE E0 0 RIVES. ETC.
" ' " " D 9 HIGH HIGH AL ARMS WILL SE DESIGNATED '( ) AHH' AND LOW LOW AL ARMS '{ ) ALL',
pg FOR EX AMPLE: L AHH OENOTE S'HIGH HIGH LEVE L AL ARM',
,7,7 ,,g ni cit p p 10- t ) K OEN0TES A CONTROL stall 04 FOR AN AN ALYZER OR OTHER INSTRUMENT SUCH AS AK FOR A CHROMATOGRAPHY PROGRAMMER OR INTR ARED CONTROL STATION RI kII *P 4Y as pg WHICH IS SEPARATE FROM THE ANALYlER ITSELF.
S' 3'I S' 5" 11. MERE SPECIAL DESIGNATION IS REQUIRE 0, PILOT LIGHTS SHALL BE IDENTIFIEDWITH tr lif ', rv vi in '
.~
48 12. PRESSURE REllEF VALVES AND RUPTURE OlSKS $ HALL BE 10ENTIF8E0 AS'PSV' AND
..y q, g3 'PSE* Rr SPE CTIVE LY l
si vir ,, ,, 11 ET REPRESENTS A P0TENTI AL TRANSFORMER I57 wT wtf ,, .
- WHE N A,0 OR E APPE AR IN TI:15 SYMBOL A = AN ALOG: 0 = OlGITAL: E =
17 A'I av na EVENT; INPUT TO COMPUTER vi vs I INSTRUME NT ATION IDENTIFICATION T ABLE Amend. 57 - Nov. 1980' D"3M e a f6 e u' "D'3'kh A XL 1.A-21 l J
t O ONE LINE & ELEMENTARY DIAGRAMS g ( GANG OPERATE 0 HIGH VOLTAGE HORN GAP SWITCH u_ LJ PRIMARY FUSE GANG 0 ERATED l \ MOTOR OPERATOR l LOAD BREAK SINGLE POLE 57
-Q l GROUNDING SWITCH A p LIGHTNING ARRESTER O
COUPLING CAPACITOR FOR MEDIUM VOLTAGE COMBINATION h ' Y k FUSED DISCONNECT SWITCH & MOTOR CONTROLLER. FULL VO LTAGE. NON-R EVERSING T COUPLING CAPACITOR WITH CARRIER CURRENT POTENTIAL DEVICE / WAVE TRAP GROUND DE1ECTION CURRENT _ TRANS. NUMBER INulCATES / CAPACITOR
\
_2 QUANT. RATIO AS NOTED GROUN'a OETECTION CURRENT TR ANS.: ERO SEQUENCE TYPE. NUMBEFt INDICATES QUANTITY
= a-RATIO AS NOTED.
1.A-22 Amend. 57 NoV. 1980
ONE LINE & ELEMENTARY DIAGRAMS (Continued) O 2 BUSHING TYPE POTENTIAL DEVICE NUMBER INDICATES ('3 BUSHING TYPE CURRENT TRANSF. NUMBER
,l QUANT. RATING AS NOTED -3 lNDICATES QUANT. RATIO AS NOTED
' PWR.TRANSF. SIZE & RATING AS POTENTIAL TR ANSF. NUMBER - N, NOTED. AA OPEN DRY TYPE.
)(
INDICATES QUANT. RATING k GA SEALED DRY TYPE, DA-AS NOTED < . NATURAL COOLING, FA FORCED k[' AIR (FAN) COOLING, FO A-FORCED OIL AIR (PUMP & AIR) COOLING REG UL ATING PWR. TRANSF.
!' GROU'ID TRANSF. &
[ SIZE RATING AS NOTED y RESISTO R, SIZE & H ATING AS NOTED n-REACTOR, SIZE & RATING GENERAL TRANSFORMER W AS NOTED 57 DISCONNECT LINKS HIGH VOLTAGE CIRCulT BREAKER, INTERRUPTING CAP., SIZE & RATING AS NOTED
\ STRESS CONE g - CABLE TERMIN ATbr1 POTHE AD COMB.3 POLE AIR CIRCulT ll: GROUND CONNECTION Aqp BREAKER & MAGNETIC CONTACTOR NUMBER INDICATES 2
SIZE i l l l Amend. 57 1.A-23 Nov. 1980
ONE LINE & ELEMENTARY DIAGRAMS (Condnued) i O - ORAWOUT DISCONNECT DEVICE gh h D A 00T E PWR.OPER. AIR CIRCulT BREAKER ORAWOUT TYPE, MAGNETIC OVER-d6 N> MP TR P, M- NNU LLY OPER. -@ ' BREAKER 3 POLE W/ THERMAL BREAKER, E ELECTRICALLY OPER. BREAKER,NA NON AUTOMATIC, & MAGNETIC TRIP U UNDERVOLTAGE ATTACHMENT C040.3 POLE AIR CKT. BKR., COMB.3 POLE AIR CKT. SKR. FULL VOLT., NON-REVERSING -
& FULL VOLT. REVERSING Ym , h SINGLE SPEED STARTER W/ THERMAL g 4 SINGLE SPEED MAG. STARTER OVERLOAO ELEMENTS. DRAWOUT 9g WITH THERMAL OVERLOAD TYPE NO. INDICATES NEMA SIZE. ELEMENTS DRAWOUT TYPE, NO.
INDICATES NEMA SIZE. COMB.3 POLE FUSED DISCONNECT SW. & FULL VOLTAGE NON. AIR CIRCUlT BKR. DRAWOUT ARTE L OVER. CE O INDICA I QUANTITY 0F POLES LOA 0 ELEMENTS. NO. INDICATES SIZE. I d PRIMARY RESISTOR REDUCED THREE WIN DING,3c PWR. VOLTAGE STARTER. NO. TRANSFORMER, SIZE & RATING p y INDICATES SIZE AS NOTED 2 l PHASE SHIFTING MOTOR - NO. IN0lCATES TRANSFORMER HORSEPOWER MOTOR GENERATOR SET l SYNCHRONOUSMOTOR SIZE & RATING ASSHOWN @ NO. lN01 CATES HORSEPOWER l RPMSTAT . MANUALLY i GPERATED RHEOSTAT MOTOR OPERATED M 0 Amend. 57 flov. 1989 1.A-24 1 l
ONE LINE & ELEMENTARY DIAGRAMS (Continued) 3 8l1l BATTERY (MULTICELL) RR (CL)
'o o] SHUNT RESISTOR o TRANSFER SWITCH M MANUAL RTD RESISTANCE TEMP. DETECT 0R OPERATED, A. AUTOMATICALLY A
OPERATED ELECTRICAL EQUIP. l SP. H l SPACE HE ATER POWER RECEPTACLF, 480V,100A 344W. T CONTROL SW.W/ RED & GREEN GENER ATOR - SIZE P IN D. LIG HTS. P IN DIC AT ES s & RATING AS NOTED PERMISSIVE. T INDICATES COMBINATION CONTROL / TEST SW. 57 START - STOP MAINTAINED START - STOP MAINTAINED 0,- O CONTACT PUSHBUTTON (4h @ CONTACT PUSHBUTTON W/ RED
& GREEN IND. LIGHTS START - STOP MOMENTARY RED & GREEN INDICATING C O C " '^ ' " S " " " " " : : ' ' " " 'S I
Amend. 57 Nov. 1980 1.A-25
i i O . ONE LINE & ELEMENTARY DIAGRAMS (Continued) i i START - STOP MOMENTARY HAND-OFF AUTOMATIC , g g CONTACT PUSHBUTTON W/ RED
@ @ SELECTOR SWITCH i & GREEN INDICATING LIGHTS i
1 l SELECTOR SWITCH W/ RED TEST SWITCH W/ RED l & GREEN INDICATING LIGHTS & GREEN INDICATING LIGHTS l L.O. CONTRO L SWITCH W/ RED
$ LOCKOUT PUSHBUTTON GREEN & AMBER IKdlCATING LIGHTS
\ i j \ INDICATING LIGHTS WITH DROPPING RESISTOR. LETTER BELL ALARM CONTACT
/ INDICATES COLOR 1
57 COMPUTER INPUT
] ANNUNCIATOR POINT b OR OUTPUT SIGNAL l
VARIABLE RESISTOR NO NORMALLY OPEN ) NC NORMALLY CLOSED ----E---- ELECTRICAL INTERLOCK t
- O l Amend. 57 1.A-26 riov. 1980 r,--- , ,,s--,v-,a--w- .-- ,- v--+, --, * ---~.-~ ,r- - - - - , , ~ , - - - - , - ----,----,a -.
O ONE LINE & ELEMENTARY DI AGRAMS (Continued)
..M--.. MECHANICAL INTERLOCK KEYlNTERLOCK
----K----
- r. .. 3 MANUAL MOTOR STARTER, C CLEAR INDICATING LIGHT I POLE WRHERMAL OVERLO AD e A AMMETER V VOLT METER 37 ADDITIONALLY FOR ELEMENTARY BLOCK DIAGRAMS SOLID CIRCLE DENOTES YM TERMINATION FOR INTERNAL WIRING O
OPEN CIRCLE DENOTES TERMINAL POINT FOR T26 EXTERNAL WIRING NUMBER ON TOP INDICATES MC2) CABLE NUMBER. NUMBER g 3 BELOW INDICATES CONDUCTOR NUMBER AND COLOR CODE OF THE CABLE ABOVE. EXAMPLE-NUMBER 3 EQUALS " RED" 37 l l l l l Amend. 57 Nov. 1980 1.A-27
O ONE LINE & ELEMENTARY DIAGRAMS (TRANSFORMER CONNECTIONS) 3 pZlG ZAG 3$ZlGZAG UNGROUNDED gp GROUNDED 3 A 3W DELTA UNGROUNDED 3 4 3W DELTA GROUNDED 3 4 4W DELTA UNGROUNDED 3p,4W DELTA GROUNDED i 3dOPEN DELTA, GND.
! 3 $DPEN DELTA ! AT COMMON PT.
O 3 $0 PEN DELTA, GND. AT 3 BROKEN DELTA
/ MID POINT O
+
-i l 3 $ WYE OR STAR, 3p WYE OR STAR, A UNGROUNDED /\ GROUNDED NEUTR AL 57
^
3 4 4W WYE OR STAR 3 ,4W WYE OR STAR,
/N UNGROUNDED RESISTANCE GND. NEUTRAL l
l Amend. 57 Nov. 1.980 1.A-28
O ELEMENTARY & 1 LINE DIAGRAMS (DEVICES) ASTERISK IN DICATES AS-AMMETER SW. PLACEMENT OF TYP. O OR [T ABBREVIATION OF CONTROL DEVICE VS VO LTAGE SW. SS SYNCHRONIZING SW. MSS METERING SE L. SW. AST E R:SK INDICAT ES PLACEMENT ASTERISK IN0lCATES 0F TYP. ABB9EVIATION Or PL ACEMENT OF TYP. (3) O* METER OR INSTRUMENT PREFIX R RECORDING ABBREVIATION OF RELAY OR DEVICE No. DENOTES QUANT. ELEMENTARY & 1 LINE DIAGR AMS (SWITCH CONTACT & MISCELLANEOUS SYMBOLS)
+- D100E -O & **""""
CONTACT, NORMALLY CLOSED l L.D. PUSHBUTTON MOMENTARY _Q @ PUSHLUTTON LOCKOUT
-O O-- CONTACT, NORMALLY OPEN _ _
T O L.
~
PUSHB UTTON CONTACTS 0F l g ,e MAINTAINED CONTACT OVERLOAD DEVICES f o TORQUE LIMIT SWITCH O $[R3$o" SIT $0NY O PROTECTIVE RELAY OR
% THERMAL E LEMENT
% SOLENOID 57 Amend. 57 tiov. 1980 1.A-29
O ELEMENTARY B 1 LINE DIAGRAMS (SWITCH CONTACT) (Continued) tI NORMALLY OPEN gy NORMALLY CLOSED 1b CONTACT (N.O.) A3 CONTACT (N.C.) l 3 MOTOR OPER. VALVE / \ INDICATING T(PE 3 3 POS. LIMIT SWITCH FUSE
-4 @ >--
CONTACTOR OR K;R AUXILIARY RELAY ^ " " OPERATING COIL ^ l A DEVICE LOCATED IN A DIFFERENT COMPARTMF'~ g WITHIN THE SWITCH C 57 {--- J OR MOTOR CONTRO' ATER POWER. GROUNDING B LIGHTING PLANS LIGHTING PANEL M SCELLANEOUS POWER I2-MOTOR DISTRIDUTION PANEL
- HORIZONTALLY MTD.
57 O MOTOR VERTICALLY MTD. TR ANSFORM E R - SIZE & RATING AS NOTED 1 l lC l CONTACTOR l MC l MOTOR STARTER OR CONTROLLER TRANSFER SWITCH ITxl SIZE & TYPE AS NOTED SIZE & TYPE AS NOTED 1.A-30 Amend. 57
?!Ov. 1980
POWER, GROUNDING & LIGHTING PLANS (Continued) R RA D CDA 480V 34 57 SOLENGID OPERATED VALVE UNIT HEATER n [p [ SABLE TRAY OR LADDER [ l SYSTEM W/ NUMBERS FOR RADIANT HEATER COMPUTER CABLE LOADING.
, , , ELEV. ARE TO BOTTOM E L. 30*-9., OF TRAY.
5 KV BUS DUCT BUS DUCT OVER 5 KV O RI I ONDulT RUN DB DIRECT BURIAL CABLE EXPOSED RIGID CONDUli EMBEDDED RIG!D CONDUlT RUN IN CONCRETE CONCEALED RIGID CONDUlf RUN , , , , , , _ ,
'' FLEXIBLE CONDuli l BELOW EL. SHOWN l
- CONDUli OR CABLE CONDUlf OR CABLE l
3 TURNING UP OR ; TURNING DOWN OR TOWARDS DBSERVER AWAY FROM OBSERVER O Amend. 57 1.A-31 M v. 1 E
) POWER, GROUNDING & LIGHTING PLANS (Continued) s./
LIGHTING FIXTURE WITH INCAN-I
' INDICATES A LETTER WHICH DESCE T R MERCURY LA PS.
D A LE ER H H IDENTIFIES FIXTURE TYPE AS C2 lDENTIFIES FIXTURE TYPE AS
*"
- SPECIFIED ON FIXTURE SCHEDULE.
SPECIFIED ON LIGHTING FIXTURE ABOVE FINISHED I Dl T8 E SCHEDULE WALL hj'g fgg NG EL' U,PPLIED FLOOR MTD. C l RCUlT NO. "2" sEILING l l , EXIT LIGHTING CEILING FLUORESCENT LIGHTING FlXTURE j j FIXT URE
, , , , , BARE LAMP AC/DC EMERGENCY FLUORESCENT STRIP LIGHTING UNIT C00SENECK LIGHTING STREET LIGHTING STANCHION & FIXTURE FIXTURE n
(,j SINGLE POLE TOGGLE SWITCH
' ^
FLOODLIGHT FIXTUhE O'[,' 0 F WS 3 IND. 3 WAY SWITCH SWITCH & SINGLE CONVENIENCE SWITCH & OUPLE X CONVENIENCE RECEPTACLE COMBINATION RECEPTACLE COMBINATION RECEPTACLE SINGLE RECEPTACLE DUPLEX A CONVENIENCE, VERTICA L A CO N VE NIEN CE, VERTICA L U SLOTS,120V,20A,3W, U S L O TS,120V, 20A, 3W, GNDED GNDED l RECEPTACLE SINGLE HOMERUN TO PANE LBO ARD - PH ASE, HO RIZO NTA L ALL UNMARKED CONDUlTS ARE S LOTS, 208V, 20 A, 3W, ~ 3/4"& CONTAIN 2 #12 UNLESS GNDED OTHERWISE NOTED. Amend. 57 floV. 1980 1.A-32
POWER, GROUNDING & LIGHTING PLANS (Continued) O J-IND. JUNCTION BOX UNDERFLOOR DUCT j~~~}--T* TB-IND. TERMINAL COX W/ JUNCTION BOX 1..j"*N J J PB-lND, PULL BOX T TELEPHONE DUCT ADD BOX NUMBER IF P-POWER DUCT REQUIRED 1 O PUSHBUTTON STATION PB3A. POWER, GROUNDING & LIGHTING PLANS (GROUNDING)
@ GROUND ROD l THERMIT WELD PROCESS ANNEALED BARE STRANDED COPPER GND. GBOUND CABLE CABLE RUN EXPOSED. RISER UP SIZE AS INDICATED ANNEALED, BARE GROUND CABLE STRANDED COPPER GND.
~~ ~~~
G DN RISER DOWN CABLE RUN CONCEALED. SIZE AS INDICATED GROUND CABLE RISER GROUND CABLE RISER. FROM UNDERMAT GND.
@ GRID PER g ; FT. LONG. TERMINATED AT GRADE FOR FUTURE CONNECTION PILE WITH GROUND WIRE GROUND TEST BOX Tc : CONNECTION TO MAY C : CONNECTION TO O
GROUND INSTRUMENT GROUND 57 1.A-33 Amend. 57 Novm L9BA --__ _---
i N CONTROL DEVICE CONTACTS (ELEMENTARY) (b ylNDICATES PLACEMENT 0F CONTROL DEVICE ONTROL DEM CONTACTS ABBREVIATION ( ENTARW (SAME AS BELOW)
'C '0F ' "
F ' CORT
- ORT LS LS COS I "
CU T-0 UT SWITCH 'CORL '00RL PRESSUR E SWITCH CS CONTRO L SWITCH , 57
' CORP '00RP T THERMOSTAT LMS LIMIT SWITCH TM TIME DELAY CLOSE INDICATES CLOSES ON INCREASE
, g TD0 TIME DELAY OPEN OF FLOW OlFFERENTIAL INDICATES OPENS ON INCREASE OPS '001F PRESSURE SWITCH OF FLOW ELECTRO PNEUMATIC INDICATES CLOSES ON g , p SWITCH RISING PHE:iSURE
'00RP INDICATES OPENS ON H HUM 10lSTAT RISING PRESSURE pgs INDICATES CLOSES ON PERMISSIVE SWlTCH ' CORT ulSING TEMPERATURE RE LAY INSTANTANEO US INDICATES OPENS ON INST T CONTACT RISING TEMPERATURE INDICATES OPENS ON RISING LEVEL As INDICATES CLOSES ON
'00RL TABULATION OR TO APPLICABLE ' CORT OESCRIPTION COLUMNS. RISING LEVEL 57 4
p Amend. 57 l Q Nov. 1980 1 1.A-34
O DEVICE ABBREVIATIONS (ELEMENTARY) SPECIAL EQUIPMENT PNEUMATIC-ELECTRIC SE PE FURN. BY MFGR. RELAY PT PRESSUR E TRANSMITTE R TE TEMPERATURE ELtMENT FI FLDWiNDICATOR FT FLOW TR ANSMITTER HtLAY TIME DELAY LEVEL TDD0 LT DROPOUT TRANSMITTER , T/C TDPU
^
THERMOCOUPLE PIC UP TZ TRANSDUCER 57 O I Amend. 57 Nov. 1980 ei l l 1.A-35
POWER, GROUNDING & LIGHTING PLANS (DEVICE ABBREVIATIONS) l
- 57 XFMR TRANSFORMER 1 INTERLOCK SWGR SWITCHG E AR STR STARTER MCC MOTOR CONTROL CENTFR HP0 HEALTH PHYSICS OF F;CE PC REHEATER POWER PANEL (AC) RHCP CONTACTOR PANEL PD POWER PANEL (DC) ATP
^
CONTROL PANEL LC LIGHTING PANEL (AC) PT POTENTIAL TRANSFORMER LD LIGHTING PANEL (DC) MTS MANUAL TRANSFER SWITCH GND GROUND OB DIRECT BURIAL CABLE ' C CONOUli LOCAL FIRE PHOTECT10N FPC PUMP CONTRO LLER EP EXPLOSION PROOF CT CUHRENT TRANSFORMER WP WEATHER PROOF POS POSITIVE 1 VT VAPOR TIGHT NEG NEGATIVE i EC EMPTY CONDUlT RL REMOTE LOCATION DT DUST TIGHT SPT S0 N A p WT WATERTIG HT (SUBMERSIBLE) MR MULTIPLE RATIO RP ISOLATEb PHASE RELAY PANEL IP BUS DUCT EBB ELECTRICAL BENCH BOARD B0 BUS DUCT MBB MECHANICAL BENCH BOARD CP CONTROL PANEL LTU LINE TUNING UNIT PE " 10 T0E L E CTRIC 57 O Amend. 57 Nov. 1980 1.A-36
COMMUNICATIONS SPEAKER
- REPRESENTS H AN D S ET.* R E PR E SENTS 4 LETTER CORRESPONDING TO g LETTEHIS) CORRESPONDING SPEAKER AND/OR AMPLIFIER TO HANDSET TYPE TYPE
_ GPEAKER AMPLiflER SPEAKER.
- REPRESENTS LE1TER 4
g
- REPRESENTS LETTER CORRESPONDING TO
- CORRESPONDING TO SPEAKER SPEAKER TYPE AMPLIFIER TYPE SPEAKER TYPE (USED WITH FOLLOWING DESIGNATIONSI A DIRECTION AL TRUMPET. 86' SOUND DISPERSION WITH 15' HORN t APPROX.); 20 BELL DIAMETER
( AFPROX.! & 3CW DRIVER. B PAGING / TALK B ACK SPEAKER 106* SOUND DISDERSION. 9' HORN ( APPROX.) 10' BELL DIAMETER ( APPROX.) C WALL MOUNTED CONE SPEAKER ASSEMBLY. WALNUT FINISHED SFEAKER BAFFLE WITH 8 Ohm 8" DIAMETER (APPROX ) CONE EF EAKER AND VOLUME CONTROL. D CORRIDOR TYPE. Bi-DIRECTIONAL BAFFLE WITH 3" DIAMETER SPE AKER AND VOLUME CONTROL. E FLUSH. WALL OR PANEL MOUNTED CONE SPEAKER ASSEMBLY WITH PROJECTING BAFFLE. 8* DIAMETER SPEAKER WITH VOLUME CONTROL. F FLUSH. SPEAKER WALL OR PANEL MOUNTED CONE SPEAKl'8 ASSEMBLY WITH FLUSH BAFFLE. 8" DIAMETER G MULTI DUTY WE ATHERPROOF HIGH. FREQUENCY SPEAKER.120' SOUND DISPERSION H DUAL, WIDE ANGLE HORN SPEAKER. 120' = 60' 3OUND DISPERSION.10* HORN ( APPROX.) 18' s 9" BELL DIAMETER (APPROX.I WITH 30W. DRIVER. SPEAKER AMPLIFIER (USED WITH FOLLOWING DESIGNATIONS) M SPEAKER AMPLIFIER ASSEMBLY PUSH-PULL. CLASS 8.12 WATT AMPLIFIER WITH INDIVIDUAL VOLUY E CONTROL; BAKED ENAMEL ON ZINC CHROMATE PHOSPHATE DONDED ENCLOSURE. N SPEAKER AMPLIFIER ASSEMBLY tAMPLIFIER TYPE SAME AS M) WEATHERPROOF ENCLOSURE. HANDSET TYPE (USE3 WITH FOLLOWING DESIGNATIONSL Q DESK-TOP ENCLOSURE. STATION WITH REMOTE HANDSET SPEAN ER. SPEAKER AMPLIFIER. VOLUME CONTROL 8 K WALL STATION WITH HANDSET. SPEAKER AMPLIFIE9 & ENCLOSURE. R DESK EDGE STATION WITH SUBSET. REMOTC HANDSET. SPEAKER AMPLIFIER & ENCLOSURE. S FLUSH PANEL STATION WITH SUBSET REMOTE HANOSET. SPEAKER AMPLIFIER & ENCLOSURE. T WEATHERPROOF WALL STATION. WITH HANDSET. SPEAKER AMPLIFIER & WEATHERPROOF g ENCLOSURE. TELEPHONE DIRECT DISPA TCH TELEPHONE COMMERCIAL TELEPHONE L0 AD DISPATCH TELEPHONE INT ERCOM SYSTEM l TS l SWIT CH80 A R D PRIVATE AUTOMATIC EXCHANGE iPAX) SO UltD POWERED TELEPHONE 8 SP 'O TELEPHONE JACK D -INDICATES DESK MTD h. W-lNulCATES WALL MTD SOUND PR00' ENCLOSED HANDSET 57, 7366-35 1.A-37 nc. 57 NOv. 1980
i l COMMUNICATIONS (CONT.) RADIATION MONITORS AREA PARTICULATE HAND & FOOT LIQUID MONITOR GAS MONITOR FRISKER MONITOR PARTICULATE /GASEQUS LAUNDRY INSPECTION MONITOR MONITOR PARTICU LATE / GASEOUS / IODINE PARTICULATE SAMPLER TRITIUM SAMPLER ALPHA MONITOR UNDERGROUND DISTRIBUTION PLANS
+ --+-- MANHOLE + HAN0 HOLE Y MH" HH#
DUCTBANK SEISMIC JulNT
"""~ -
SINGLE PHASE TRANSFORMER THREE PH ASE TR ANSFORMER 1 SIZE & RATING AS NOTED SIZE & RATING AS NOTED
]# STREET LIGHTING REGULATOR O C RATING AS NOTED 1
Amend. 57 1.A-38 Nov. 1930
HAZARD MONITORS-LOOP / LOGIC USE O HFVAPOR FIRE-lR AE HF ACID VAPOR MONITOR AE INFRARED FLAME DETECTOR FIRE-l FIRE-T THERMAL HEAT DETECTOR-AE lONIZATION SMOKE DETECTOR AE FIXED TYPE FIRE-P FIRE-TL PHOTOELECTRIC SMOKE THERMAL HEAT DETECTOR-DETECTOR
^
LINE TYPE FIRE-UV FlRE-TR THERMAL HEAT DETECTOR-AE ULTRAVIOLET FLAME DETECTOR AE RATF OF RISE TYPE HAZARD MONITORS-SYSTEM ARRANGEMENT USE AREA 1 lONIZATION SMOKE LETECTOR 1R INFRARED FLAME DETECTOR DUCT FIXED lONIZATION SMOKE DETECTOR THERMAL HEAT DETECTOR-I T (ON HVAC DUCTING) FIXED TYP. RATE PHOTOELECTRIC SMOKE THERMAL HEAT DETECTOR-DETECTOR RATE OF RISE TYPE LINE T;'ERMAL DETECTOR-UV ULTRAVIOLET FLAME DETECTOR T LINE TYPE 57 l l Amend. 57 Nov. 1980 1.A-39
. - - - - . - - . _ - - . - . - - - . _ . . _ _ - ~ - - .-- . . . _ _ - -
1 j CHAPTER 4.0 REACTOR 4 r 4.1
SUMMARY
DESCRIPTION O f The Clinch River Breeder Reactor Plant (CRBRP) uses a dxed (Pu-U) ) oxide fueled, sodium cooled fast reactor having a total thermal output of 975 Mwt. A schematic of the reactor is shown in Figure 4.2-36. The reactor i vessel, the closure head, the inlet nozzles and the core barrel are iden-tified in this figure. The core support plate and the support cone form 4 the principal pressure boundary inside the vessel. The fuel, control,
! blanket and removable shield assemblies are supported by the core support i plate which also supports a fixed radial shield. Each of these reactor
! assemblies has two load pad areas which match the elevation of the core former rings. The rings are supported by the core barrel which is welded to the core support plate. The upper internals structure, located above the core, is i supported from the intermediate rotating plug of the vessel closure and
- keyed to the upper core former ring permitting vertical motion while re-
- straining lateral and rotational motion. The structure laterally stabilizes primary and secondary control rod shroud tubes. In case of a loss of hydraulic balance, the upper internals structure acts as a secondary holddown device. L 57 The four support columns of the upper internal structure have jacks for lifting' j the upper internals structure with its keys clear of the core former ring and i reactor assemblies for refueling. The in-vessel transfer machine rotates with i i the upper internals structure for removing and replacing of reactor assemblies ,
dt refueling. t t 4 A vortex suppressor plate is provided just below the sodium .' pool surface to minimize gas entrainment in the sodium exiting from the outlet plenum. Fuel transfer and contingency storage positions are pro- , I vided in the annulus formed between the core barrel and the reactor vessel thermal liner.
- The active length of the core is 36 inches and the equivalent
, diameter is 79.5 incher. The fuel region consists of a single enrich-i ment zone with a total fissile plutonium loading of sl500 Kg. The reactor control systems include 9 primary and 6 secondary control rads. The two systems are independent and diverse. Both the systems are capable of (' shutting down the reactor from full power to hot standby conditions. The core mid-plane details are shown schematically in Figure 4.3-1. 4.1.1 Lower Internals The lower internals structure positions and restrains the reactor assemblies. The main components of the structure are: the core support structure composed of the core support plate and core barrel, horizontal i baffle, core former rings, fixed radial shielding, lower inlet modules, bypass flow modules and fuel transfer and storage assemble. These compon-51 ents are shown in Figure 4.2-36. Most of these components are also shown
- in Figure 4.2-37.
4,1 1 Amend. 57 I Nov. 1980 i O i _ _ _ ~ _ _ _ _ . _ _ _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ -
The core support cone, a component of the reactor vessel, is an inverted truncated conical shell that connects the thick circular perfor-ated core support plate with the reactor vessel wall. These two welded-in g components (plate and cone) form the upper boundcry for the reactor vessel inlet high pressure plenum. The core barrel is a thick wall right circular cylinder that extends upward from the outer edge of the core support plate to the top plane of the reactor assembly outlet nozzles. The horizontal baffle forms the upper boundary of the annular region between the core barrel and the reactor vessel and separates sodium ih the outlet plenum region from bypass flow sodium below the baffle. It limits leakage to the outlet plenum and heat flow to the bypass sodium to provide for cooling components within the annulus and the reactor vessel above the baffle while m:aimizing flow which bypasses the core. Supported inside the core barrel are upper and lower core former rings. These rings are contoured inside to the cutline formed by the outer surfaces of the upper and lower load pads of the outer row radial shield assemblies. A small gap is provided in the cold condition between the rings and shield assemblies to allow a small amount of free bow at power operation. This gap facilitates replacement of reactor assemblies at refueling. Fitted to the inside of the core barrel is fixed radial radiation shielding for protecting the barrel from structural damage from neutron fluence. There are 61 lower inlet modules for the core. A lower inlet module is shown in Figure 4.2-40. They are surrounded by six bypass flow modules. The lower inlet modules are inserted into lined holes in the core support plate. The bypass flow modules (Figure 4.2-41A) are installed on the core support plate. Each lower inlet module holds and distributes sodium coolant flow to the inlet nozzles of seven reactor assemblies. The bypass flow modules also receive low pressure coolant from the lower inlet modules and distributes the coolant to the removable radial shield assemblies. The core lattice of equilateral triangular pitch is established by the core support plate and the lower inlet modules. The inlet nozzles of the reactor assemblies are held on this lati' ice by their respective modules which allow some on-plant rotation of the nozzles as part of the core restraint sp tem. btcral above-core restraint of the reactor assem-blies is provided by the core former rings, reference Figure 4.2-37, which are located at the upper and lower hard-faced pad elevations of the reactor assemblies. These rings act on the outer row radial shield assemblies and contribute to the stable control of reactivity as the reactor power and coolant temperature are changed. In addition, they provide the lateral 51 support required to withstand seismic events. i 4.1-2 Amend. 51 Sept.1979
i l 1 l
, orific to provide coolant flow to the bypass flow module / removable 1 radial shielding and the core barrel / reactor vessel annulus. Although
'he modules are 30 year life components, they will be designed to be 3
removable. The inlet modules have the following principal functional requirements: a) Support, vertically position and restrain downward, position j and restrain horizontally the reactor assemblies during assembly,
- operation and refueling of the reactor using hydraulic balance features where required. ,
b) Distribute and provide coolant to the various reactor assemblies (fuel assemblies, blanket assemblies, control rod assemblies, and removable radial shielding). c) Provide a source of low pressure bypass coolant to the bypass
, flow module / removable radial shield assemblies and the core barrel / reactor vessel annulus.
d) Provide features to assure correct placement of the reactor assemblies in a safe location. e) Maintain a pressure boundary between the high pressure region and low pressure region within the reactor vessel and lirait the leakage flow across the boundary.
/' ' f) Provide a low impedance flow path through the LIM for the Secon-dary Control Rod System bypass flow.
g) Provide retention of loose debris greater than 0.25 in. in diameter to preclude blockage of the reactor assembly rod bundles, h) Provide for the retention of the modules during normal reactor operation using hydraulic balance features where required. i) Assure that the LIM can be removed through the Upper Internals Structure and the IVTM port. j) Maintain nominal primary coolant flow and preclude any adverse change of flow paths, k) Provide a capability to use multiple coolant flow sources in the core support structure module liner. 4.2.2.1.1.3 Bypass Flow Modules The Bypass Flow Modules (BPFM) shown in Figure 4.2-41A, are supported by the Core Support Structure (CSS). Subsequently, the BPFM 51 supports and positions the Fixed Radial Shield (FRS) and Removable Radial , 4.2-120 Amend. 53 j Jan. 1980 l i f
Shield (RRS) assemblies. The BPFM is des Jned as a permanent component. The BPFM receives low pressure coolant from the inlet modules and distributes the coolant to the RRS assemblies. A total of 6 BPFM provides a flow to a total of 264 RRS and each is designed to meet the following functional requirements. ... a) Support and position a total of 44 removable radial shield assemblies per module during reactor assembly, operation, and refueling. I b) Di s t r4 ' nd provide coolant to each of the radial shield ac~ morted. c) .ve mechanical discrimination features to insure
- p. .,r only removable radial shield assemblies into any o f t .,: receptacles in the bypass flow modules.
d) Provide a redundant flow path for the coolant which feeds the RRS assemblies. e) Support and position the FRS. f) Provide a thirty year life (22.5 full power years) with no planned maintenance. 4.2.2.1.1.4 Fixed Radial Shielding The fixed radial shielding is located inside the core barrel beyond the radius of the removable radial shielding and rests on the bypass #10w module and beneath the lowar core former. The fixed radial shielaheg is designed for the 30 year plant life. The functional requirements for this component are: a) In conjunction with the removable radial shielding assemblies, the fixed radial shielding will provide radiation protection for the core barrel and reactor vessel. This shielding will contribute to the overall reactor shielding system. The minimum ductility provided by the combination of the fixed and removable radial shielding is 10% residual ductility for the core barrel and reactor vessel. This value is based on the total elongation at end of design life and includes effects due to stretch conditions. b) Operate for a thirty year life at seventy-f:ve percent plant capacity factor. The minimum ductility limits (based on tott.1 elongation) for the fixed radial shielding are the following: Shielding Material at Attachments 10% 57 51 hend. 57 4.2-121 Nov. 1980
b) Requirement - Operate for a thirty year life at seventy-five percent piant capacity factor which also includes effects due to stretch conditions. The minimum ductility limits (based on total elongation) for the fixed radial shielding are the following: Shielding Material at Attachments 10% 57l Bases - The combination of removable and fixed radial shielding must be designed such that the fixed radial shielding remains ductile within the limits stated above. The ability of the material to yield locally would be reduced if the ductility of the material dropped below the stated limits. This situation would produce uncertainties concerning the integrity of the structure. The ductility of the structurai regions of the fixed radial shielding must be maintained at a level which will insure that the ductile failure mode analysis used in analyzing the design remains valid. 4.2.2.1.2.5 Fuel Transfer and Storage Assembly a) Requirement - Provide features which allow cooling of the stored fuel assemblies. Bases _ - The removed fuel assembly cannot be allowed to increase in temperature such that the gas pressure would build up and result in cladding failure. Additionally, a removed fuel assembly inherently contains useful infonnation which could be compromised or destroyed if the fuel temperature subsequent to removal is- permitted to exceed significantly the peak operating tempera ture. Thermal analysis has indicated that cooling of the transfer and storage assembly is required to remove the heat generated in a spent fuel assembly. b) Requirement - Provide interface features that minimize leakage with the horizontal baffle. Bases - The outlet plenum is filled with the hot core effluent, while the plenum underneath the horizontal baffle is filled with the cooler core inlet sodium. The upper portion of the fuel transfer and storage assembly is a part of the horizontal baffle and provides the guidance for insertion of the core component pot and the stored fuel into the fuel transfer and storage 51 assembly. If significant leakage flow were permitted in between l 4.2-138 ; - Amend. 57 < Nov. 1980 1 j
the fuel transfer and storage assembly and the horizontal baffle it could lead to possible insufficient material fatigue strengths to withstand the potential hot and cold flow oscillation sweeping the interface surfaces between the horizontal baffle and the fuel transfer and storage assembly. c) Requirement - Provide interface features which permit differential thermal expansion between the fuel transfer and storage assembly and the horizontal baffle. Bases - The core barrel and the fuel transfer and storage assembly will have somewhat different average temperatures. The different temperatures, which occur over a long length, could induce stresses in both the core barrel and the fuel transfer anu storage assembly if some provision is not included to allow free differential expansion between the two components. 4.2.2.1.2.6 Horizontal Baffle a) Requirement - Limit heat flow from the outlet plenum sodium to the core barrel / reactor vessel annulus sodium to an acceptable level. Bases - Bypass flow in the core barrel / reactor vessel annulus cools components within and on the periphery of the annulus as well as the vessel and vessel liner. Therefore, heat flow across the baffle must be within limits which will provide acceptable bypass cooling flow temperature to components which utilize bypass flow for cooling. b) Requirement - Form an effective hydraulic barrier between the core barrel / reactor vessel annulus and the outlet plenum sodium. Bases - Random mixing of hot sodium flowing through and exiting through the fuel region and sodium rising in the annulus between the core barrel and the reactor vesse; creates unstable flow conditions characterized by thennal plumes and/or widespread turbulence. In addition, bypass flow must be directed to provide coolant to the reactor vessel and liner above the baffle. c) Requirement - Insure that temperature oscillations on metal surfaces in or adjacent to the coolant flow paths are compatible with the horizontal baffle service life requirements. Bases - Thermal stratification exists in the flow exiting the core. The flow through the radial blanket will be cooler tnan that in the main core region. The radial blanket flow will tend to be split off at elevation just above the core barrel
, (base of the UIS). This cooler flow will mix with that circula-51 .- ting in the outlet plenum, and thennal striping can result.
Amend. 51 4.2-139 Sept. 1979
Loads from weight, hydraulic pressure drop and seismic acce-leration are transmitted by the support plate to the reactor vessel. Sizing analysis for internal pressure, flow blockage, control rod drop, and seismic loads indicate that under normal operating loads with flow blockage the inlet module meets the ASME Section III cri-teria for primary stresses. Six bypass flow modules, surrounding the lower inlet modules, distribute low pressure coolant received from the lower inlet rodules to the removable radial shield assemblies. The bypass flow modules provide receptacles to accept the removable radial shield assemblies that are not positioned in the lower inlet modules. 51 The details of the FRS are provided in Section 4.2.2.2.1.4. 54 4.2-164 Amend. 54 k May 1980
The general design rule of 5.07 minimum residual ductility insures that non-ductile fracture will not occur Juring short term loadings in reactor internal structures. This criterion is based upon the minimum residual total elongation of 10.0% and the established relationship between total and uniform residual elongation of *t = Eu + 5% as noted in Table 4.2-53. This relationship is based upon the end-of-lite tensile test data in Tables 4.2-54 through 4.2-57 and data from References 178, 179 and 180. It is conservatively based upon a data set showing the least uniform elongation for a total elongation of 10.0%. An evaluation of all current data indicates that when the degradation on ductility is greatest at a particular fluence level the uniform elongation tends to be a greater fraction of the total than this relationship indicates. Since this limit is based upon uniaxial test data a correction for the multiaxial state of stress for actual reactor component conditions is required. This correction can be performed using scientific paper 67-lD0-CODES-P1, " Applied Mechanics in the Nuclear Industry Applications of Stress Analysis". For a typical thermal stress conditions which causes an equibiaxial stress state the 5.0% would be reduced to 0.9%. The elongation available to insure ductile behavior can be determined by considering the factor of safety, consistent with the ASME Code Section III factor of safety protecting against ultimate failure. The use of the factor of safety of 3.0 would reduce the elongation for a equibiaxial state of stress to 0.307,. The applied strain considered relevant to this elongation limit is the maximum value of the three principle strains and represents an accumulation of elastic plus plastic strain at the end of life. Tnese limits vould apply at a minimum to membrane plus bending strains regardless of whether the loading is primary or secondary. Theraial transient strains in reactor in-ternal components are less than the 0.30% membrane plus bending. Therefore, from the tensile data base that is presently available, the ductility required at the end-of-life in reactor internal components is sufficient to insure their integrity when 10% residual total elongation is available and the criteria described is applied. In locations where significant fatigue damage occurs in the low cycle regime, which is also affected by the ductility of the material, corrections to the fatigue design curves are applied using accepted theories of fatigue design curve construction which are based upon reduction in area. A test program is presently in place which will experimentally characterize the fracture toughness of reactor component materials when subjected to a fast-neutron irradiation environment. This program includes tests of sc.ooth, notched and welded specimens. The establishment of the fracture toughness and fatigue crack propagation characteristics will provide a basis for confirmation of the described criteria or the substitution of a 57 more refined criteria. 4.2.2.2.1.2 Lower Inlet Module Sixty-one inlet modules support and position the reactor assemblies on the core support plate. These modules distribute the coolant to the following reactor components: fuel assemblies, 51 blanket assemblies, removable shield assemblies, control rod Amend. 57 Nov. 1980 4.2-165
^
- assemblies, core barrel, pressure vessel thermal liner, fuel transfer V) and storage assembly and horizontal baffle. Each module fits into a liner integral with the core support plate and supports and positions seven reactor assemblies while providing orificing that is unique to specific reactor assembly locations. Figure 4.2-41 shows 1/6 of the core and indicates the relative position of fuel and radial blanket assemblies and orifice zones.
The module stem acts as a strainer which collects and prevents loose debris from directly blocking the various reactor assemblies. Mechanical discrimination features are designed into each module to assure placement of the reactor assemblies into core lattice positions that will not result in assembly undercooling. Angular alignment of each module for its lattice position is assured by an alignment pin between the module liner and the module. The modules are welded 304 stainless steel structures. There are several internal configurations, excluding discrimination differences, due to the differing flow requirements of the reactor assemblies. 4.2.2.2.1.3 Bypass Flow Module The bypass flow modules shown in Figure 4.2-41 A, are functionally similar to the lower inlet modules in that they provide support and position removable radial shield assemblies and direct low
) pressure flow to cool these assemblies. There are a total of 6 identichl modules designed to rest on the core support plate and conform to the periphery formed by the 61 lower inlet modules. A flow pipe attached to the bottom of a bypass flow module mates with a hole in the core support plate. This provides a flow path for the coolant between the lower inlet module and bypass flow module.
The bypass flow modules distribute 1.22% of the total nominal reactor flow to 264 removable radial shield assemblies, 44 of which are in each module. Flow enters each of the six bypass flow modules through a bottom entry port. Each bypass flow module is hydraulically inter-connected to the adjacent two bypass flow modules giving multiple flow sources for all the RRS assemblies served by the bypass flow modules. The removable radial shield assemblies fit into receptacles integral with the bypass flow modules. These receptacles are designed with a mechanical discrimination feature to assure placement of only the removable radial shield assemblies into the bypass flow module. The low pressure existing within the region of the outer removable radial shielding results in negligible hydraulic forces and consequently a hydraulic balance system is not required. The assemblies are simply slip fitted into the receptacle permitting 51 easy insertion and removal.
~ Amend. 57 4.2-166 [Jov. 1980
\ O 4.2.2.2.1.4 Fixed Radial Shield l l l The fixed radial shield is a segmented annular rinc of type 316 ( stainless steel located between the renovable radial shielding and the core barrel as shown in Figure 4.2-42. The segments rest on the bypass flow mod-
- ules and extend upward to the lower core former structure. The segments are laterally positioned by captured pins at the lower end to the bypass
- fits; modules and at the upper end to the formu structure. The rfnning i
a rr ar.cecient acccmoda tos dif#prontial therm 1 f xcansion and ro u'ts in the O O 4.2-167 Amend. 51 Sept. 1979 l \
i O V l fixed radial shield being a simple unrestrained structure. The fixed radial shield weight is carried by the bypass flow modules and the seismic loads are transmitted through the core former structure and bypass flow modules to the core support structure. The fixed radial shield in conjunction with the removable shielding protects the core barrel and vessel from radiation damage to assure the retention of ductility for a design lifetime of thirty years. 4.2.2.2.1.5 Fuel Transter and Storage Positions Reactc refueling requires bringing new fuel into the reactor vessel and removing the spent fuel. The fuel is brought in and out through the vessel head by an ex-vessel transfer machine and is handled inside the vessel by an in-vessel transfer machina. A fuel transfer position is required to set down an assembly so that one machine can decouple and move out of the way so the other machine can grapple the assembly to continue the fuel handling operation. The component surveillance program necessitates placement The fuel transfer, fuel of sp{ecimens stora outside the e, and surveillance core barrel. specimen positions are provided by the five wells in the reactor vessel / core barrel annulus. The wells are fabricated of Type 304 stainless steel and are attached to the core barrel and the horizontal baffle. Thus all v) dead weight and earthquake loads are transmitted to the core support structure. 4.2.2.2.1.6 Horizontal Baffle The horizontal baffle shown in Figure 4.2-44 forms the upper boundary of the core barrel / reactor vessel annulus and physically separates hot sodium in the outlet plenum from the cooler bypass flow sodium in the core barrel / reactor vessel annulus. The baffle maintains the temperature of the sodium in the core barrel / reactor vessel annulus close to reactor inlet tempe-ature to reduce temperature differences across components below the baffle and to provide for decay heat removal from the irradiated reactor assemblies, stored in the fuel transfer and storage assembly. In addition, the boundary formed by the baffle forms a part of the flow path which diverts bypass flow between the reactor vessel and reactor vessel thermal liner, through uniformly spaced holes in the thermal liner below the baffle, to provide cooling for the reactor vessel and reactor vessel thermal liner. A small pressure differential must be maintained across the horizontal baffle to provide the head for this flow. The pressure differential, approximately 0.5 psi, causes leakage through the seals at the edges of the horizontal baffle base plate and at the FT&SA inlet port nozzles. However, the leakage is s limited to 12.5% of bypass fl)w to insure that sufficient cooling flow Si is provided to the vessel and vessel thennal liner. 4.2-168 Amend. 51 Sept. 1975
The horizontal baffle design incorporates a single 1.5 inch thick simply supported base plate restrained on the outer diameter by a segmented ring outer edge attachment. Each outer ring segment includes a top ring segment, a spacer block and a bottom ring segment, all of which are bolted to the vessel liner flange with a single bolt that extends through the ring segments and spacer block and into the vessel liner flange. At the inner diameter the base plate is supported on a ledge on the core barrel wall. It is held vertically by a con-57 tinuous retaining ring and located radially by a spacer ring. Circumfer-ential motion of the base plate relative to the core barrel is restrained through a key. Radial movement of the base plate is not restricted and wear resistant surfaces of Haynes 273 are provided on both sides of the plate at the inner and outer diameters and on the ring segments to accom-rrodate relative radial and angular rotation displacements due to thermal and seismic effects ct the outside diameter and angular rotation displace-I ments due to thermal effects at the inside diameter. The base plate normally operates with a 150-200 F temperature difference through the thickness. Since the upper surface is hotter, the plate will tend to develop an upwardly convex spherical curvature. The plate edges, however, are restrained vertically to the relative vertical thermal displacement between the vessel thermal liner flange I and the core barrel ledge. As a result of the vertical restraint,a i thermally induced vertical downward force will act on the vessel liner flange and an equal upward force will act on the core barrel. These vertical reaction forces provide a positive seal at the base plate outer and inner diameters. During down transients, the direction of the holddown forces can reverse due to the reversal of the through-the-thickness temperature gradient. The core barrel ledge will be in compression (down load) and the upward load at the vessel liner flange is carried through the top ring segments. i The portion of fuel transfer and storage asserably associated with the baffle consists of five penetrations through the base plate at i a radius of 85.62 inches with an inlet port nozzle at each penetration. The penetrations allow access to the portion of the FT&SA located in the core barrel / reactor vessel annulus. The nozzles are fabricated from Inconel 718 because of the thermal striping (alternate washing of a metal surface with hot and cold fluid) anticipated in the FT&SL inlet 41 i ports. 49 The horizontal baffle, except for the Inconel 718 FT&SA inlet 41 port nozzles, is fabricated from Type 316 stainless steel. 4.2-169 980
l 1 l 1 Flow patterns in the region immediately above the ; ore have been p investigated in water table tests. These tests have shown that a torroidal flow !d pattern exists in the mixing chamber located directly above the core. A large portion of the stream to stream temperature differences are reduced in this chamber before the flow exits. Temperatures in these flow streams differ sub-stantially, hence the mixing adjacent to the inner surface of the mixing chamber 57 41 results in thermal striping. The material selected for the exposed surfaces in the mixing chamber must therefore have an endurance stress limit in excess of the maximum anticipated stress amplitude produced by fluid mixing. This requirement led to the selection of Inconel 718 for the exposed surfaces of mixing chamber components. 4.2.2.2.1.8 Core Restraint System Design of the CRBRP core restraint system is based upon the limited free bow concept. Essential features of this concept are illustrated in Fig-ure 4.2-47. Fixed peripheral formers provide lateral support to the core assemblies at two locations above the active core. A third support at the core support plate elevation completes the lateral support configuration. Relief of restraint loads for refueling in the limited free bow core restraint concept is achieved by allowing the core assemblies limited freedom for unrestrained bowing during the core startup and shutdown transients. The amount of free bow permitted is controlled by sizing the gaps between core assembly load pads, and between the peripheral load pads and the n adjacent core formers. The upper bound of the allowa,le b core and former gaps is defined by a conservative analysis of the effect on critical core components () of a step compaction of the core through the range of free motion permitted by the gap configuration. The resulting core step reactivity insertion is not allowed to produce transient heating rates in the fuel which would result in the fuel pin upset condition damage limits being exceeded. It is evident that the core restraint system in its entirety includes all the reactor assemblies plus elements of the core support structure and the upper internals structure. Only the core formers, their associated retention and positioning hardward and the removable radial shield assemblies are categorized as core restraint hardware, The core formers are comprised of profile milled segments assembled into continuous rings, as illustrated in Figure 4.2-45 and centered in the 41 core barrel cavity by means of radial shims. The above core load plane former, called 'he lower core former, is mounted on a ledge machined in the inner diameter of the core barrel. A spacing cylinder provides holddown for the lower core former and support for the top load plane fonner called the upper core former. The upper core former has six lugs that fit slots in the top of the core barrel to transmit seismic and other loads to the core barrel. A series of L-shaped keys are circumferentially slipped into the groove on the inside of the core barrel, between each of the six lugs, ar.d trapped by means of a radially oriented dowel pin on either side of each slot. These L-shaped keys prevent vertical displacement of the core fonners away from the load g 49 Planes. L) Amend. 57 l 4.2-174 Nov. 1980 ! l l
4.2.2.2.1.9 Removable Radial Shield The radial shield assemblies are made up of stainless steel rods held within thin walled hexagonal ducts. These assemblies are designed to be as flexible as possible in order not to contribute to the off-power restraint loads. A close-fitting support block is inserted inside the duct at the ACLP to provide axial restraint for the shield rods and to absorb seismic loads that are transmitted through the ACLP to the core former. 4.2.2.2.1.10 Maintainability All the reactor internals except for the reactor assemblies, are designed for a 30 year life. However, provision has been made to permit removal of the lower inlet modules to assure full plant life and malfunction recovery capability. All iter,s of core support structure equipment with any significant potential for maintenance are located in the removable lower inlet modules. Items having some potential for maintenance include:
- 1. The reactor assembly receptacles, subject to insertion and removal of reactor assemblies.
- 2. Strainers and orifices, subject to coolant induced changes, such as wear or partial plugging.
4.2.2.2.1.11 Surveillance and In-Service Inspection Surveillance Material surveillance coupons are contained within special assemblies located in removable radial shield positions and a fuel transfer and storage assembly. In addition to these special assemblies, irradiated removable shield assemblies will be available for material surveillance. 4.2.2.3 Design Criteria The design criteria presented in this section are those that were in effect at the time analyses were performed. These analyses will be updated as required, to reflect the 1974 edition of the ASME Code, which provides the basic design criteria for 51 these components. Amend. 51 4.2-L 5 Sept. 1979 O
4.2.2.3.1 Lower Internals Structures (LIS) The LIS components and Core Former Structure (CFS) are evaluated as nuclear components in accordance with the rules of: The i.5ME Boiler and Pressure Vessel Code, Section III. Where these rules cannot be applied, the following rules are invoked:
- a. Code Case 1592, Class I Components in Elevated Temperature Service, Section III
- b. RDT F 9-4 Components at Elevated Temperature (Supplement to ASME Code Cases 1592,1593,1594,1595, and 1596)
- c. RDT F 9-5, Guidelines and Procedures for Design of Nuclear Systems Components at Elevated Temperatures (Non-mandatory)
- d. The special purpose strain controlled high-cycle fatigue criterion discussed in Section 4.2.2.3.2.2 may be applied to 304 and 316 austenitic stainless steel at temperatures 57 up to 1100 F.
Material properties not given in the Code are taken from the A Nuclear Systems Materials Handbook, TID-26666, and Section 4.2.2.3.3.1 below. V 4.2.2.3.1.1 Core Support Structure (CSS)
- The CSS was analyzed using the following additional rules:
- a. The 1974 Edition of the Code, Subsection NB and selected portions of Subsection NG with Addenda through Summer 1975.
- b. RDT Standard E 15-2NB, November 1974, (Supplement to ASME Code Section III, Subsections NA and NB).
- c. Modification to the high temperature design rules for Austenitic Stainless Steel - same as para. 4.2.2.3.2.2f.
4.2.2.3.1.2 Lower Inlet Module (LIM), Bypass Flow Module (BPFM), and Core Former Structure (CFS) The 1974 Edition of the Code with Addenda through Winter 1976 were used for the LIM, BPFM, and CFS analyses. 4.2.2.3.1.3 Horizontal-Baffle (HB), Fuel Transfer & Storage Assembly , (FT&SA), and Fixed Radial Shield (FRS) The HB, FT&SA, and FRS are internals structures and are not fq covered by mandatory Code rules, but the Owner's designee has required that
- g. the rules stated in 4.2.2.3.1 be applied to the design and analysis of these Amend. 57 4.2-176 Nov. 1980
i l l components. The HB and FT&SA use the 1974 Edition of the Code with Addenda through Winter 1976, and the FRS uses the 1977 Edition of the Code with Addenda through Winter 1977. 4.2.2.3.2 Upper Internals Structure (UIS) Code criteria applicable to the analysis of the UIS are divided into two categories as follows: 4.2.2.3.2.1 Class 1 Appurtenances Those portions of the UIS support columns located within the boundary of code jurisdiction for appurtenances and the ITM port plug cap shall be analyzed as Class 1 appurtenances in accordance alth:
- a. 1974 Edition of the ASME Boiler and Pressure Vessel Code, Section III, Subsection NB with addenda through Winter 1974 and,
- b. RDT Standard E15-2NB, November 1974.
4.2.2.3.2.2 Internal Structure Even though the existing ASME Code does not provide rulgs for the analysis of components operating at temperatures in excess of 800 F, those portions of the UIS located within the primary pressure boundary shall be analyzed as Class 1 components in accordance with the following: For temperatures below 800 F the 1977 Edition of the ASME Boiler and Pressure Vessel Code, Section III, with addenda through Summer 1977, Subsections NA and NG shall be used. U For temperatures in excess of 800 F the following shall be used:
- a. The 1977 Edition of the ASME Boiler and Pressure Vessel Code, Section III, with addenda through Summer 1977.
- h. Code Cases 1592, 1593, and 1594.
- c. RDT E15-2NB (Supplement to Section III).
- d. RDT F9-4 (Supplement to Code Cases 1592,1593,1594, and 1596).
- e. The Nuclear Systems Materials Handbook, TID 26666, "Inconel Alloy 718", Technical Bulletin T-39, International Nickel Company and the Alloy 718 design fatigue curve, Figure 4.2-48 shall be used.
l 49 i 4.2-177 Amend. 51 Sept. 1979
e r D To meet criterion: De < D (7) A cycle limiting criterion is required to verify the applicability of the modified rule. The effective number of al!awable design cycles { is: , n n* = i (De/ Where n is the total number of significant strain cycles between hold periods. Low amplitude high cycle strain fluc;uations (such as normal power fluctuations) need not be consider ed in it if they are i elastic excursions that result in negligible fatigue damage. i i For the modified rule to be applicable, n shall not exceed 3000 for j 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 j evaluation may be mo,dified as described herein. i i (1) The material is austenitic stainless steel Type 304 or 316 solution j treated. 1 (2) The structure does not require a Code Stamp under existing Code rules. f (3) Simplified or rigorous inelastic analysis is used. (4) Stress 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 '
l temperature may be higher than the service temperature in order to more closely simulate the actual component lifetime and the stress level. 1 I (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. j (6) The stress rupture test data include data up to 1/60 of the component i 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. 4.2-180 Amend. 51 x Sept.1979 4
~ - - . - . . . . . . . - .- . , . . - , . .
(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-fication is to use a peak ; tress to rupture design curve based upon 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 snooth speciment 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 r tigue a damage is determined by adding to the creep damage and fatige 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 4.2-47a) 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 ceacentrations and the Code Case 1592 49 stress to rupture design curve shall be used. High Cycle Strain Controlled Fatigue Limits For those 304 and 316 Stainless Steel components which are outside ASME Code jurisdiction, the fatigue damage for strain controlled cyclic deformations in excess of 1 106 cycles may be evaluated using allowable strain ranges obtained from Figure 4.2-47B, provided metal temperatures do not exceed 11000F. Fatigue life reduction factors must be applied indepen-57 dently for slow strain rates and hold times, in accordance with ASME Code requirements. 4.2-181 Amend. 57 Nov. 1980
l 1-l' Equivalent strain range shall be used to evaluate whether or not , the allowable strain range limits have been met. The equivalent strain range i value for entering the curve shall be calculated in accordance with the procedures specified in ASME Code Case 1592 except that one of the following formulations shall be used: i Formula 1. When the elastic and plastic components of the total i strain range are not known, the equivalent strain range shall be calculated i as: Ac equiv. " ! i v) - O(Cl- c2) + A(c2 - c3)2 + A(C3 - cl) 1 where v = Poisson's ratio for elastic strains. 1 Formula 2. When the elastic and plastic components of the total strain range are known, the equivalent strain range shall be calculated as the sum of i equivalent elastic strain range. OC ! equivalent "2 + v) A(c1 - c2) + A(c2 - c3) + A(c3 - c1) and equivalent plastic strain range Ac P = ! equivalent A(cl - c2) + A(c2 - c3) + A(c3 - cl) where v = Poisson's ratio for elastic strains.
, Formula 3. The following formula is included as an alternative to formula 2 l as it represents the method of calculating total equivalent strain range l employed in some computer routines, e.g. ANSYS. The total equivalent strain l
range is calculated as: j I Ac equ i v. = d A(cl - c2)2 + A(c2 - c3)2 + A(c3 - c1)2 1/2 20+vg
- where vg is a generalized Poisson's ratio found as
1
; vg = 0.5 - (0.5-v) (Es/E) t j where v = Poisson's ratio for elastic strains-
- Es = The material secant modulus prior to the last plastic strain
, increment 57 E =. The material elastic modulus
- 4.2.2.3.3 Additional Material Properties i 4.2.2.3.3.1 Inconel 718 Fatigue Properties 1
The Alloy 718 design fatigue curve, Figure 4.2-48 proposed for I 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 4.2-181a Amend. 57 Nov.-1980 i
REFERENCES (Continued) b V
- 14. P. E. Blackburn, A. E. Martin, J. E. Battles, and others, " Sodium-Fuel Interactions," in Proceedings of the Conference on Fast Reactor Fuel Element Technology, pp. 479 493, American Nuclear Society, Hinsdale, Ill.
(1971).
- 15. P. E. Blackburn, C. E. Johnson, I. Johnson, and others, " Chemical Engineer-ing Division Fuels and Materials Chemistry Semiannual Report, July-December, 1971," ANL-7877, April, 1972.
- 16. M. G. Chasanov, C. E. Johnson, J. E. Battles, and others, " Chemical Engineering Division, Fuels and Materials Chemistry Semiannual Report, January-June, 1972," ANL-7922, October, 1972.
- 17. M. G. Chasanov, C. E. Johnson, P. E. Blackburn, and others, " Chemical Engineering Division Fuels and Materials Chemistry Semiannual Report, July-December, 1972," ANL-7977, April, 1973.
- 18. P. E. Blackburn, C. E. Johnson, J. E. Battles, and others, " Chemical Engineering Division Fuels and Materials Chemistry Semiannual Report, January-June,1971," ANL-7822, September,1971.
- 19. T. L. Markin and E. J. McIver, " Thermodynamic and Phase Studies for Plutonium and Uranium-Plutonium 0xides with Application to Comnatibility Calculations," in Plutonium 1965, pp. 845-857, The Institute n ~ Metals, London,1967.
- 20. N. A. Javed and J. T. A. Roberts, " Thermodynamic and Defect-Structure Studies in Mixed-Oxide Fuels," ANL-7901, February,1972.
- 21. R. E. Woodley, " Equilibrium Oxygen Potential Composition Relationships in O.75Pu0.2502-x' " HEDL-SA-269, 1970.
U
- 22. M. H. Rand and T. L. Markin, "Some Thermodyn;rnic Aspects of (U, Pu) 0 2 Sodium Solutions and Their Use as Nuclear Fuels," in Thermodynamics of Nuclear Materials, 1967, pp. 637-650, International Atomic Energy Agency, Vienna, 1968.
- 23. M. G. Chasanov, C. E. Johnson, N. D. Dudey, and others, " Chemical Engineering Division Fuels and Materials Chemistry Semiannual Report, January-June,1973," ANL-8022, December,1973.
- 24. M. Housseau, J. F. Marin, F. Anselin, and others, " Etude Des Reactions i Entre Le Combustible (U, Pu) 0 Et Le Sodium" in Fuel and Fuel Elements for Fast Reactors, Vol. 1, pp.2277-291, International Atomic Energy Agency, Vienna, 1974.
- 25. D. L. Smith, "0xygen Interactions Between Sodium and Uranium-Plutonium 0xide Fuel," Nucl. Technol. 20, pp. 190-199, December, 1973.
V. Amend. 51 Sept. 1979 4.2-316
l l REFERENCES (Continued) t 26. K. E. Gregoire, P. E. Novak, R. E. Murata, " Failed Fuel Performance in Naturally Convecting Liquid Metal Coolant", GEAP-13620, June, 1970.
- 27. K. Q. Bagley, E. Edmonds , H. J. Powell, and others , " Fuel Element l
Behavior Under Irradiation in DFR", in Fuel and Fuel Elements for- } Fast Reactors , Vol .1, pp. 87-100, International Atomic Energy Agency, Vienna, 1974. { 28. V. R. Marian and J. V. Marron, " Core Restraint System Tests and Analysis, Phase I: Planar Externally Clamped Core", GEAP-10798, i March, 1973, i i
- 29. "0xide Fuel Element Development Quarterly Progress Report for Period Ending June 30, 1974", WARD-0X-3045-14, Sep tembe r, 1974.
(Availability: USAEC Technical Informm ion Center). ! 29a. HEDL-TME-73-46, " Flow Induced Vibration Tests of a Prototypic FFTF Control Rod Absorber Pin Bundle Aseembly", June 1973. l 16
- 30. Deleted ii i 57
- 31. C. R. Brinkman and G. E. Korth, " Strain Fatigue and Tensile Behavior of Inconel 718 from Room Temperature 650 C", ANCR-ll25, August, 1973.
- 32. "AEC Fuels and Materials Development Progress Report No. 79",
GEMP-1013, June, 1969.
- 33. Rules for Construction of Nuclear Power Plant Components, ASME Boiler and Pressure Vessel Code, Section III, The American Society o f Mechanical Engineers, New York,1971.
, 34. R. T. Collins, G. W. Smith, and E. E. Spier, " Manual for Structural l
Stability Analysis of Sandwich Plates and Shells", NASA-CR-1457, l December, 1969. l 35. P. Seide, "A Survey of Buckling Theory and Experiment for Circ ular
; Crnical Shells of Constant Thickness", in Pressure Vessels anc Piping: Design and Analysis, Vol. 1, pp. 590-615, The American Society of Mechanical Engineers, New York,1972.
- 36. G. Gerard, Introduction to Structural Stability Theory, McGraw-Hill, New York, 1962.
- 37. V. 1. Weingarten and P. Seide, " Buckling of Thin-Walled Truncated Cones", NASA-SP-8019, September,1968.
l' Amend. 57 t 4.2-317 Nov. 1980 i i
REFERENCES (Continued) G
- 54. " Microstructural Dependence of Failure Threshold in Mixed Oxide LMFBR Fuel Pins", HEDL-TME 75-9, October, 1974.
55.
" Mechanical Properties During Simulated Overpoger Transients of Fast Reactor Cladding Irradiated from 700-1000 F", HEDL-TME 75-28, June,1975.
- 56. " Base Technology FSAR Support Document-Prefailure Transient and 57l Failure Threshold Behavior", HEDL-TME 75-47, November, 1975.
- 57. " Evaluation of FFTF Fuel Pin Transient Design Procedures", HEDL-TME 75-40, August, 1975. (See also reference 173.)
*58.
D. C. Jacobs, "The Development and Application of a Cumulative flechanical Damage Function for Fuel-Pin Failure Analysis in LMFBR Systems", CRBRP-ARD-57 0115, August 1976.
*59. R. Sim, A. Veca, "FFTF Fuel Pin Final Design Support Document",
Westinghouse Advanced Reactors Division Internal Memorandum, FCF-213, December,1971. p. 2.1 and 2.2.
*60. R. D. Coffield, P. L. Wattelet, "An Analytical Evaluation of Fuel ~T Failure Propagation for the Fast Flux Test Facility", WARD-2171-ll, December, 1970.
(L)
- 61. W. H. McCarthy, K. J. Pecry, G. R. Hull, J. W. Bennett, " Examination of F3A Series Unencapsulated Mixed-0xide Fuel Pins Irradiated in EBR-II", Nuclear Technology, 16, pp 171-186 (1972).
- 62. B. F. Rubin et. al., "Fucl-Cladding Mechanical Interaction in LMFBR Fuel Rods", Nuclear Technology 16, pp 89-99 (1972).
- 63. R. F. Hilbert, E. A. Aitken, P. R. Pluta, "High Burnup Performance of LMFBR Fuel Rods - Recent GE Experience", Transactions of the American Nuclear Society 20, pp 315-318 (1975).
- 64. R. G. Sims, S. Vaidyanathan, " Cladding Loading Sensitivity for LMFBR Fuel Rods", GEAP-13969, January,1974.
- 65. G. E. Culley, J. E. Hanson, W. L. Partain, J. H. Scott, A. E. Wal ter,
" Fast Reactor Safety Implications of Recent Assessments of Fuel Pin Transient Behavior", Proceedings of International Conference, Engineering of Fast Reactors for Safe and Reliable Operation, 11 pp. 626-641, Gesellshaf t Fur Kernforschung M.B.H. , Karlsruhe, 1973. (CONF-72-1006).
1
- References annotated with an asterisk support conclusions in the Section.
dl Other references are provided as background information. b Amend. 57 4.2-320 Nov. 1980
l l REFERENCES (Continued) ' t
- 66. S. Chen, " Vibrations of Nuclear Fuel Bundles", Nuclear Engineering and Design 35, pp 399-422 (1975).
- 67. K. G. Eickhoff, C. Betts, F. E. Buckley, C. E. Yliffe, G. McAreavey, ;
J. O. Poinder, H. Hughes, D. White, " Theoretical and Experimental ; Studies Supporting the Design of Fast Reactor Fuel Elements", ' Proceedings of a Symposium on Fuel and Fuel Elements for Fast Reactors I, pp 329-149, International Atomic Energy Agency, Vienna, 1974.
- 68. J. Van Miegroet, D. Dewaldelcer, A. Michel, K. Kummerer, G. Van Massenhove, " Design and Preparation of the MOL-78 Irradiation Experiment", Proceedings of a Symposium on Fuel and Fuel Elemeqts for Fast Reactors I, pp 163-183, International Atomic Energy 51 Agency, Vienna, 1974.
l Amend. 53 Jan. 1980 4.2-321
s_/ . 'x TABLE 4.2.5 2 REFERENCES SUPPORTING CRBRP FUEL ROD LOADING (See Resconse to Question 241.68) Exemplary References , Loading FFTF EBR-II U.S. LMFBR Foreign L!!FBR Category Design Madiation Proaram Program I. Fuel-Cladding Differen- 59 62 64, 65 67, 68 tial Expansion
- 2. Fission Gas Released 59 61, 63 64, 65 67, 68 from Fuel
- 3. Differential Thermal 59 ---
64 67, 68
& Irradiation Induced Expansion a
- 4. Support System Inter-1 action 5
A. Fuel Rod-Spacer 59 63 64 67 B. Bundle-Duct 59 63 64 ---
- 5. Flow Induced Vibration 59 --- 66 67
- 6. Accident Loading A. Fission Gas Ejection 60 ---
65 69 B. Fuel Coolant 60 --- --- 69 32 Interaction en N e" "e G$ y < 12
m U N N N N w M ^ N N""" N N N" N - N m N N N m O O O N O O h Cm Uu C% .w. ;J, e e e O r O m N n m O O N 45 67 C 2V d e m O
= cC O-N X M X X e W 6 g X X x x C O 3 _J C M M
G W w ea w C.". au . *f. M. o. O. m O. C. O. O. O w O 7 Q M m u1 " L m M y)
- U C
C1 C e Ge O es 3 as - - GJ G3 % ed e q Q = 7 *J 1 kt O O O o O o o O O O O
- C~O C'~* m e r- e-* " '-* n #
e r= e M ~~ Uo H Q C LLJ C - H L&J s M U U C
~5
. C10 %e Q
s 80 L LM L C Q> M M c) CD ** W tt3 M ,p ep wy. o3 o /'. O. O. O. O. O. O. O. O. O.
> 0 O O O. O.
LAJ w,C v O O O O O O O O O
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-w C - +J oa m e r- " r= e e m 6; O X Q "3 % X X X X
% x X X X X X X r3 >~ - C O - r 6 %.-
y) g > M. M. e. .a. e. v. . 10 N. us. v. (f) s C 43 O r= M M L'3 Y N t > W - m U1 N Z C
. W O V w L *J 7~ a O Cm O O O O y
M O O O O O O O O _c g C_; = L u) O C C O O O O O O
., J v.-< L D - g3 o c3 a s c3 y =g a ca 2 CO H 3, ~~-w se g < < L c-H ~ L . -
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- g-CL ? :[5 7a 3t bm O O O O O O O O O O O N [ a cC O
N O N O O O O O O O O O
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J
< C3 CO c3
- *P CD c w - e - C3 er - CC 7 O O ~ ~ N O y .m c1 M M M U M t%
U O M M O N U O M W b M M M M C M C M
*J M M M M C M H M t.1 W W m 4/3 m = W
.T *s M
Z E O H t U a q e g c C O m L O L w 4P C3 .s a: O - es es e O E C U M - G C e 3 C -- O L U C C U Q --. m J c) CC C y MC C 3 m " O Q 3 C c 9 ~3 w 2 O O w M "3 L U 23 O 23 a
+J L
C=* C - U- O C C. Nm O U.T M E O 4 o d . C; a 7- 6 %C l.a O 3o O u C "U O V OO C 63 O V CL C L O c- L "3 c C C.
- L dZ 41 O Q1 O O d Oi-3 O cJ Q 3 H u l LO C C to
~
O O I t 23 d W J J vr _.J e a D oC L
- c. L rJ ~3 x, 3 C Lu -u o m o
= a w0 vn- La- w -
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+ k1 E r-O O3 3 c-- V ft3 33 Q. O O ay U LA. M O C O >,e= 0 M
Lt. Cf LA Jw1 cc LA. E 4 r% s to tn a rs 4.2-421 Amend. 57 m Nov. 1980
0 1 l 1, ) l .I I 1 i i i FIGURE 4.2-43 DELETED G 1 i I i I l l I t 1 . i i l i t 4.2-528 Amend. 51 , Sept. 1979 1
FT&SAINLET PORT N0ZZLE pp OUTLET FORMER PLENUM RESTRAINT D SEGMENTED RING 7 s OUTER EDGE j \j ATTACHMENT
/ ,7 WEAR RESISTANT N s [
/ s
\ SURFACES s
/ /// // / / f sS) / \
/ N'NC : : '
h
/ HORIZONTAL /
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PLATE ' 8
/
FT&SA __._ m, / /
/ -
THERMAL LINER
/
CORE BARREL / REACTOR VESSEL [
/ l Q
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/
SURFACES / j/I
- SPACER RING p
N REACTOR VESSEL WALL CORE RARREL ATTACHMENT Figure 4.2-44. llorizontal IL file l Amend. 57 4.2-529 Nov. 1980
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E L E VA Tl0N 80 0
/
SLOPE = 0.0288 IN/FT / (EXTENSION N0ZZLE) 02R 005 R (SHROUD TUBE)
.009 R (EX TENSION N0ZZLE) j 00 HE AD PE NE TR ATION 2 / 1 l
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@ THE CINTE RiiNE Of THE PHIMARY CROM TOP OF SCR0 BUSHING ISCRS) ' 10f Su$HM IS TO 87 MHW M8 pgg ' 1%CHE S OF THE 14DICAfl0 SiOPE Ai El E VATION 80 JUNCT10N OF UPPE R AND '
LOWE R SHROUD TU8E @ rHt v AtCES USED IN THf SkiTCH ARE I 0 086 R FOR A ROW SEVIN CORNER CONTROL ROD A00ifl0N AL D A T A IS GIVE % 1% THE 0 015 H insti af t0W { si t t Ris H4C 0 396 0 403 0 432 h o us o 3ri 0 388 80T TUM 0F L OWE R 341 65 SHROUD TUBE h -p 350 15 - 0 396+ 0 365- h t ARGE DIMENSION IS ASSEMBt v 10 CORE CDRE ASSEMHL Y E0cr THR0 UGH CORE SMAttDiMEm SiON IS ASSEMBL Y TO C RE EDGE
*- h TO GEi THE MISALIG4 MENT FOR THE PCA H A%DLING $00kEilD AT E LE V 344 15 f A00 0 0081% TO THE VAtutS CIVEN Ai f EttV 350 15 h 0 Af uM A IS THE CINTERLl%E4EOF T:
D ATuM 8 IS THE CE NT E Ril4E Of THE t OWE R SH A000 TUBE A f E t 347 65 f DATUM C 15 THE CENTERLINE OF THE LOWE R SHROU0 ill8E A T f t 215 5. l O ATUM i IS THE CE 4TERLI%E OF THE HE AP STUS %ollLC l 0 002 - - - DATUM F IS THE Cf NTE Rtl%E OF THE uf rER Se*ROUD TuSE AND THE EXTE NSION NO2 ELF AT Et 813
/ O AiuM A,8. C,i AND F A RE PAR AL L E L 10 f C AT UM 0 f
RE CEPTACLE 80T TOM h ALL DIMENSIONS ARE IN INCHES
, UNLESS OTHERWISE NOTED 0238 R Figure 4.2-95 A Control llod System Operating Misalignment Emelope 3975-4a 4.2-591 Amend. 57 Nov. 1980
ELEVATION CRD UPPER Gul0E BUSHING [ ,! .C) SLOPE = 0.229 IN/FT - (EXTENSION N0ZZLE)
- 0.2 R F j (SHROUD TUBE NOTES l .005 R j 00 HE AD PENETRATIONh 1
'009 R b U ^'U" 0 '8 '"' "' # C'0" 8""
3
- g~i 3 k i INSTALLED PLUM 8 CENTERLihE FOR (E X T. N 0 Z) RE E ERE =CE OAiuu o HAS BEEN DR AWN TOP OF UPPER SHROUD ( THROUGH THE APPLICA8LE PE%EIRA TUBE "'""'
(STUB N0ZZLE) .001 { l @ THE vat UES REPRESENT maximum RADIAL MISALIG4MENTS OF COMPONE NT a' (FE ATURE)CENTERLl4ES ACTUAL
; INiil AL MAGeituoE Ann 0RiEnf atiO9 0F t
e SL OPE MISAllGhME %i AT E ACH ELEVATION 1 RE ACf 0R CENTER LINE MAY BE ANYWNRE WpHIN A CIRCLE OF l OF.005 l R ADIUS AS 1401CATE D. TEMPERATURE , IN/F T ASSUME 010 SE 400*F. j l h THE CE Nit RLINE OF THE PCROM GUIDE
; SUSHi%G AT E LEVA 760N 80 SHALL 8E Wif H14 0 04814CHE10F THE 14DICATEs l T0P OF DASHPOT SEAT ; 025 R St OPE .
190 0 _ 0 086 R 215 5
- h THE VALUES USED IN THE SMETCH ARE JUNCit0N OF UPPER FOR A ROWSEVEN CORNER CONTROL l
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/
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$104 5 ASSEM8tv To CORE EOGE.
1 h TO GET THE MISALIG4 MENT FOR THE PCA j HANDLING SOCKET 10 AT ELEV 34415 r. ADD 0 00814.TO THE VALUES GIVEN Af E LE V. -350 T5 h DATUM A15 THE CENTERLINE OF THE PCA ' DISCRIMIN ATOR AT E L 512.15 l DATUM 8 IS THE CENTERLibt OF THE { CORE AS$lMRl y
- e LOWER SHROUO TUBE AT EL-34165.
T HE RMAL SHHf NK AGf DATUM C IS THE CENTE RLINE OF THE f LOWER SHROUD TUBE AT EL-215 5. DATUM E IS THE CE%iERLINE OF THE
~
HE AD STUB 4022LE AT EL 813. D ATUM F IS THE CENTERLINE OF THE l UPPERSHROUD TU8E A%D THE EXTE45404 CORE ASSEMBLY I 4022LE AT EL 813 RECEPTACLE EDTTOM datum A.B.C.E AND F ARE PARALLEL T0 512.15 D AT UM D. j 0.M8 R # h ALL DIMENSIONS ARE IN INCHES UNLESS OTHEPWISE NOTED l Figure 4.2-95B Contiot Rod System Refueling Misalignment Envelope ! D 3975-Sa 4.2-592 Amend. 57 Nov. 1980
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3Hn1VH3dW31 Amend. 53 1764 48 4.2-593
i l V Bypass Flow Figure 4.4-7 shows the bypass orificing which regulates the by-1 pass flow that is used for radial shield and vessel thermal liner cooling. This bypass orificing is located in the 24 peripheral inlet modules. In 18 of the modules, the bypass orificing feeds a controlled portion of the flow downward into the low pressure plenum below the module where it reverses direction and flows upward and is t'en directed radially outward through the core support structure low pressure manifold to the reactor vessel / core barrel annulus. In the other 6 modules, the flow is directed upward to the bypass flow module where it is distributed to those radial shield assemblies (264) outside the inlet module region of the core. Radial Shielding The radial shielding is made up of fixed shielding attached to the core barrel and removable shielding supported by the core support structure as described in Section 4.2. The removable shielding is cooled by external flow through the inter-assembly gaps as well as internal flow through the assemblies. The orifices in the assembly are sized to pro-vide adequate cooling to meet assembly lifetime requirements. External cooling alone was found to be adequate for the fixed shielding. Vessel Cooling Flow The flow from the annulus between the reactor vessel and the core barrel passes upward into an annulus formed by the reactor vessel and the vessel thermal liner. A horizontal baffle is installed between the liner and the core barrel; it minimizes leakage from the liner-barrel annulus into the outlet plenum. From the vessel-liner annulus the coolant discharges into the outlet plenum region above the suppres-sor plate. Two percent of total flow enters the vessel-liner annulus; a fraction of this flow appears as leakage at the outlet nozzle and the 57 makeup nozzle penetrations in the thermal liner. Leakage Seals between the core support structure and core inlet module liner, and other mechanical interfacing locations are sources of leakage. This leakage is estimated to be 1.05% of total flow and is assumed to 51 pass upward through the core interstitial and peripheral region without contributing to the cooling of any reactor component.
4.4-9 Amend. 57 Nov. 1980
Outlet Plenum All fuel, blanket, control, and a portion of the radial shield assembly flow discharges into the upper internals structure. The coolant first enters a mixing chamber before entering the chimneys (Figure 4.4-8). The chimneys duct the flow vertically upward and discharge the flow into the upper region of the vessel outlet plenum. The flow is directed into the upper region of the plenum to minimize flew stratification in this region during a reactor trip transient. The flow from some of the removable radial shields which are located outside of the peripheral skirt of the upper internal structure discharge directly into the outlet plenum. Also, sl43 of total reactor flow from the fuel, blanket, control aad radial shield assemblies bypasses the chimneys through the gap between the top of the core assemblies and the skirt of the upper internals structure and discharges direct'y into the outlet pler. ..n. The coolant leaves the reactor vessel outlet plenum through three 36-inch diameter outlet nozzles. 4.4.2.5 Fuel and Blanket Assemblies Orificing 4.4.2.5.1 Orificina Philosoohv. Approach and Constraints Core orificing, i.e. , flow allocation to the various fuel and blanket assemblies ;s an important step in the core thermal-hydraulic design. Since the assembly temperatures are directly dependent on the amount of flow and since the flcw allocation is the only thermal-hydraulic design parameter which can be varied, within certain limits, by the designer, it logically follows that the core T&H design and performance is only as " good" as the core orificing. Therefore, much attention in the CRBRP core T&H design has been placed on core orificing. Previous experience has indicated that a successful orificing should account 'a priori" for all the various aspects to be considered through the design, in order to avoid time consuming and costly itera-tions when the analyses are well in progress. Thus, a systematic orificing approach was developed, which accounted for lifetime /burnup, transient, upper internals temperature constraints. This new approach represented a change in philosophy and a significant improvement over the previous maximum temperature equalization method. Characteristic features of this approach are determination of the limiting temperatures (see Section 4.4.2.5.2) for all types of assemblies and simultaneous orificing of the fuel and blanket assemblies. Finally, both first and second core conditions were investigated in determining the orificing constraints and the most restrictive in either core was used in deriving the orificing configuration. This guaranteed, a priori, that the thermai-51 hydraulic performance would satisfy the constraints considered in both 4.4-10 Amend. 51 Sept. 1979
4500 a v O 4000 b FUEL Pg 3600 3200 - CONDITIONS:
- ASSEMBLY -901, BOC 3 2800 PLANT EXPECTED OPERATING CONDITIONS 0_ HDT CHANNEL F ACTORS o
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/
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LOWiR UPPER AX1AL BLKT CORE AXtAL BLKT 0 14 50 64 i Figure 4.4-1 Typical Asial Temperature l'rofile in Cittlitl' Fuel Itod 1668-69 Amend. 51 O 4.4-127 Sept. 1979
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* ~ ~
D D F 3 ~\{ oo o 5.J o
l Riser Elastomer Seals
') The balance of the seals on the riser assembly operate at temperatures Il below 1250F.
Upper Internals Structure Jacking Mechanism The UIS jacking mechanism utilizes metal buffered seals in the 400UF areas. These seals are part of the mechanical assemblies. The seals will be removed with components at the appropriate maintenance period. Elastomer 41 l seals are located in the cooler regions, have a service life of five years, 57 and will be replaced using hands-on maintenance, or with special tools. Liquid Level Monitor Ports Plugs c Four of these components, operating at 400 F, are located on the reactor vessel head and provide receptacles for holding the liquid level monitors. Three small port plugs are attached to the top surfaces of the closure head rotating plugs by partial penetration welds, two on the intermediate and one on the large rotating plug. One large port plug is bolted to the top surface of the large rotating plug and is sealed to the plug by double metal "0" rings. The seals remain attached to the pirt plug during installa-tion and removal. An inerged cask V:ill be used to install and remove the liquid 25 level monitor while at 400 F, requiring no hands-on operation. Because the port plug remains stationary relative to the head assembly, the metal "0" rings 57 beneath the plug flange are not expected to require maintenance. p) ( 5.2.1.4 Guard Vessel The guard vessel provides for the retention of the primary sodium coolant in the event of a leak in the portion of the primary coolant boundary which it surrounds. The guard vessel geometry assures reactor vessel outlet nozzle submergence after such a leak which will maintain continuity in operating pri...ary coolant loops to provide core cooling. The guard vessel also provides a uniform annulus for in-service inspection
- the reactor vessel, with clearances that preclude contact with the reacar vessel and piping under accident conditions. Insulation for the reactor vessel and a heating system for the reactor vessel to be used prior to sodium fill and during prolonged shutdown are also mounted upon the guard vessel.
The maximum and minimum widths of the radial gap between the guard vessel and the reactor vessel have been conservatively calculated, taking into account all relevent factors such as tolerances on the diameters of the two vessels, permissible out-of-roundness of the two vessels, possible deviations from straightness due to manufacture and subsequent operation, thermal expan-sion, initial deviations in the alignment of the two vessels, etc. The trans-porter for the television camera will be designed to accommodate itself to this maximum possible range of gaps as it moves in the space between the two vessels. 25 in v
)
Amend. 57 5.2-4a Nov. 1980
5.2.l.5 Reactor Vessel Preheat The Reactor Vessel Preheat System will control the dry heat-up and cool down of the Guard Vessel, Reactor Vessel and Internals between ambient (70 F) and 400 F and if required will provide make-up heat for that lost to the Reactor Cavity during prolonged shutdowns. The heat will be provided by tubular electrical heaters mounted between the Guard Vessel and insulation. These heaters will be arranged circumferentially around the Guard Vessel and will be grouped and controlled in zones of uniform heat output. Temperature sensing devices will monitor the Guard Vessel temperature in each of these zones and provide the necessary feedback for power level adjustments in the heaters. The heaters will be mounted to the same framework which supports the Guard Vessel insulation. Ceramic offsets will be used to offset the framework and heaters from the Guard Vessel surface. The heaters and framework will therefore be electrically isolated from the Guard Vessel. Convective barriers, reflective sheaths and the Guard Vessel insulation will be used to optimize heat input to the Guard Vessel and minimize losses to the Reactor Cavity. Preliminary preheat, startup, and shutdown analyses have been performed on the Reactor Versel and Guard Vessel to determine the temperature differences which will result in opening and/or closure of the annular gap between the two vessels. By necessity the preheat analysis is very preliminary since no firm preheat procedure has yet been developed. Figures 5.2-4 through 5.2-6 show the temperature differences between the Reactor Vessel and Guard Vessel in the inlet and outlet plenum regions for the three transients in question. As shown the largest positive temperature difference between the Reactor Vessel and the Guard Vessel occurs in the outlet plenum region during startup (335 F) while the largest negative temperature difference occurs in the outlet plenum region during shutdown (-214 F). The nominal radial gap between the reactor vessel and guard vessel is 8 inches at assembly and at the end of preheat. This gap decreases to approximately 7.6 inches minimum during start-up and increases to approximately 8.3 inches maximum during shutdown. During preheat the gap also increases but to a lesser value than during shutdown due to the smaller maximum temperature difference. Variations in the axial gap between the bottom of the reactor vessel and the inner surface of the guard vessel are noted between the states shown in the trbie. Thus the largest axial gap is 11.0 inches at the dry cold condition and the smallest gap is 6.2 inches at the end of the heating phase of preheat. g 5.2.2 Design Parameters Overall schematic views of the reactor vessel, closure heeJ assembly, inlet and outlet piping, and guard vessel are shown in 56 Figures 5.2-1, l A and 13. The top view is given in Fiqure 5.2-2. l 5.2-4b Amend. 56 Aug. 1930
a l 41l 17l load transmitted to the support ledge is reduced due to the large mass presented by the spring-coupled closure head / reactor vessel sy: tem compared j O to the mass of the postulated sodium slug. 36 5.2.2.2 Closure Head The closure head consists of three rotating plugs which will be constructed of SA 508 Class 2 steel. Each plug contains a major penetration eccentric to its outside diameter. These rotating plugs are interconnected 4 17 l by means of a series of plug risers. Sealing between the plugs is accompluned by sodium dip seals and double inflatable seals of elastomer material. At 36 its top, the large rotating plug has an outer diameter of 257.38 in., and an i inner diameter of 176.50 in. The large rotating plug provides access to the vessel interior liquid level for the ex-vessel transfer machine and the core coolant monitors. The intermediate rotating plug (175.50 in. 0.D._and 68.94 in. I.D.) provides access to the vessel interior for the control rod drivelines, upper intervals support columns, and the liquid level monitors. 41l The small rotating plug (67.94 in. 0.D.) provides access to the vessel for the In-Vessel Transfer Machine. The thickness of each rotating plug is , 41 l 22.0 in. Rotation of the plugs will be accomplished by a gearing and bearing system attached to the plug risers. The nozzles for each penetration will 41 l be constructed of an austenitic stainless steel. 4 Each rotating plug is provided with a system of mechanical locks and electrical interlocks which prevent plug rotation during reactor operation and refueling when plug rotation is not desired. 4 l The echanical locks include the following:
- a. Each plug includes a separate positivc lock to assure t1at the plug cannot be moved, and will not drift from its normal operating position during reactor operation.
i This lock w'll be installed to prevent relative rotation between eact bull gear and its outer riser whenever the control rod irivelines are connected. The locks shall be manually installed at the end of each refueling cycle, < and will be removed only during the refueling period when i plug rotation is necessary.
- b. The plug drives are designed to be self locking to react to any seismic torque occurring during refueling, which could rotate the plugs and thus damage a fuel or blanket assembly during removal from the core.
The electrical interlocks include the following:
- a. During reactor operation, the plug drive and control system keyswitch is in the OFF position, the control system is deenergized, and there is no power to the g plug drive motors.
5.2-6 Nov.1977
~ ,_.._ _ _ __ - _ _ . . _ - _ _ _ - _ _ , _ , _ , _ _ _ - _ _ _ _ - . . - _ , _ . . - .
- b. Electrical interlocks are provided to prevent the plugs from being inadvertently rotated by their drive system unless the upper internals are raised and locked, and the IVTM and EVTM are in a safe condition.
- c. An electrical interlock is also provided to prevent vertical operation 56 24 af the IVTM or operation of the EVTil over the HAA during operation of the plug drive system.
Each rotating plug has attendant thermal and radiological shiesding extended to a depth of 74.65 in. beneath the top of each plu9 forging. The shielding is composed of a series of plates fabricated from carbon steel, 17land stainless steel. The cover gas between each set of plates attenuates thermal conduction and thereby acts to decrease the heat flux imparted to the rotating plug. A heating and cooling system is provided to maintain the closure head at 4000 F (nominal) as well as providing neating and cool-17 ing for other small head mounted subassemblies. A gas entrainment suppressor plate assembly is positioned beneath 45l17lthe head thermal and radiological shielding at a depth of 122.65 in, beneath the top of each rotating plug. It protects the head shielding from being contacted by the core coolant and minimizes the amount of cover gas entrain-ed in the core coolant. The assembly is designed to accommodate all normal, 17l upset, emergency, and faulted conditions. O 56 25 In plan view, the subassembly consists of 33 plates at the same 42le levation with horizontal gaps between them. (Fig. 5.2-3) These plates have penetrations in line with the head penetrations to allow the passage of the head mounted c:mponents into the outlet plenum. Each plate ie supported by means of a cent-al support column affixed to the lower shield plate. These central columns, when possible, consist of tubes which surround closure head penetrations. Support columns which do not surround penetrating equipment will be capped to minimize the amount of cover gas entrained. The support columns will be inserted through oversized penetrations in the lower shield plate, accurately positioned and then attached to the top surface of the lower plate by means of bolting. The support columns will be attached to the suppressor plate by means of welding. This attachment wi d is located above the region of the suppressor plate where high thermal gradients occur by using a plate with an extruded weld neck. The top end of the support column, which protrudes through the lower shield plate is composed of 2h Cr-1Mo. material to minimize the differential expansion with the carbon steel shield plate. The lower, in sodium, portion is austenitic stainless 57 42 steel. The use of a single support provides adequate support while lessening 2 the thermal stresses by permitting the plates to flex freely under the expected thermal gradient, Amend. 57 O 5.2-6a tiov. 1980
The riser has been designed to maintain a maximum temperature of ex 125U F in the region of the elastomer seals. Thermal analysis hag been ( ) completed for this design which shows that this temperature (125 F) is maintained by natural circulation cooling of the r ser structure. 25 i 5.2.2.3 Guard Vessel ! The guard vessel is a bottom-supported, right circular cylindrical vessel surrounding the reactor vessel. It will be fabricated from SA240 Type 304 stainless steel. The purpose of the guard vessel is to assure outlet nozzle submergence in the event of a leak in the e 2 j s, t ! v (- ()\ 5.2-6b Amend. 57 Nov. 1980
Figure 5.2-8 gives enlarged views of the elastomer seal areas j to show the sealing arrangement. The upper view shows typical seals for the top end of the riser. The inflatable dynamics seals are mounted into a removable seal retainer ring, which is sealed to the riser top by elastomer o-ring seals. The space enclosed by the two inflatable 20 seals is a static buffer. The CRBRP inflatable seals are the same in cross-section and made of the same material (nylon-reinforced nitrile rubber) as the FFTF-IVHM inflatable seais. Extensive testing was conducted to qualify the IVHM inflatable seal at an operating temperature of 150 F 0 (Reference 41l 5.2-6). Additional performance testing has been conducted (Reference 5.2-7) on an FFTF-IVTM type inflatable seal to assure these seals will function reliably under the CRBRP speed and pressure operating conditions. This testing was also conducted at 1500F. Specific objectives of the CRBRP program included buffer cavity leakage and seal drag measurements under combinations of seal and buffer cavity pressure, measurement of gas diffusion through the seal, breakaway seal drag as a function of dwell times, effect of lubrication and long duration (life) cycling on the wear characteristics of the seal, and the effect of horizontal and vertical runout of the seal runner on seal performance. Testing to support the present design adequacy under dynamic operation is nearing completion. The verification of elastoner life at 1250F is documented in Reference 5.2-8. The life of the elastomers for upward of 5 years in this temperat.ure and environment has been domonstrated 25 O by tests. V The seal retainer ring is sealed to the inner riser by two solid elastomer o-ring seals. These have an inerted, static buffer space between them which acts similar to the one described above. The seal follower is sealed to the outer riser in the same way , but with the space between seals 20 purged with a small argon flow instead of static buffering. 49
~
The primary source of experimental data for CRBRP seal design is the on-going Cover Gas Seal Development Program being performed by Atomics International (AI). Testing includes static, rotary, and reciprocating seal leakage tests, compression set tests, gas permeability measurements, and seal lubricant evaluations including elastomer compatibility, thermal stability and friction testing. Elastomers tested include silicone, ethylene-propylene, urethane, nitrile (Buna N) and butyl rubber supplied by three different seal verdors. Test temperatures ranged from 100 to 3000F depending upon the elastomer being tested. In addition to the numerous quarterly reports which have been published, two sunmary reports 41l have been issued (References 5.2-8 and 5.2-9). The information contained ) in these report amply demonstrate the ability of several elastomers to ' meet CRBRP design requirements at 1250F. These tests demonstrate that 25 elastomer life of upward of 5 years in this temperature and environment is achievable. O \> Anend. 49 5.2-9b Arril 1979 l l
O The closure head has several penetrations that permit access to the reactor interior. Penetrations will be sealed in one of two ways. One is to provide two seals with a pressurized buffer space between them. In the event of a failure of the inner seal, that between the reactor interior and the buffer space, buffer gas will laak into the vessel. In the event of a failure in the outer seal, buffer gas will leak into the head access area. j In each ceae, the leak will be detected and repaired before a leak can occur between the reactor interior and the head access area. The other method is to use a hermetic seal. Figure 5.2-8 shows the method of sealing the riser bases to the vessel head. Soft coated metallic c-rings wi th an inerted, purged space between them are used in the same way as the elastomers at the 20 49 riser top. A continuous metal "C" ring or canopy seal is welded at both ends completely around the periphery to provide a hermetic seal at the base of the large outer riser. 36 The base of the small and intermediate outer risers are welded directly to the closure head and therefore have no ieak 57 path at the juncture. General approach to eal selection is:
- 1. For openings which are at high ternperature and/or require long life, metallic seals, hermetic or double buffered, are used.
- 2. Seals which operate at low temperature and can be replaced relatively frequently are elastomer sealed.
1 , t Amend. 57 5.2-9c Nov. 1980
O inspection of materials for the primary pressure coundary (i:eactor Vessel and Closure) will be in accordance with the P.bi miterial standards for the particular materials. Inspection during fabrication will be in cccor-dance with Section III of the ASME Code and RDT ElS-2 f,C-T, Class I r!uclear
.;1l 27 l C omonen ts . Inspection of materials and insrection during fabrication of the Guard Vessel will be in accordance with ASME Code, Class ! and RDT ElS-41l 271 2NB-T. The overall inspection and test plans for the three structures will be prepared by the fabricator and approved by the purchaser prior to fabri-17 cation.
5.2.7 Eacking, Packaging & Storage Applicable requirements to assure adequate quality during shipping 43 and storege will be in the respective equipment specifications rather than ir. RDT 5tandards. Tne specifications will require that packaging and packing te acequatc to protect items while at the suppliers' facilities, during tran:,- 31 portaticn to the aelivery paint and during storage at the site. ine specifications will, where appropriate, provide reacirements fc,r sealing tne openings in the components, purging the components ar.d/or their tainers, selecting and using desiccants, selecting and using noterials 9 3t centacting the components which are suitably f ree of chlorides, flourices, lead, copper, zinc , cadmium, sul fur, reercury, etc. During storage, the equipment will be maintained in a dry gar environ-57 31 m nt, where appropriate, to protect it from contamination. The purge gas, container inteority, etc. , will be monitored to assure compliance with previously prepared procedures. 57l Protective measures to be taken during interim storage and con-struction will be provided by the constructior contractor. 25 1 O 5.2-lla / mend. 57 Nov. 1930
i REFERENCES FCd SECTION 5.2 I
- 1. G.F. Carpenter, N.R. Kuopf, E.S. Byron, "Anomolous Embrittling Effects Observed during Irradiation Studies on Reactor Vessel Steels", Nuclear Science and Enoineerina, Vol 19,1964, pp 18-38. l l
l 2. "Inconel Alloy EC0", Huntington Alloy Products Bulletin T-7, I 1969.
- 3. " Steels for Elevated Temperature Service", U.S. Steel, June 1976, pp. 70, 71.
- 4. S. Schrock, S. Shiels, C. Bagnall " Carbon - Nitrogen Transportation in Sodium Systems", Summaries of kechnical Papers, CONF-76-05-3-SUM
- 5. " Heat Treating, Cleaning and Forming",.etals Handbook, Vol. 2, ASME .
196", pp. 149-166. l 22 Atonics International Report No. TR-707-310-004, " Test Report (Deselocment) IVHi' Reactor Refueling Plug Dynamic Seal Test",
!'ay 8, 1974.
i A ctics International Report No. TR-707-810-009, " Test Report I Jevelopr.ent) CRBRP Rotating Plug Inflatable Seal", July 9,1975. tomics Ir.ternational Report No. Al-AEC-13145, "Desian Guide for i Reas tor Cover Gas Elastomeric Seals", March 7,1975.
- 9. Atc ics In ternatianal Report No. Al-AEC-13146. "Penetra tion, Lonage, w,d Ct.'ruression Set Testina of Elastreric Seals fo*
LFP' Use", Acri! 2, 1975. 25 l l l O' i 5.2-11b Amend. 25 Aug. 1976 i
... - -. - . - ~ . _ - - . - . - . - . _ - - - - - _ - ._. _-_- -_ . - . - .._... --- -- -.
l O O O TABLE 5.2-1 i
SUMMARY
OF CODE, CODE CASES AND RDT < j STANDARDS APPLICABLE TO DESIGN AND MANUFACTURE OF REACTOR VESSEL, CLOSURE HEAD AND GUARD VESSEL Closure Head
- Pressure Internals Guard Component / Criteria Reactor Vessel
- Boundary (asappropriate) Vessel 57 -
, Section III Addenda thru Winter Addenda thru Winter Addenda thru Winter Addenda thru ASME Code, '74 '74 '74 Summer '75 1974 Edition Class 1 Class 1 Class 1 Class 2* ASME Code Cases 1521-1,1592-2,1593- 1682,1690 1521-1 1592,1593,1594 0,1594-1,1595-1, 1592-4,1593-1 if elected by sup-1596-1,1682,1690 plier 1521-1 & cn . 1682 n3 RDT Standards E8-18T, 2/75 E15-2NB-T, 11/74 E15-2NB-T, 11/74 E15-NR-T, 11/74
. 57l Manda tory E15-2NB-T, 11/74 Amend thru 6/75 Amend thru 6/75 Amend thru 6/76 n'
Amend thru 1/75 i F2-2, 8/73 F2-2, 8/73 F2-2, 8/73 F2-2, 8/73 Amend thru 3/74 Amend thru 7/75 Amend thru 7/75 Amend thru 7/75 F3-6T, 12/ 74 F3-6T, 12/74** F9-4, 9/74 F3-6T, 10/75 l F6-5T, 8/74 F6-5T, 8/74 F6-5T, 8/74 Amend thru 2/75 Amend thru 2/75 Amend thru 11/75 F7-3T, 11/74 F7-3T, 6/75 F7-3T, 6/75 F9-4T, 9/74 M1-1T, 2/75 F9-4, 9/74 1 gy gp 42 M1-2, 3/75 5{ Amend thru 7/75 , g- - ' *For those reactor vessel and closure head components internal to the pressure boundary special purpose high cycle fatigue curves and creep damage rules have been developed as discussed in 57 Appendix 5.2A. 1 i
TABLE 5.2-1 (Continued) Closure Head Pressure Internals Guard Component /Cri teria Reactor Vessel Boundary (as appropriate) Vessel 57 l l RDT Standards 'M1-IT, 3/75 M1-2T, 4/75 M1-4T, 3/75 M1-4T, 3/75 M1-6T, 4/75 Amend 1-7/75 4 M1-6T, 4/75 M1-10T, 3/75 M1-10T, 3/75 M1-11T, 3/75 Amend.1-7/75 f11-11T, 3/75 M1-17T, 3/75 t11-17T, 3/75 M2-2T, 12/74 f12-2T, 12/74 M2-7T, 3/75** M2-ST, 1/75 M3-10T, 7/75 Amend 1-2/75 m M2-7T, 2/75 M7-4T, 3/75 h M2-18T, 4/76 b 57 ;M2-21T, 12/77 i M3-6T, 3/75
' M3-7T, 4/75 lMS-1T, 11/74 MS-2T, 5/73 M5-3T, 12/74 MS-4T, 1/75 kN 57l M6-3T, 2/75 b .E M6-4T, 2/75 e
gm M7-3T, 11/74 42 tion-Mandatory F9-ST, 9/74 , F9-5T, 9/74 ,F9-5T, 9/74
- Functionally designated Class 2, and constructed to rules for Class 1, but not hydrostatically tested er code stamped.
** Except for the three rotating plugs, for which the applicable issues are: F3-6T, 3/69 for LRP &
42 SRP; F3-6T, 5/74 for IRP. M2-7T, 2/69 for LRP & SRP; F2-7T, 2/74 for IRP. O O O
p ,a p O 0 NA TABLE 5.2-3 MATERIALS FROM WHICH THE REACTOR VESSL CLOSURE HEAD AND GUARD VESSEL ARE FABRIL ED Reactor Vessel Product Form Material Comment
,upport Ring Ring Forging SA 508 Class 2 Vessel Flange Ring Forging SA 508 Class 2 1 Transition Shell Plate SB 168 Inconel 600 17l Shell Cources Pla te SA 240, Type 304 Austenitic stainless steel Core Support Ring Forging SA 182, Type F304 Austenitic stainless steel Core Support Cone Pla te SA 240, Type 304 f rmed into arcs and welded Inlet Plenum Plate SA 240, Type 304 Therml Liner Plate SA 240, Type 316 Austenitic stainless steel Forging SA 182, Type F304 Austenitic stainless, formed into 1 l Thernal Liner Support Ring segments and welded
'lozzl es Forging SA 182, Type F304
,m Closure Head to k Rotating Plugs Penetration Nozzles Forging Forgino SA-508, Class 2 SA-182, Type F304 Austenitic stainless steel Shield Plates Plate SA-516, Grade 60 Shielding Support Skirts Plate SA-516, Grade 60 Plate formed into arcs and welded Reflector Plates Plate SA-240, Type 304 Austenitic stainless steel Reflector Plate Supports Forging or SA-182, Type F304 or Pipe SA-312, Type 304 Austenitic stainless steel Suppressor Plates Plate SA-240, Type 316H Austenitic stainless steel Suppressor Plate Support Forging SA-182, Type F316H Austenitic stainless steel Column and SA-336, Gr F22 Lower portion welded to 24 C.
- 1 Mo. upper portion Spacer Bars Bar SA-387, Class 2, Grt b 22 Margin Ring Bar SA-Y0, Class 1, Grade B-24 gg Margin Ring Keeper Bar SA-533, Class 1
< a> Suppressor Plate Column Plate or SA-387, Grade 22,
'E 57 Caps Forging Class 2 or SA-336, y Grade F22 os
TABLE 5.2-3 (Cont'd.) 1l Guard Vessel Product Fonn Material Comment Vessel top flange Bar, Pla te, SA 479, SA 240, Type 304 Forging SA 182 Vessel Plate SA 240 Type 304 Vessel to support skirt ring Bar, Forging SA 479, SA 182 Type 304 Support Skirt Plate SA 240 Type 304 Support Flange Plate SA 240 Tyne 304 Nozzles Plate, Forging SA 240, SA 182 Type 304 Guard Pipe Flanges Bar, Plate SA 479, SA 240 Type 304 Guard Pipe Welded Pipe, Plate SA 409, SA 2f,0 Type 304 Guard Pipe Elbows Welded Fitting, SA 403, SA 240 Type WP 304 g Plate P Cleanout Nozzle Forging SA 182 Type F 304 ';o Cleanout Nozzle Cap Forging, Plate SA 182, SA 240 Type F 304 y 41 Ef sr
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O BULL GEAR AND INNER RISER STUDS IC BEARING INNEP RACE l BEARING OUTER RACE MARGIN SEALS y s
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- OR DUTER PLUG TYPICAL RISER BASE SEALS Figure 5.2-8 Riser Sealing Details 5.2-22 Amend. 41 Oct. 1977 1
I
5.3 PRIMARY HEAT TRANSPORT SYSTEM (PHTS) Q) 5.3.1 Design Bases 5.3.1.1 Performance Requirements 17 The Primary Heat Transport System extracts reactor generated heat and delivers this heat to the intermediate coolant system under all normal and off-normal operating conditions. Specific performance requirements include: 17 Heat Transport and Flow Performance
- a. Transport of reactor generated heat (975 MWt) through the primary to intermediate coolant system while maintaining an adequate flow rate for controlling reactor temperature cor.aitions within limits which preclude damage to the reactor vessel, fuel and reactor internals.
- b. Regulation of heat transport system flow in response to plant process control over the full operating power range of 40 to 100 percent reactor thermal power.
- c. Transfer of decay heat to the intermediate coolant system by pony motor operation or natural circulation under all normal and off-normal conditions including failure of a heat transport system component or loop. Specifically, there will be capability to remove decay heat by natural circulation with two or three loops
[] following operation' on two loops at the appropriate power.
'v 17
- a. Containmrnt of primary sodium coolant and radioactive fission products within the primary coolant system by providing a bound-ary for primary coolant confinement and a separation of the pri-mary and intermediate coolant systems all within the confines of the containment building.
- e. Transport of reactor generated heat to the intennediate coolant system with two-loop operation at the appropriate power input.
17
- f. Provide a sodium coolar.t system which can be easily filled, vented and drained.
- g. Support of operation in a hot stand-by condition - nominally 71/2 to 10% of full flow at a normal temperature of 600 F.
Structural Performance
- a. Design, fabrication, erection and testing of the PHTS components which comprise the sodium boundary shall be in accordance with the ASME Boiler and Pressure Vessel Code, Section III, Division 1, 5 fluclear Power Plant Components,1974 Edition. The addenda to b)7 V
5.3-1 Amend. 57 flov. 1980
Section III that are apolicable for the various components are given in Section 5.3.1.2. The applicable ASME Code Cases (1592,1593,1594,1595,1596) for elevated temperature compo-aents shall also be employed. In addition, the RDT Standards E15-2NB-T and F9-4T with revisions as given in Section 5.3.1.2 shall apply, along with the RDT Standards F2-2, F3-6T and F6-5T 27 57 of appropriate revisions coincident with the compor. ant contract date of concern.
- b. The natural frequencies of all components will, where possible, avoid resonance with all expected pump driving frequencies.
Where not possible, the component design shall insure that struc-tural damage will not occur as a result of resonance.
- c. Structural design shall provide for dry HTS piping and component heat up at a rate of 30F/hr.
41
- d. Structural design shall provide for a system fill under condi-tions of full vacuum with system components at an average tem-perature of 400 F and hot spot temperatures of 600 F.
Transients
- a. The primary heat transport system shall be designed to accommo-date the thermal transients resulting from the normal, upset, emergency and faulted conditions described in Appendix B of this document.
- b. The system shall be designed such that a normal or upset event does not adverself affect the useful life of any HTS component.
- c. Following an emergency condition, resumption of operation must be possible following repair and re-inspection of the components, except that the primary ccolant pumps (damaged or undamaged) must maintain capability to provide pony motor flow following all emergency conditions except in the affected loop for pump mechan-ical failure.
- d. During and following a faulted condition, the heat transport system j must remain sufficiently intact to be capable of performing its 33 decay heat removal function, including maintenance of primary coolant pump pony motor flow.
- e. The primary heat transport system will accommodate, without loss of decay heat removal capability, the pressures imposed on the intermediate system, by a major sodium water reaction and the sodium hammer resulting from a primary loop check valve closure caused by the most severe flow degradation, such as a primary pump seizure.
Amend. 57 Nov. 1980 E.3-2 l l
- f. The primary heat transport system will acconinodate, without
/] loss of integrity of the primary coolant boundary, the 'Q pressure transients calculated to be the result of a hypo- 33 4 thetical accident releasing energy in the reactor vessel.
These " Third Level Design loadings" for various parts of the' system are given in Section 15.1.1.3. Seismic Loads i The PhTS components and supports (ASME Class 1) under the jurisdiction of the ASME Code, Section III, shall be designed to accommodate the load combinations (including seismic loads) prescribed herein without producing total combined stresses, total strains (for all loading cycles)
' and cumulative creep / fatigue damage (for all loading cycles) in excess of those allowed by the Code. No component of stress or strain of an individual loading condition shall be included which would render the cont)ination non-conservative. Transient loadings shall be included as requimd by the Code. For elevated temperatures, Code Case 1592 27 I 57l supplemented by RDT Standard F9-4T will apply.
The ASME Code Class 1 PHTS seismic Category I components shall be designed to withstand the effects of the SSE and of the OBE and remain functional. The OBE will be considered as an upset condition and the SSE 1 5 as a faulted condition. For the lov-ter<perature portfor of the desicn analysis, the rules in the ASME-III code proper will be followed. Accordingly, the OBE shall be included in the analysis to the Code
" Design Condition" requirements and limits. For the elevated-temperature portion of the design analysis, the requirements set forth in Code Case 1592 will be followed. Therefore, the OBE shall be included in the analysis to the Code " Upset Condition" requirements and limits.
The reason for this difference is that the rules of the Code proper do not require an assessment of Primary Membrane stress plus Bending except in the " Design Condition" case. Complete details for loading combinations of normal and transient condition loadings with seismic loads are provided in Section 3.9.1.5. Thermal and Hydraulic design basis parameters are given in Tabl e 5.3-1. Component structural design pressures and temperatures are given in Table 5.3-2. Pressures and temperatures used for structural evaluation at steady state operating conditions are given in Table 5.3-3. 4 35 1 5.3-3 Amend. 57 Nov. 1980 4
' - ~
_ -- . r y n--
5.3.1.2 Applicable Code Criteria and Cases The PHTS pressure containing components shall be designed, fabricateo, erected, constructed, tested and inspected (in compliance with 10CFR50, Section 50.55a) to the standards listed below: COMP 0NENT APPLICABLE STANDARD AND CLASS PHTS Pump ASME III, Class 1 PHTS IHX ASME III, Class 1 PHTS Check Valve ASME III, Class 1 PHTS Piping (including ASME III, Class 1 flowmeter, thermowells and pressure tap pene-trations,etc.) All primary heat transport system components shall be analyzed as Cla 1 nuclear components in accordance with the rules indicated by the documents in the following: 1 2 ASME Code RDT-E15-2NB-T 3 4 EDITION & EDITION & RDT F9-4T CODE CASE 1592 COMPONENT ADDENDA AMENDMENTS EDITION REVISION PHTS IHX 1974 plus November 1974 September 1592-1 Summer 1974 Amendment 1 1974 Addenda Primary Pump 1974 plus November 1974 September 1592-4 Guard Vessel Sumner 1974 Amendments 1974 Addenda 1,2,3 IHX Guard 1974 plus November 1974 September 1592-4 Vessel Summer 1974 Amendments 1974 Addenda 1,2,3 PHTS Piping 1974 plus November 1974 January 1592-7 Addenda to Amendments 1976 Summer 1975 1, 2, 3 PHTS Check 1974 plus November 1974 September 1592-2 Valve Addenda to Amendments 1974 Summer 1975 1, 2 NOTES: 1) ASME Boiler and Pressure Vessel Code, Section III
- 2) RDT E15-2NB-T (Supplement to Section III)
- 3) RDT F9-4T (Supplement to Code Case 1592)
- 4) ASME Code Case 1592, " Class 1 Nuclear Components 57 Amend. 57 Nov. 1980 5.3-3a
b V 27 57 The " Nuclear Systems Materials Handbook" (Ref.11 shall be used to obtain material properties data not available from the above sources. As required in RDT F9-4T, the use of additional or alternative material properties shall require the approval of the purchaser. Code Case 1521, "Use of H-Grades of SA-240, SA-479, SA-336, and SA-358, Sec-tion III," may be used for H-Grades of Type 304 and 316 austenitic stainless steels. RDT F9-ST, Sept.1974 (Section 6) provides alternative procedures for satisfying the strain limits of Appendix T of Code Case 1592 which are acceptable 27 l to the purchaser. Section 6 of RDT F9-5T, Sept.1974 also provides time / temper-36 atures below which the primary plus secondary and peak stress limits of Section III may be used in place of the limits of Appendix T of Code Case 1592. The scope of the analysis of Code Case 1592 shall be used even if tne limits from Section III are used. For example, the primary plus secondary stress intensity range due to emergency as well as normal plus upset conditions is limited. In addition, Code Cases 1593, for fabrication and installation of elevated temperature components,1594 for their examination,1S95 for their testing, and 1596 for their overpressure protection shall apply for the pri-mary heat transport system components. p 5.3.1.3 Surveillance Requirements V~ Changes in fracture toughness of the PHTS piping and components may be caused by carburization, plastic creep straining and the thermal environ-ment though not radiation effects. The need for surveillance of stainless steel piping and components for changes in fracture toughness will be determined by ongoing programs. If a requirement is identified by ongoing programs, a surveillance program will be designed in accordance with the philosophy of 10CFR50, Appendix H. 5.3.1.4 Materials Considerations 5.3.1.4.1 Basis for High Temperature Design / Analysis Rules governing the construction of Class 1 components which are to experience temperatures above those now provided in Section III shall be constructed in accordance with the following considerations:
- a. The rules for materials in NB-2000 shall apply except as modi-fied by Code Case 1592, and fl
'/ 5.3-4 Amend. 57 Nov. 1980
The guard vessels for the IHX and primary pu.",p components shall
- 44) be fabricated from Type 304 Austenitic Stairless Steel.
\ The material specifications for the construction of the primary heat transport system components are given in Tables 5.3-4 thru 5.3-8. The supplier may substitute other ASME-approved material than that specified in Table 5.3-4 thru 8 af ter review and approval by purchaser prior to use or procurement. If the supplier elects to use other ASME-approved materials, the supplier's request for approva' shall contain topical reports demonstrating the adequacy of the selected alternate material. Selection of alternate materials shall be based on the mechanical properties, metallurgical stability, sodium compatibility, and response to radiation under the applicable design and environmental conditions. When recommending the use of alternate materials, the supplier shall document the justification which shall include, as a minimum, a summary of available test or experience data and a discussion of the adequacy of the recomcended materials relative to 304 or 316 stainless steel or other purchaser-specified alternate. 5.3.1.4.3 Additional Requirements The following requirements are modifications or additions to the reqbirements of the materials specifications identified in Tables 5.3 (3 thru 5.3-8. Lj Hydrostatic Test of PHTS Piping The pHTS piping will be hydrostatically tested in accordance with the ASME Code, Section III, Article NB6000 and Code Case 159' (supplemented by RDT E15-2NBT, October 1975), as described below. The lengths of straight pipe will be tested by the manufacturer at his facility to satisfy the Code material specifications. The thermowell body sub-assemblies will be tested under external pressure, by the manufacturer. The IHX Vent-line flow restrictor will be tested by its manufacturer. All other items are con-sidered by the Code to be materials and will not be tested separately. Rather, each completed PHTS piping sub-assembly (spool) wi:1 be hydrosta-tically tested by the spool fabricator at his facility prior to their being shipped to the plant site. Thus, all piping items such as fittings and branch connections will be tested, as well as the spool welds, prior to installation. Specific procedures have not yet been written, but they will be in compliance with the Code requirements noted above as they 57 apply to the design conditions at each location in the system. Strength Tests of HTS Components Hydrostatic or pneumatic strength tests shall be conducted in accordance with the ASME Code Section III Code Case 1595-1 and implement 27] any requirements of RDT Std E15-2NB-T, Oct.1975. 16 Amend. 57 5"3-7 Nov. 1980
Water shall be used as the test medium for hydrostatic tests unless it cannot be shown that residual water can be completely removed. If complete water removal cannot be accomplished, a substitute liquid may be used. For example, the cold leg check valve hydrostatic test.will be. performed using liquid Freon as the test medium. In any case where a suitable liquid cannot be used, a pneumatic strength test will be employed in accordance with the Code. Guard vessels are not pressure tested. In addition to hydrostatic and/or pneumatic strength tests, helium leak tests of certain liquid metal containing parts or assemblies will be required. Tarticularly, tube to tube sheet welds, tube bundles, single pass welds and thin sections (e.g., bellows, rupture discs and seal welds) will be helium leak tested in accordance with the ASME Code, Section V, Article 10. Helium leak tests will be perfonned af ter a pneumatic test so that any porosity or minute defect will be exaggerated thereby increasing the sensitivity of the helium leak test. If a component to be hydro-statically tested will also be helium leak tested, the helium leak test will be performed first to preclude water molecules from plugging minute leak paths which the belium leak test otherwise would detect. Gas Pressure Test - Leak Test Prior to sodium fill of the reactor and the PHTS, a system pneumatic test wil1 be performed in accordance with 6320 of ASME Code Case 1595-1. This Code Case requires that the test pressure shall not exceed the lowest of the maximum test pressuras allowed for any of the components in the system which in the case of the CP.BRP would be 18 psig (1,2 times the 15 psig design pressure of the reactor vessel upper plenum). The principal purpose of the test is to leak check field welds. Individual components and piping spool pieces will have been pressure tested prior to installation in ac-cordance with individual specification requirements as noted above. This test will be conducted to supplement visual penetrant and radiographic examination of field welds by checking for gas leaks. A trace gas (helium) may be added to the test ri!ediun or a bubble test may be performed. These tests may be supplemented by a gross leak rate test. The various options that may be used in preparing a specific test procedure have not been evaluated, but will be included in the FSAR. I 16 Chemical Analysis The materials specified in Tables 5.3-4 thru 5.3-8 for the primary heat transport system components shall conform to the chemical compositions specified by the applicable RDT Standard and the additional requirements noted. 5.3-7a Amend. 16 Apr. 1976
27 Shell Assembly The shell, which is the main IHX enclosure, is fabricated from Types 304 and 316 stainless steel. The shell is welded directly to 27 the lower edge of the cylindrical hanging support through the "Z" junction. The shell cyhnaer is . fabricated from three cylindrical sections. The use
~
of Type 316 ' stainless steel reduces the' thermal and creep ratcheting problems, thus 27 providing a shell design without the need for a thermal liner in the high temper-ature region. The bottom portion of the shell consists of a shell cylinder, lateral support ring, hemi-head and primary outlet nozzle, all of Type 304 l27 stainless steel. The lateral support ring has sixteen spacer guides attached to it which serve as guides and restraints for the tuoe bundle lower tubesheet. To preclude potential cover gas accumulation in the IHX, there is continuous venting sodium from the top of the IHX shell to the primary pump tank below the minimum safe sodium level. This sodium flow will carry any gas evolved in the IHX to the pump tank where the gas will migrate to the cover gas space. Tube Bund _le Assembly The tube bundle assembly consi'ts of two major sub-assemblies: (1) the bundle, consisting of tubesha. s. tubes, support plates, tierods and spacers, outer shroud, hemi-head, downce wr, strongback and by-pass seal, and (2) the channel assembly consisting of replaceable bellows, upper head, intennediate outlet nozzle, intermediate vent, inner and outer channel b7 cylinders, upper downcomer pipes, and "Z" junction forging. Tube Bundle Subassembly The tube bundle contains 2850 7/8 inch 0.D. x 0.045 inch minimum wall tubes spaced on a 1-5/16 inch triangular pitch. The tubes are joined to both the upper and lower tubesheets. The tube-to-tubesheet joints are made by front face fillet welding the tube ends to specially prepared stubs in the tube sheet face. An automatic T.I.G. welding procedure is used to join tubes to tubesheets. The tube ends will be expanded into the tubesheet holes The upper and lower tubesheets are designed with a minimum inner and outer rim 27 so as to permit better thermal response with the perforated portion. O V Amend. 27 5.3-23 Oct. 1976
The strongback pipec 34-5/8 inches diameter, is welded to the upper tubesheet and extends down to approximately 6 inches above the lower tube-sheet. The 0.D. of the strongback is machined to provide little distance 2d between it and the inner support plates reducing by-pass flow away from the cubes. The lower portion of the strongback contains three slots which engage matching keys attached to the downcomer. The slot and key combination pro-vides torsional and lateral stiffness for the lower tubesheet-head complex 27l during shipping and operation. A mechanism (gas trap) is incorporated into the lower downcomer region to prevent gas from being entrapped in the annulus between the strongback and downcomer. The outer shroud is prr,vided as the outer bound of the flow in the heat transfer zone of the bundle. The uppermost portion of the outer shroud is the distribution cylinder which forms the entrance zone to the heat trans-fer tubes. This distribution cylinder is designed to insure uniform circum-ferential flow into the bundle. This perforated entrance cylinder is welded to the upper tubesheet. Eight additional rings and eight cylindrical sections welded together make up the remainder of the outer shroud complex. The machined rings serve as stiffeners. One ring is utilized as a primary bypass seal attach-ment point. The ring at the primary inlet nozzle is a support point for saddle and strap supports during shipment. The outer shroud terminates 1 to 2 inches below the lowermost support plate to allow for exit flow fron the heat transfer zone 57 to the primary outlet nozzle, I The tubes in the tube bundle are supoorted by nineteen support plates. 27 The uppermost support plate extends to the I.D. of the shell. This plate forms the upper bne .dary of the inlet plen im. Tne plate limits direct impingement of sodium upon the face of the upper tulesheet. The second support plate is a complete plate, i .e. , it extends from the 0.D. of the strongback to the I.D. of the outer siroud. This plate contains flow holes preferentially drilled to i further ensure ui:iform bundle flow. The lowermost plate is also complete and 2 71 aids in uniform distribution through the tube bundle exit. The sixteen inter-mediate support-baffles are overlapping "douphnut" baffles which contain unifonn flow holes resulting in a combination of cross and axial flow. The plates are supported by eighteen tierods and spacers, 6 at the
! .D. , 6 intermediate and 6 at the 0.D. The tierods are threaded into tapped holes in the shell side of the upper tubesheet.
support for heat exchangers. The lower tubesheetThis is a standard method of and hemispherical head is a floating assembly. The head is welded to the lower tubesheet and serves as a pressure boundary between the i itermediate and primary systems. A baffle flow ring is provided to uniformly distribute the intermediate flow Amend. 57 5.3-24 Nov. 1980 I l
Although equipment sizing calculations, plant transient analysis and ps safety evaluations are not based on expected operating conditions, predic- \j tions of fuel life are computed based on nominal parameters plus 2 standard deviations of the reactor inlet and outlet temperatures based on the above mentioned Monte Carlo technique. 31 It is to be emphasized that accident analysis, structural evaluation of pemanent plant components and component sizing are based on a fixed set of parameters conservatively chosen and are not based on the expected temperatures and flows resulting from a statistical analysis. 25 The thermodynamic and physical properties of the fluids and materials are based on References 1 and 2. 5.3.3.1.1 Structural Evaluation Plan (SEP) A structural evaluation plan for each segment of the Heat Transport System is not presently available. However, to facilitate the orderly review and verification of the stress report to be provided to the owner by the man-ufacturer, structural evaluation plans shall be prepared for each major Class 1 component of the primary heat transport system. The SEP for each component shall provide a description of the methods of analysis which the manufacturer contempiates using in various phases of the structural analysis. The plan shall indicate the degree to which the manufacturer anticipates using elastic, simplified inelastic and rigorous inelastic methods of analysis in design iterations and for demonstrating compliance with the requirements of /^N RDT Standard F9-4T. The manufacturer shall identify any computer programs to C be used and shall describe, or provide the basic theory of the program, and identify the assumptions involved in their use. The manufacturer shall also im Amend. 31 5.3-33b Nov. 1976
5.3.3.1.2 Stress Analysis Verification The SEP for each major Class I component of the primary heat trans-port system (i.e., primary piping, primary pump, IHX, guard vessel *, and pri-mary check valve) will specify that a checklist be provided for identifying the anticipated analytical requirements for each component under normal, upset, emergency and faulted plant conditions. A sample of the structural design and analysis checklist to be used for each component is given in Figure 5.3-17. Structural evaluation details of how the manufacturer of each primary system Class I component intends to demonstrate compliance with structural requirements shall be as described in the following categories. Duty Cycle To confirm the structural integrity of the PHTS equipment and piping, the sequence of the application of loads or cycles must be selected to provide the most conservative loading history of the applicable events. This is especially important if inelastic analysis methods are being used. Per the PHTS Equipment Specifications, the designer or manufacturer must determine the most conservative sequence of the applicatior of the start-up to shutdown cycles using simplified analysis techniques subsequent to the detailed thermal transient and stress analysis. The equipment designer 57 in the ASME Code stress report must substantiate and document the load history used in the analysis of the equipment for the most severe cases. The application of material properties to be used in structural analysis of the PHTS equipment and piping is in accordance with ASME Code Case 1592 and RDT Standard F9-4T, and this application will give conservative results; sensitivity studies are not required. For example, the standard requires average yield stress properties to be used when calculating residual stress rupture damage. 19 Failure Modes The manufacturer shall identify the locations and failure modes which are expected to be dominant and the load conditions (pressure, thermal, seis-mic, etc.) associated therewith. Any failure modes not identified in the 27l 57l ASME Code Section III, Code Case 1592 and/or RDT Standard F9-4T to be guarded against for specified loads shall be identified. 41l
- Guard vessels, although classified as ASME III, Class 2. will be designed dnd analyzed as ASME Class 1 components but will not be tode stamped.
O Amend. 57 5.3-34 N v. 1980
References
- 1. TID-26666, " Nuclear Systems Materials Handbook", Hanford Engineering V 57 Development Laboratory, 1974.
la. " Sodium Components Development Program, Mass Transfer Investigations in Liquid Metal Systems, Quarterly Progress Report No. 9, Macch-May 1969," GEAP-10036, June 1969. Ib. W. T. Lee, " Biaxial Stress-Rupture Properties of Austenitic Stain-less Steels in Zirconium-Gettered Sodium," NAA-SR-12353, October 1967. ' Ic. R. C. Andrews, R. H. Hiltz, L. H. Kirschler, and R. J. Udavcak,
" Limited Comparison of the Relative Merits of 304 Versus 316 Stain-less Steel for Liquid Metal Service, Topical Report No.10."
MSAR-67-216, December 1967. Id. I. A. Rohrig, " Residual Elements and Their Effect on Applications of Austenitic Stainless Steels in the Power Industry," in Effects of Residual Elements on Properties of Austenitic Stainless Steels, pp. 78-89, American Society for Testing and Materials,1967. le. P. D. Goodell, T. M. Cullen, and J. W. Freeman, "The Influence of Nitrogen and Certain Other Elements on the Creep-Rupture Properties of Wholly Austenitic Type 304 Steel," Trans. Am. Soc. Mech. Engrs. 89, Series D, pp. 517-524 (1967). If. L. Colombier and J. Hochmann, Stainless and Heat Resisting Steels, St. Martin's Press, New York,1967.
- 19. D. E. Sonon. and G. V. Smith, "Effect of Interstitial Impurities, Grain Size and Temperature on Plastic Flow of Type 330 Stainless Steel," Trans. American Soc, Metals 58_, pp. 353-359 (1965).
Ih. R. B. Gunia and G. R. Woodrow, " Nitrogen Improves Engineering Properties of Chromium-Nickel Stainless Steels," J. Mater. 5_ pp. 413-430 (1970). li. P. N. Flagella, R. Hundel, and R. Ammon, "Effect of a Liquid-Sodium Environment of the Mechanical Properties of Type 316 Stainless Steel," Trans. Amer. Nucl. Soc. 14, pp. 589-590 (1971). lj. R. Hundal and P. N. Flagella, "The Effcct of Sodium Exposure on Stress Relaxation for Type 316 Stainless Steel at 1200 F and 1325 F," paper presented at the ASM/AIME Materials Engineering Congress, 17 Chicago, Illinois, October 1973. 5.3-74 Amend. 57 Nov. 1980 1
O O O TABLE 5.3-3 STEADY STATE OPERATING CONDITIONS FOR STRUCTURAL EVALUATION Hot 40% 80% 100% Refueling Standby Power Power Power T P T P T P T P T P System Location ( F) psig ( F) psig psig ( F) psig psig _( F) ( F) Reactor Outlet Nozzle 400 6.8 600 6.6 910 6.1 974 5.4 1015 4. 9 Reactor Inlet Nozzle 400 18.5 600 18.1 610 33 674 95 715 134 Pump Inlet 400 9.1 600 8.9 910 8.2 974 6.9 1015 6.0 Pump Discharge 400 14.4 600 14.3 910 34 974 115 1015 168
.m IHX Inlet 400 6.5 600 6.4 910 26 974 104 1015 155 Y
4 s Check Valve Inlet 400 1.3 600 1.4 610 19 674 88 715 133 NOTES:
- 1. "100%" power is actually 115% of rated power or 1121 MWt. The flow rate in each loop = 14.03 x 106#/hr = 1.015 times design flow. .
- 2. "80%" power is actually 897 MWt or 115% of rated 80% power (780 MWt). The loop flow = 11.22 x 106#/hr.
- 3. "40%" power is actually 40% of rated power (390 MWt) which is 35% of the 115% power condition. The loop flow is 4.84 x 10b#/hr.
i
TABLE 5.3-4 INTERMEDIATE HEAT EXCHANGER MATERIALS SPECIFICATION Product Form RDT Standard
- Grade Plate MS-1 Nov. 1974 304 & 316 Fo gings M2-2 Dec. 1974 F304 M2-4 Nov. 1974 F8 & F8 m Tubing M3-2 Dec. 1974 TP 304H Pipe M3- 3 Nov. 1974 TP 304 Bars M7-3 Nov. 1974 Bolting M6-1 Feb. 1975, M6-3 Feb. 1975 Nuts !!6-4 Feb. 1975 Sprinos M8-1 May 1975 27 Studs !!6-3 Feb. 1975 B8tt Diaphragm MS-1 Nov. 1974 TP 316 Spool Pieces M3-6 Apr. 1976 TP 316 57 Flange M2-4 Nov. 1974 The following additional chemistry controls apply:
Carbon - 0.04 to 0.08% for material <0.25 in. thick (Types 304 and 316)
- 0.05% minimum for material s0.25 in. thick (Type 304H)
*RDT Materials Standards apply only to those parts forming portions of the pressure retaining boundaries j or that are exoosed to liquid sodium or sodium con-talning environments.
Amend. 57 5.3-79
V TABLE 5.3-7 PRtMARY HEAT TRANSPORT SYSTEM PIPING MATERIALSSPECIFICATIONS(I) PRODUCT ITEM FORM ttATERIAL GRADES COVERED Pipe Welded Seamless (2f2) Types 316H, 304H, 304, 2 1/4 Cr-1 Mo Types 316H, 304H, 304, 2 1/4 Cr-1 Mo Fittings Types 316H, 304H, 304, 2 1/4 Cr-1 Mo Welded Seamless (2f2) Types 316H, 304H, 304, 21/4 Cr-1 Mo Branch Forging Types 316H, 304H, 304, 2 1/4 Cr-1 Mo Cor.nections Thermowells Fabrication Type 316H, 2 1/4 Cr-1 Mo Flued Heads, Forging plus Types 316H, 304H, 304 Pipe Support Fabrication Hangers Fabrication Various - as allowed by ASME Code \ Snubbers Fabrication Various - as allowed by ASME Code Clamps Fabrication Type 304 Auxiliary Steel Fabrication Various - as allowed by AISC Code
& Hardware IHX Vent-Line Fabrication Type 316H Flow Restrictor Shop Fabrication Fabrication Types 316H, 304H, 304, 2 1/4 Cr-1 Mo of Pipe Sub-Assemblies (Spools)
(1) The CRBRP Materials Specifications are based on ASME Code Section III and RDT Standards requirements. They are used for all large diameter sodium piping in both Primary and Intermediate Heat Transport Systems. (2) Welded and seamless products to these specifications are intercha 57l for the intended service. O Amend. 57 5.3-82 Nov. 1980
TABLE 5.3-8 IHX AND PRIMARY PUMP GUARD VESSEL MATERIALS SPECIFICATION
- ASME 44l Product Form , acification Grade Stainless Steel Forgings SA 336 F8 Plate SA 240 304 Bolting SA 540 B22 Class 2 j 44
*No RTD Materials Standards apply.
44 Amend. 44 5.3-83 April 1978
m 1 TABLE 5.3-9 d PRIMARY HEAT TRANSPORT SYSTEM COMP 0NENT WELD FILLER MATERIALS SPECIFICATIONS Weld Material Filler Material Form RDi 'tandards Classification 2 71 Stainless Steel Mi-1 March 1975 E 308L-15 or 16 Covered Electrodes E 308-15 or 16 E 316L-15 or 16 E 316-15 or 16 E 16 2 27 l Stainless Steal Mi-2 July 1975 ER 308, ER 308L Bare Rods and ER 316, ER 316L Electrodes ER 16 2 WELD MATERIALS FOR GUARD VESSELS Stainless Steel None Covered Electrodes E 308-15 Stainless Steel None ER 308 Bare Rods and 57 Electrodes Amend. 57 5.3-84 Nov. 1980
TABLE 5.3-10 1 PRIMARY REACTOR COOLANT PRESSURE 801'NDARY VALVES AND PUMPS Status Normal Component Active / Inactive Operating Mode
- PHIS Pumps Active Running 40l PHTS Check Valve Inactive NO PHTS Fill and Drain Valvest Inactive LC Reactor Coolant Make-up Pumpst Inactive Running PHTS High Point Vent Valvet Inactive LC
*NO - Normally Open LC - Locked Closed
+These pumps and valves are part of the primary coolant pressure boundary, but are not actually parts of the heat transport system. These components are discussed in Section 9.3 and 9.5.
O Amend. 40 5.3-85 July 1977
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p ( CHAPTER 7.0 INSTRUMENTATION AND CONTROLS TABLE OF CONTENTS PAGE
7.1 INTRODUCTION
7.1-1 7.1.1 Identification of Safety Related Instrumentation 7.1-1 and Control Systems 7.1.2 Identification of Safety Criteria 7.1-1 7.1.2.1 Design Basis 7.1-2 7.1.2.2 Independence of Redundant Safety Related Systems 7.1-3 7.1.2.3 Physical Identification of Safety Related Equipment 7.1-4 7.1.2.4 Conformance to Regulatory Guides 1.11 7.1-4
' Instrument Lines Penetrating Primary Re'ctor Containment" and 1.63, " Electric Penetration Assemblies in Containment Structures for Water-p Cooled Nuclear Power Plants" 7.1.2.5 Conformance to IEEE No. 323 "IEEE Standard for 7.1-4 Qualifying Class lE Equipment for Nuclear Power Generating Stations" f
7.1.2.6 Conformance to IEE No. 336 " Installation, Inspec- I.1-4 tion and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations" 7.1.2.7 Conformance to IEEE No. 338 " Periodic Testing of 7.1-5 ) Nuclear Power Generating Station Protection System" ; 7.1.2.8 Conformance to Regulatory Guide 1.22 " Periodic 7.1-5 Testing of Protection System Actuation Functions" 44l 7.1.2.9 Conformance to Regulatory Guide 1.47 " Bypassed 7.1-6 and Inoperable Status Indication for Nuclear Power Plant Safety Systems" 7.1.2.10 Conformance to Regulatory Gtide 1.53 " Application 7.1-6 of the Single Failure Criterion to Nuclear Power Plant Protection Systems"
/-i Amend. 44 April 1978
7.1.2.11 Conformance to Regulatory Guide 1.62 " Manual 7.1-6 Initiation of Protective Functions 7.1.2.12 Regulatory Guide 1.89 " Qualification of Class IE 7.1-6a Equipment for Nuclear Power Plants" 1 22 7.2 REACTOR SHUTDOWN SYSTEM 7.2-1 7.2.1 Description 7.2-1 7.2.1.1 Reactor Shutdown System Description 7.2-1 7.2.1.2 Design Basis Information 7.2-6 7 . 2.1. 2.1 Primary Reactor Shutdown System Subsystems 7.2-7 57 7.2.1.2.2 Secondary Reactor Shutoown System Subsystems 7.2-9 7.2.1.2.3 Essential Performance Requirements 7.2-11 7.2.2 Analysis 7.2-13 7.3 ENGINEERED SAFETY FEATURE INSTRUMENTATION AND CON. TROL 7.3-1 7.3.1 Containment Isolation System 7.3-1 7.3.1.1 System Description 7. 3 -1 7.3.1.2 Design Basis Information 7.3-2 7.3.1.2.1 Containment Isolation System Subsystems 7.3-2 7.3.1.2.2 Essential Performance Requirements 7.3-3 7.3.2 Analysis 7.3-3 7.3.2.1 Functional Performance 7.3-3 7.3.2.2 Design Features 7.3-4 7.4 INSTRUMENTATION AND CONTROL SYSTEMS 7.4-1 EQUIRED FOR SAFE SHUTDOWN 7.4.1 Steam Generator Auxiliary Heat Remaval 7.4-1 Instrumentation and Control System 7.4.1.1 Design Description 7.4-1 7.4.1.1.1 Function 7.4-1 l 7.4.1.1.2 Equipment Design 7.4-1 O Amend. 57 7-i. i Nov. 1980
7.6 OTHEF INSTRUMENTATION AND CONTROL SYSTEtiS REQUIRED FOR STFETY 7.6-1 34 7.6.1 Plant Service Water and Chilled Water Instrumentation and Control Systems 7.6-1 7.6.1.1 Description 7.6-1 7.6.1.2 Analysis 7.6-1 7.6.2 Fuel Handling Safety Interlocks 7.6-1 7.6.2.1 Design Description 7.6-1 7.6.2.2 Design Analysis 7.6-3 7.6.3 Overflow Heat Removal Service Instrumenta-tion and Control 7.6-3 7.6.3.1 Design Description 7.6-3 7.6.3.1.1 Function 7.6-3 7.6.3.1.2 Design Criteria 7.6-3 7.6.3.1.3 Equipment Design 7.6-3a 7.6.3.1.4 Initiating Circuits 7.6-3c 44l 7.6.3.1.5 Bypass and Interlocks 7.6-3c 7.6.3.2 Design Analysis 7.6-3d 7.6.4 Heating, Ventilating, and Air Conditioning Instrumentation and Control System 7.6-3e 7.6.4.1 Design Basis 7.6-3e 7.6.4.2 Design Criteria 7.6-3e 7.6.4.3 Ftnctional Control Diagrams 7.6-3f 34 23 7.6.4.3.1 Reactor Containment Building HVAC I&C 7.6-3f 7.6.5 SGB Flooding Protection System 7.6-3f 7.6.5.1 Design Basis 7.6-3f 7.6.5.2 Design Requirements 7.6-3f O\ V 49 7.6.5.3 Design Requirements 7.6-3f Amend. 49 7-vi Apr.1979
i PAGE 7.6.5.3.1 Instrumentation 7.6-39 ! 49 7.6.5.3.2 Controls 7.6-39 34 7.7 INSTRUMENTATION AND CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 7.7-1 7.7.1 Plant Control System Description 7.7-1 7.7.1.1 Supervisory Control System 7.7-2 7.7.1.2 Reactor Control System 7.7-3 7.7.l.3 Primary and Secondary CRDM (Control Rod Drive Mechanism) Controller and Rod Position Indication 7.7-4 , 57 l 7.7.1. 3.1 Primary CRDM Control 7.7-4 7.7.1.3.2 Rod Position Indication System 7.7-6 i 7.7.1.4 Sodium Flow Control System 7.7-7 7.7.1.5 Steam Generator Feedwater Flow Coritrol System 7.7-8 7.7.1.5.1 Feedwater Flow Control Valve Control System 7,7-8 7.7.1.5.2 Main Feedwater Isolation 7.7-9 7.7.1.5.3 Feedwater Pump 57eed Control System 7.7-9 i r ! I i l Amend. 57 Nov. 1980 t . - - - . . - . - . . _ _ _ _ .. __ _ _ _ __
r~x LIST OF TABLES () TABLE N0. 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
.1- Deleted 49 7.1-6 Safety Related Electrical Instrumentation 7.1-13 23 and Control Equipment 7.2-1 Plant Protection System Protective 7.2-18
,- Functions 7.2-2 PPS Design Basis Fault Events 7.2-19 7.2-3 Essential Performance Require.nents 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 Functions and 7.5-34 Summary 7.5-2 Reactor and Vessel Instrumentation 7.5-39 44 7.5-3 Sumation 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 Control Room Arrangements 7.9-8
, Amend. 50 7-1x June 1979
LIST OF FIGURES FIG. NO. PAGE 7.2-1 Reactor Shutdown System 7.2 24 7.2-2 HTS Coolant Pump Shutdown 7.2-25 7.2-2A Typical Primary PPS Instrument Channel 7.2-26 1 L;ic Diagram 7.2-2AA RSS Bypass Function Block Diagran 7.2-27 15 7.2-2B Primary PPS Logic Diagram 7.2-28 7.2-2C ~;pical Secondary PPS Instrutrent Channel 7.2-29 Logic Diagram 7.2-20 Secondary PPS Logic Diagram 7.2-30
)
44l 7.2-3 Typical Primary Subsystem 7.2 31 7.2-4 Typical Secondary Subsystem 7.2-32 7.2-5 Functional Block Diagrams of the Flux-Delayed 7.2-33 Flux, High Flux, Flux-Pressure, and Reactor Vessel Level Protective Subsystems 7.2-6 Functional Block Diagrams of Primary Pump 7.2-34 Electrics and Primary to Intermediate Speed Ratio Protective Subsystems 7.2-7 Functional Block Diagrams of the IHX Primary 7.2-35 Outlet Temperature and Steam to Feedwater Flow Mismatch Protective Subsystems 7.2-8 Functional Block Diagrams of the Flux-Total 7.2-36 Flow, Startup Nuclear, Modified Nt: clear Rate, and Primary to Intermediate Flow Rate Protective Subsystems 7.2-9 Functional Block Diagrams of the Steam 7.2-37 Drum Level and Loss of Condenser Vacuum Protective Subsystems 57 7.2-10 Functional Block D'agrams of the Evaporator 7.2-38 Outlet Sodium Tempercture and Sodium Water Reaction Protective Subsystems 7.3-1 Containment Isolation System Block Diagram 7.3-6 O l Amend. 57 Nov. 1980 7-x
PAGE l 7.3-2 Containment Selection System Logic Diagram 7.3-7l1 7.4-1 SGAHRS Initiation Logic 7.4-11 7.5-1 CRBRP Flux Monitoring System Block Diagram 7.5-43 7.5-2 CRBRP Flux Monitoring System Instrument Range
- Coverage 7.5-44 7.5-3 Failed Fuel Monitoring System 7.5-45 i 7.5-4 Main Sodium Stream First Pass Hydrogen 1
Concentration Change vs. Leak Rate 7.5-46 7.5-4a Main Sodium Stream First Pass Oxygen 49 Concentration vs. Leak Rate 7.5-46a 7.5-5 Hydrogen Concentration vs. Time for Various Water Leak Rates 7.5-47 17 5-6 Sodium-Water Reaction Pressure Relief 44 Control System (Typical for each SGS loop) 7.5-48 l 7.6-1 Functional Control Diagram for Reactor Contain-j mett Building Exhaust Fans - Start
. 7.6-5 7.6-2 Functional Control Diagram for Reactor Contain-
] ment Building Exhaust Fans - Stop 7.6-6 7.6-3 Functional Control Diagram for Reactor Contain-ment Building Supply Fans 7.67 1 5 1 7.6-4 Functional Control Diagram for Reactor Contain-ment Building Exhaust Fan Dampers 7.6-8 1 1 7.6-5 Functional Control Diagram for Reactor Contain-mnt Building Supply Fan Dampers 7.6-9 7.6-6 Functional Control Diagram for Reactor Contain-ment Building Main Exhaust and Supply Dampers 7.6-10 7.6-7 Functional Control Diagram for Reactor Containment Building Exhaust and Supply Relief Dampers 7.6-11 7.6-8 Functional Control Diagram for Reactor Containment Building Operating Floor Area Dampers 7.6-12 23 p Amend. 49 l- () 7-xi Apr. 1979 I_ , _ _ . __ _ . . _ _ _ _ _ _ . _ _ _ _ . , - . _ _ . _ - . _ . _ , _.
l I 7.6-9 Functional Control Diagram for Reactor Con-tainment Building 733 Foot Branch Duct Damper 7.6-13 7.6-10 Functional Control Diagram for Reactor Con-tainment Building Operating Floor AC Unit and Dome Recirculating Fans 7.6-14 7.1 11 functional Control Diagram for Reactor Con-tainment Building Operating Floor AC Unit Fan
' Dampers 7.6-15 7.6-12 Functional Control Diagram for Reactor Con-tainment Building Operating Floor Unit Cooler Fans 7.6-16 ,23 7.C-13 Safety Cliss Equipment n! ital Fus t-ootor 34 l ~.6-1 7.7-1 Plant Control System 7.7-20 7.7-2 Supervisory Control System 7.7-21 7.7-3 Reactor Control 7.7-22 7.7-4 CRDM Controller and Power Train for Primary Rods 7.7-23 7.7-5 Block Diagram of Primary Rod Group Control 7.7-24 57l 7.7-6 Deleted 7.7-25 7.7-7 Sodium Flow Control System Flow / Speed Control 7.7-26 7.7-8 Fuel Handling and Storage Control System 7.7-27 7.9-1 Control Room Layout 7.9-11 7.9-2 Typical Control Panel (Side View) 7.9-12 7.9-3 Main Control Panel Plan View 7.9-13 57 7.9-4 Typical Control Panel Wiring Layout 7.9-14 7-xii Amend. 57 Nov. 1980
O 7.0 IflSTRUMENTATION AND CONTROLS
7.1 INTRODUCTION
This chapter includes a description of the instrumentation and Control Systems provided for the CRBRP. Particular emphasis is placed on the description of safety related systems, which include the Plant Protection System and the safety related display instrumentation required to maintain the plant in a safe shutdown condition. The Plant Protection System includes all equipment to initiate and carry to completion reactor heat transport and balance of plant shutdown, decay heat removal and containment isolation. Safety related display instrumentation assures that the operator has suffi-cient information to perform required manual safety functions and monitor the safety status of the plant. Major control systems not required for safety are described and analysis is included to demonstrate that even gross failure of those systems does not prevent Plant Protection System action.
' Analysis is also included to demonstrate that the requirements of the NRC 43 General Design Criteria, IEEE Standard 279-1971, applicable NRC Regulatory Guides and other appropriate criteria and standards are satisfied.
7.1.1 Identification of Safety Related Instrumentation and Centrol Systems Table 7.1-1 lists the Safety Related Instrumentation and Control p\ Systems and includes the definition of Safety Related Equipment from Section 3.2.1. The entire Plant Protection System, including the Reactor Shutdown System, the Containment Isolation System and the Shutdown Heat Removal System is safety related. The Reactor Shutdown System input variables are described in Section 7.2. The Containment Isciation Instr mentation and Control System 26 described in Section '.3 while the Shutdown Heat Removal System Instrumentation and Control System is described in Section 7.4 and Section 7.6. The instru-33l mentation which provides input signals to the Plant Protection System is also safety related and is described in Section 7.5. Safety Related Display Instrumentation, which assures that the operator has sufficient information to monitor the safety status of the plant and maintain it in a safe shutdown condition, is discussed in Sections 7.5 and 7.9. Other safety related instru-33l mentation and control systems including Emergency Chilled Water System, Emergency Plant Service Water System, and Fuel Handling and Storage Interlocks , 57 are described in Section 7.6. ll 7.1.2 Identification of Safety Criteria In addition to meeting the requirements of the CRBRP General Design Criteria (refer to Section 3.1), the safety related I&C systems will be designed to meet the applicable requirements of the Regulatory Guides and IEEE Standards listed in Tables 7.1-2 and 7.1-3. The means of compliance with the guides and standards applicable to all safety related instrumentation and control equipment are described in paragraphs 7.1.2.2 through 7.1.2.11. O v 7.1-1 Amend. 57 Nov. 1980
Compliance with guides or standards applicable to specific I&C systems or equipment are described in the paragraphs related to those systems. In addition to meeting the requirements of the Regulatory Guides and IEEE Standards, the safety related equipment will be designed to meet the applicable requirements of the RDT Standards listed in Table 7.1-4. The instrument error and other performance consideration are addressed in the description of individu?1 subsystems. 7.1.2.1 Design Basis The Plant Protection distem (PPS) includes the Reactor Shutdown System (RSS), the Containment Isolation System and the Shutdown Heat Removal Systems. The Reactor Shutdown System consists of a Primary and a Secondary System either of which is designed to initiate and carry to completion trip of the control rods and sodium coolant pumps to prevent the results of postulated fault conditions from exceeding the allowable limits. Table 45l 4.2-35 shows the basis for Primary and Secondary RSS performance for the defined fault categories. The performance limits for the fuel and cladding are identified in Section 4. The Reactor Shutdown Systems are described in Section 7.2. The Containment Isolation System (CIS) is designed to react auto-matically to prevent or limit the release of radioactive material to the outside environment. The system acts to isolate the interior of the contain-ment by closing the containment isolation valves in the event that radioactive material is released within the containment. Radiation monitors within the containment boundary are used to activate the CIS. A description of this system is given in Section 7.3. The Shutdown Heat Removal Instrumentation and Control System is designed to provide assurance against exceeding acceptable fuel and reactor coolant system damage limits following normal and emergency shutdowns. The description of this instrumentaon and control is given in Section 7.4 for the removal througn the auxilt ..y steam / water system (Steam Generator Auxiliary Heat Removal System (SGAHRS) and Outlet Steam Isolation System (OSIS)) and Section 7.6 for removal through tFe NaK to air system (Direct Heat Removal 26l 57 System (DHRS)). Sufficent instrumentation and associated display equipment will be provided to permit effective determination of the status of the reactor at any time. Section 7.5 provides a description of the instrumentation pro-vided. In Section 7.9, a description of the control room, control room 57 l layout, operator-control panel interface, instrument and display groupings and habitability are given. 7.;-2 Amend. 57 O Nov. 1980
1 1 O 7.1.2.2 Independence of Redundant Safety Related Systems
- To assure that independence o/ redundant safety related equipment is preserved, the following specific physical separation criteria are
, imposed for safety related instrumentation, o All interract PPS wiring shall be ran in conduits (or equivalent) , with wiring for redundant channels run in separate conduits. l Only PPS wiring 'shall be included in these conduits. Primary RSS wiring shall not be run in the same conduit as secondary RSS wiring. Wiring for the CIS may be run in conduits containing
- either primary RSS wiring or conduits containing secondary RSS i wiring, but never intermixed. 24 l
0 Wiring for other safety related systems may be run in conduits I containing either primary RSS wiring or conduits containing secondary RSS wiring, but never intermixed, provided that no
; degradation of the separation between primary and secondary 24 i
RSS results. o Wiring for redundant channels shall be brought through separate , containment penetrations with only PPS wiring brought through ! these penetrations. Primary RSS wiring shall not be brought I through the same penetration as secondary RSS wiring. Wiring for the CIS and other safety related systems will be brought through the same penetration as the RSS wiring with which it is routed. 24 57 o Instrumentation equipment associated with redundant channels shall be mounted in separate racks (or completely, metallically enclosed compartments). Only PPS channel instrumentation shall
; be mounted in these racks. Primary RSS equipment shall not be 57 located in the same rack as Secondary RSS equipment.
9 The physical separation between conduits, penetrations, or racks containing redundant instrument channels shall be specified on an individual case basis to meet the requirements of Regulatory Guide 1.75. This separation shall provide assurance that credible single events do not simultaneously degrade redundant channels 57l or redundant shutdown systems. o The wiring from a PPS buffered output which is used for a non-PPS purpose may be included in the same rack as PPS equipment. The PPS wiring shall be physically separated from the non-PPS wi ri ng. The amount of separation shall meet the requirements of IEEE 384-1974. o Electrical power for redundant PPS equipment shall be supplied from separate sources such that failure of a single power sou ce t
! Amend. 57 7.1-3 Nov. 1980
does not cause failure of more than one redundant channel . The power sources and associated wiring shall be separated, as speci-fied in Section 8. The criteria for cable tray fill, cable derating, cable routing in congested or hostile areas, fire detection and protection in cable areas, and cable markings are defined in Section 8. Separation of redundant safety related equipment within the control boards is described in Section 7.9. 7.1.2 3 Physical Identification of Safety Related Equipment The Plant Protection System equipment will be identified distinc-tively as being in the protection system. This identification will distin-guish between redundant portions of the protection system such that qualified personnel can distinguish whether the equipment is safety related and, if so, which channel. Color coding, cabinet and wire labeling and other techniques as appropriate will be used. 7.1.2.4 Conformance to Regulatory Guides 1.11 " Instrument Lines Penetrating Primary Reactor Containment" and 1.63 " Electric Penetration Assem-blies in Containment Structures for Watercooled Nuclear Power Plants" Thore are no instrument lines as defined in Regulatory Guide 1.11 which penetcate primary reactor containment. All electric penetration assem-blies in the containment vessel will be designed, constructed and installed in accordance with Regulltory Guide 1.63 and IEEE Standard 317-1972. 7.1.2.5 Conformance to IEEE Standard 323-1974 "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations 1 All Class IE equipment will be q;alified to confirm the acequacy of the equipment design under normal, abnormal, and postulated accident conditions for the perfomance Class IE functions. This will be accomplished through a disciplined program af quality assurance and testing. Type testing, operating experience and analysis will be used to assure that for each type of Class IE equipment the design and manufacturing processes are such that there is a high degree of confidence that the equipment of the same type will perform as required. 7.1.2.6 Conformance to IEEE Standard 336-1971 " Installation, Inspection and { Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations" l i The installation, inspection and testing of the instrumentation, electrical and electronic equipment during construction will conform to the requirements of IEEE Standard 336-1971. The quality assurance program for the safety related instrumentation and control equipment will conform to the l i requirements of Regulatory Guide 1.30. Refer to Chapter 17 for a descrip-tion of the quality assurance program. Amend. 1 July 1975 7.1-4
7.1.2.7 Conformance to IEEE Standard 338-1971 " Periodic Testing of y Nuclear Power Generating Station Protection System" Capability for periodic testing is provided ta ensure that the Plant Protection System and safety related instrumentation and equipment meet the necessary performance objectives. This testing capability includes provisions for on-line-testing, testing during shutdown, and calibration as appropriate. The basic objective of the testing is to assure that the Plant j Protection System is performing within its specifications; or conversely, to detect functional failures of redundant components or degradation of impor-tant perfonnance features. Further, neither the testing of the equipment nor the features incorporated to permit testing will compromise the inde-57 ) pendence of redundant components. Since both Reactor Shutdown Systems use 3 redundant channels, on-line testing is possible. Specific testing feature? are described with the associated hardware. 7.1.2.8 Conformance with Regulatory Guide 1.22 " Periodic Testing of Pro-tection System Actuation Functions" Plant Protection System actuation devices are periodically tested with the RSS during reactor operation. Since the RSS scram circuit breakers and scram solenoid valves are arranged in 2 out of ' ., incidence logic, scram circuit breakers or solenoid valves may be idually tested during reactor operation. The HTS breakers are ar- ed to allow testing d by using a test breaker to maintain power to the coc.ont pumps while tcsting the breaker. Knile the Reactor Shutdown Systems can be tested through the scram circuit breakers or solenoid valves, the primary and secondary rod release capability cannot be tested during reactor operation since dropping a single control rod will initiate a reactor scram. Scram actuator and control rod insertion times will be functionally tested every time the plant is shut down for refueling. The Containment Isolation System actuated equipment can be tested during reactor operation since containment isola-tion can take place without necessitating a reactor scram. Shutdown Heat 57 Removal actuation system can be tested cn line. Specific details concerning tett and calibration for each pro-tection system channel will not be available until the FSAR. However, the capability for on line functional testing will be provided for all protection system channels. The design will satisfy the requirements of Section 4.10 of IEEE 279-1971 and the recommendations of Regulatory Guide 1.22. The functional performance of a protection system channel will, in general, be tested by inserting a signal in the instrument channel as near as is practical to the sensor. For all tests requiring disconnection of the ! sensor or modification of the analog input signal, the channel will be placed in a safe condition by tripping the channel comparator associated with the protection system charmel under test. Exception to this rule 57 l is made for the functional test of the PPS nuclear flux channels. Since addition of the nuclear flux test signal to the analog signal always drives pd the channel under test toward a safe (i.e. , tripped) condition, the compa- ! rator output is not placed in trip before the functional test begins. All 16 Amend. 57 Nov. 1980 7.1-5
l protection system instrument channels are functionally tested by varying the magnitude of the test signal through the trip point to verify that the comparator trips, then readjusting its magnitude to reset the compa-rator. After this functional test is completed, the test signal is removed from the instrument channel, and the instrument channel operation 57 1 is restored. Calibration checks to assure that the protection system channel meets its performance requirements will be accomplished at periodic intervals during regul irly scheduled shutdowns. Actuated equip-ment will, in general, be testable on line. In cases where this is not practical (for example, a control rod cannot be dropped during operation without scramming the reactor), the recommendations of EICSB 22 will be met. 16 7.1.2.9 Conformance to Regulatory Guide 1.47 " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems" Administrative procedures for indicating the bypass or inoperable status of portions of the protection system or supporting systems will be supplemented by a system that automatically indicates at a system level the bypass or deliberately induced inoperability of the Protection System, systems actuated or controlled by the Protection System, or supporting systems that must be operable for the Protection and related systems to perform their safety related functions. An indication of each bypass or deliberately induced inoperability will be displayed in the control room in accordance with Regulatory Guide 1.47. 7.1.2.10 Conformant to Regulatory Guide 1.53 " Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems" Any single failure within the protection system will not prevent proper protection action at the system level when required. The Plant Protection System is periodically tested so that failures are detected. Test schemes will be designed to duplicate as closely as possible the operation being tested. Design precautions include two independent, redun-dant diverse sht.tdown systems, each capable of shutting down the reactor; physical independence between redundant channels; and physical barriers utilized for separation between redundant channels. The use of fire retardant materials in construction, fire retardant cable and wire insu-lation jackets, and physical separation between redundant circuits is relied upon to prevent or mitigate the consequences of a fire, 7.1.2.11 Conformance to Regulatory Guide 1.62 " Manual Initiation of ProteClive Functions" The Plant Protection System will provide for 21ual actuation of each protective action at the system level. The mamnl initiation of a protective action will result in a Protection Systm response identical to automatic actuation of the same protective actior . For example, manual trip buttons parmit operator initiation of reactor scram and containment isolation. The amount of equipment conmon to both manual and automatic initiation is minimized. Manual initiation of protection actions is designed to go to completion once initiated. No single failure within automatic, 57 manual, or common portions of protective subsystems will prevent ini-tiation of a protective system action by manual or automatic means. Amend. 57 7.1-6 Nov. 1980
s./ 7.1.2.12 Regulatory Guide 1.89 " Qualification of Class .lE Equipment for Nuclear Power Plants" IEEE Standard 323-1974 will be applied to the safety related instrumentation and control equipment as indicated in Section 7.1.2.5. This guide further recommends the use of a source term that is equivalent to one based on the failure of all safety related equipment designed to prevent or mitigate the condition from which the source term is derived. The purpose of qualification is to assure that the safety related equipment will perform under the environmental conditions 57 to which it may be subjected. It is highly inconsistent to require the qualification of equipnent to radiation levels which could not be 57l reached quali fied.even as a result of complete failure of the very equipment being As such, Regulatory Guide 1.89 source term requirement will not be imposed on the cafety-related instrumentation ar.d control systems and components. Radiation environments to be considered in qualifying safety-related instrumentation and control systems and components will be determined considering the source tenn vrring and/or after the applicable desig1 basis events, the spatial location, shielding and equipment. '22 (D L) /~N V 7.1-6a Amend. 57 Nov. 1980
TABLE 7.1-1 f) v SAFETY RELATED INSTRUMENTATION AND CONTROL SYSTEMS
- Reactor Shutdown Systems Includes all RSS sensors, signal conditioning calculation units, compara-57 tors, buffers, 2/3 logic, scram actuators, scram breakers, scram solenoid valve power sources, control rods, HTS shutdown logic, coolant pump breakers, and mechanical mounting hardware (equipment racks).
Containment Isolation System Includes radiation monitoring sensors, signal conditioning, comparators, 2/3 logic, containment isolation valve actuators and valves. 57l Shutdown Heat Removal System Instrumentation and Control System Includes initiating sensors, signal conditioning, calculation units, compa-rators, logic, auxiliary feedwater pump actuators and controls including feedwater turbine pump, PACC DHX actuators and controls, steam relief valve actuators and valves; sensors, signal conditioning, logic and 26 57) actuators related to shutdown heat removal functions of DHRS including control of sodium and NaK pumps and air blast heat exchangers; and sensors, signal conditioning, logic and actuators related to removal of heat from the EVST. p V Other Safety Related Instrumentation and Control Includes Instrumentation and Controls for portions of the following functions to assure the plant is maintained in a safe shutdown condition: Emergency Chilled Water System 33 . Emergency Plant Service Water System Instrumentation necessary to assure plant is maintained in safe 49 shutdown status (See Table 7.5-4) Fuel Handling and Storage Safety Interlocks Heating, Ventilating, and Air Conditioning System Recirculating Gas Cooling System 1
*The Clinch River Breeder Reactor Plant (CRBRP) structures, systems, and components important to safety are to be designed to remain functional to the event of a Safe Shutdown Earthquake (SSE). These plant features are also designated as safety-related features in the SAR. These include, but are not limited to, those structures, systems and components which are necessary:
- a. To assure the integrity of the Reactor Coolant Boundary;
- b. To shut down the reactor and maintain it in a safe shutdown condition;
- c. To prevent or mitigate the consequences of accidents which could result l in potential off-site exposures comparable to the guideline exposures of 10CFR100.
h v NOTE: Class IE equipment loads are identified in Chapter 8. ! 1 Amend. 57 7.1-7 Nov. 1980
TABLE 7.1-2 LIST OF REGULATORY GUIDES APPLICABLE TO SAFETY RELATED INSTRUMENTATION AND CONTROL SYSTEMS 1.6 Independence Between Redundant Power Sources and their Dis-tribution Systems (as discussed in Sections 8.3.1.2 and 8.3.2.2) 1.12 Instrumentation for Earthquakes 1 27" 1.17 Protection of Nuclear Power Plants Against Industrial Sabotage 1.22 Periodic Testing of Protection System Actuation Functions 1.28 Quality Assurance Program Requirements (Design and Construction) 1.29 Seismic Design Classification 1.30 Quality Assurance Program Requirements for the Installation, 22 Inspection, and Testing 2f Instrumentation and Electric Equipment 1.32 Use of IEEE Std 308-1971 " Criteria for Class lE Electric Systems for Nuclear Power Gea ratina Stations" 1.40 Qualification Tests of Continuous Duty Motors Installed Inside the Containment of Water Cooled Nuclear Power Plants 1.47 Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems 1.53 Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems 1.62 Manual Initiation of Protective Actions 1.63 Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants 1.64 Quality Assurance Prngram Requirements for the Design of Nuclear Power Plants 1.73 Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants 1.75 Physical Independence of Electric System T.70 control Room Habitability During Chemical Release (as discussed in Section 6.3), 1.89 Qualification of Class IE Equipment for Nuclear Power Plants (as discussed in Section 7.1.2.5). 7.1-8 Amend. 22 June 1976
l tab. 7.1-3 LIST OF IEEE STANDARDS APPLICABLE TO Q SAFETY RELATED INSTRUMENTATION AND CONTROL SYSTEfiS IEEE-279-1971 IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations IEEE-308-1974 Criteria for Class lE Power Systems for Nuclear Power Generating Stations IEEE-317-1972 Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations IEEE-323-1974 IEEE Trial-Use Standard: General Guide for Qualifying Class IE Electric Equipment for Nuclear Power Generating Stations 57l IEEE-323-A-1975 Supplement to the Foreword of IEEE 323-1974 IEEE-334-1971 IEEE Trial-Use Guide for Type Tests of Continuous-Duty Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations l IEEC-336-1971 IEEE Standard: Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations O) ( IEEE-338-1971 IEEE Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-344-197 5 IEEE Std. 344-1975, IEEE Reconmended Practices for Seismic Qualification of Class 1 Equipaent for Nuclear 46 Power Generating Stations IEEE-352-1972 IEEE Trial-Use Guide: General Principles for Reliability Analysis of Nuclear Power Generating Station Protection Systems IEEE-379-1972 IEEE Trial-Use Guide for the Application of the Single-Failure Criterion to Nuclear Power Generatir.g Station Protection Systems IEEE-384-1974 IEEE Trial Use Standard Criteria for Separation of Class IE Equipment and Circuits IEEE-420-1973 Trial-Use Guide for Class IE Control Switchboards for Nuclear Power Generating Stations 01 v . 7,j_g Amen *. 27 Nov. 1980
TABLE 7.1-4 LIST OF RDT STANDARDS APPLICABLE TO SAFETY RELATED INSTRUMENTATION AND CONTROL SYSTEMS Cl-1T Instrumentation and Control Eauipment Grounding and
- Shielding Practices, January,1973, Amendment 1, j January 30, 1975.
C4-5T 27i Permanent Magnet Flowmeter for Liquid Metal Piping Systems, April, 1974. C5-lT Inductive Level fleasurement Sensor for Use in Liquid fietal, 44 March, 1975 C6-lT NaK Transmission High-Temperature Pressure Transmitter for Liquid Metal Service, March,1971, Amendment 1, May 1971-Amendment 2, November 1971-Amendment 3, October 1973-Amendment 4, January,1974-Amendment 5, June 1974. C7-6T Thermocouple Material and Thermocouple Assembly, February, 1975. C10-lT Thermocouple Signal Transmitter, November 1971. C15-3T Current Pulse Preamplifier for use with Fission Counters, August 1971, Amendment 1, June 1973-Amendment 2, October 1974. C15-5T Fission Type Neitroc Detector, December,1971-Amendment 1, Oc tobe r, 1973. C15-6T Logarithmic Mean Square Voltage (MSV) Intermediate Range Neutron Flux Monitoring System, July,1971 C15-7T Gamma Compensated Ionization Chamber Assembly (Tixed Electrical Compensation) July ,1971, Amendment 1-August 197 3-Amendment 2, March 1974. C15-8T Direct Current Power Range Neutron Flux Monitoring System, July, 1971. 44 C15-10T Logarithmic Count Rate Source Range Neutron Flux Monitoring System, July,1971. C16-lT Supplementary Criteria and Requirements for RDT Reactor Protection Systems, December,1969. C16-2T Protection System Logic, April,1972-Amendment 1, June,1973. 27' C16-3T PPS Buffers, October,1971-Amendment 1, December 1971. Amend. 44 7.1-10 April 1978
i I n TABLE 7.1-4 (Continued) (_) C16-4 Plant Protection System Comparators, April 1972 - Amendment 1, 44 June 1973 57 C17-5T Metal-Sheathed, Mineral-Insulated Cable Bulk Material February 1973- Amendment 1, April 1974. E6-ST Collapsible-Rotor, Roller Nut Control Rod Drive Mechanism for Sodium Service, March 1971, Amendment 1, December 1972-Amendment 2, August 1973- Amendment 3, September 1974. F2-2 Quality Assurance Program Requirements, August 1973, Amendment 1, December 1973 - Amendment 2, March 1974 - Amendment 3, July 1975. 44 F2-4T Quality Verification Program Requirements, December 1974 41 F3-2T Calibration System Requirements, February 1973. F3-39T Testing of High Temperature Cable for Nuclear Detectors, g August 1971. F7-2T Preparations for Sealing, Packaging, Packing, and Marking of Components for Shipment and Storage, February 1969, Amendment 27 , 1, October 1971 - Amend 2, September 1972. 7.1-11 Amend. 57 Nov. 1980 v.)+
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O I l l 45 This page intentionally blank. O l l 7.1-12 1 Amend. 45 July 1978 O
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,O (--) 7.2 REACTOR SHUTDOWN SYSTEM 7.2.1 Description 7.2.1.1 Reactor Shutdown System Description The Reactor Shutdown System (RSS) consists of two independent and diverse systems, the Primary and Secondary Reactor Shutdown Systems, either of which is capable of Reactor and Heat Transport System shutdown. All anticipated and unlikely events can be terminated without exceeding the specified limits by either system even if the most reactive control rod in the system cannot be inserted. In addition, the Primary RSS acting alone can terminate all extremcly unlikely events without exceeding speci-fied limits even if the most reactive control rod in the system cannot be inserted. To assure adequate independence of the shutdown systems, mecha-nical and electrical isolation of redundant components is provided. Functional or equipment diversity is included in the design of instrumentation and electronic equipment. The Primary RSS uses a local coincidence logic con-figuration while the Secondary RSS uses a general coincidence. Sufficient redundancy is included in each system to prevent single random failure degradation of either the Primary or Secondary RSS. As shown in the block diagram of the Reactor Shutdown System, Figure 7.2-1, the Primary RSS is composed of 24 subsystems and the Secondary g RSS is composed of 16 subsystems. Figure 7.2-2A is a typical Primary RSS q~! instrument channel logic diagram. Each protective subsystem has 3 redundant sensors to monitor a physical parameter. The output signal from each sensor is amplified and converted for transmission to the trip comparator in the control room. Three physically separate redundant instrument channels are used. When necessary, calculational units derive additional variables from the sensed parameters with the calculational units inserted in front of the comparators as needed. The comparator in each instrument channel determines if that instrument channel signal exceeds a specified limit and outputs 3 redundant signals corresponding to either the reset or trip state. The 3 outputs of each comparator are isolated and recombined with the isolated outputs of the redundant instrument channels as inputs to three redundant logic trains. The recombination of outputs is in a 2 out of 3 local coin-cidence logic arrangement. 1 Operating bypasses are necessary to allow RSS functions to be bypassed during main sodium coolant pump startup, ascent to power, and two loop operation. Operating bypasses are accomplished in the instrument channels. For bypasses associated with normal three loop operation, the bypass cannot be instated unless certain permissive conditions exist which assure that adequate protection will be maintained while these protective functions are bypassed. Permissive comparators are used to determine when bypass conditions are satisfied. When permissive conditions are within the 57 allowable range, the operator may manually instate the bypass. If the O Amend. 57 7.2-1 Nov. 1980
permissive condition goes out of the allowable range, the protective function is automatically reinstated. The trip function will remain reinstated until the permissive conditions are again satisfied and the operator again manually initiates the bypass. Operator manual bypass control is not effective unless the bypass comparator indicates that permissive con-ditions are satisfied. A functional diagram of the Primary and Secondary bypasr permissive logic is shown in Figure 7.2-2AA. Two loop bypasses are established under administrative control by changing the hardware configuration within the locked comparator cabinets. These bypasses are also under pennissive control such that the plant must be shutdown to establish two loop operation and if the shutdown loop is 57 activated the bypass is automatically removed. Bypass permissives are part of the Plant Protection System (PPS), and are designed according to the PPS requirements detailed elsewhere in this section of the PSAR. Continuous local and remote indication of bypassed instrument channels will be provided in conformance with Regulatory Guide 1.47, "Byeassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems". 15 O 7.2-la Amend. 57 Nov. 1980
Figure 7.2-28 is a logic diagram of the Primary RSS logic trains. The outputs from the comparators and 2/3 functions are inputs to a 1 out of 24 general coincidence arrangement. The output of the 1/24 is an input to a 1 out of 2 with the manual trip function to actuate the scram i breakers. The scram breakers are arranged in a 2 of 3. When 2 or more i logic trains actuate the associated scram breakers, power to the control rods is open circuited and the control rods are released for insertion to shutdown position with spring assisted scram force. Open circuiting the
! control rod power initiates Heat Transport System shutdown.
In the Secondary RSS, the sensed variables are signal conditioned i and compared to specified limits by equipment which is different from the Primary RSS es uipment. The secondary logic is configured in. general rather j than local coincidence to provide additional protection against common mode
- failure. Each instrument channel comparator outputs its trip or reset i signal to a 1 of 16 logic module. The 3 redundant secondary instrument
, channels from each subsystem feed 3 redundant logic trains, which are coupled
] to the secondary scram actuators. Figure 7.2-2D is a logic diagram for the l Secondary RSS logic. f The Secondary RSS consists of 16 protective subsystems and monitors i a set of parameters diverse from the Primary RSS as shown in Table 7.2-1. i However, since a measure of nuclear flux is necessary in both the Primary and Secondary RSS, nuclear flux is sensed with compensated ionization chambers in the primary while fission chambers are used in the secondary. The Primary RSS monitors primary and intermediate pump speed while the Secondary RSS monitors primary and intermediate coolant flow. Similarly, the steam flow ! to feedwater flow r tio is used in the Primary RSS while the steam drum level is sensed for the Secondary RSS. Figure 7.2-2C is a typical Secondary RSS instrument channel logic diagram. Each protective subsystem has 3 redundant sensors to monitor a
- physical parameter. The output signal from each sensor is conditioned for I transmission to the trip cnmparator located in the control room. Redundant
- instrument channels are used. When necessary, calculational units are placed
- in front of the comparators to derive additional variables. The output of the comparators are input to redundant logic trains in a general coin-
~
j ]; cidence arrangement. Bypass of secondary comparators is implemented in the same fashion as in the primary system except that different equipment is used to provide the permissive comparator function.
, 57 Figure 7.2-2D is a logic diagram of the Secondary RSS logic trains.
i The outputs from the instrument channels are. input to a 1/16 general coinci-I dence arrangement. The 1/16 output controls the solenoid power sources l through isolated outputs. Isolated outputs are also provided to initiate Heat Transport System shutdown. A trip latch-in function is provided to - assure that once initiated, the scram will go to completion. The remaining l 43 redundant logic trains provide the other two signals for the 2/3 function. !O 7.2-2 Amend. 57 1 1 Nov. 1980
Figure 7.2-2 shows the RSS interface with the Heat Transport System (HTS) pump breaker control. Two HTS pump breakers are connected in series for each HTS pump. Each HTS breaker receives input from the Primary RSS and Secondary RSS pump trip logic. Upon receipt of a reactor trip signal from either Primary or Secondary RSS, the HTS pump breakers open to remove power from the primary and intermediate pumps. Provisions are made to allow testing of the HT: breaker actuation function during reactor operation. A test breaker is used to bypass the main HTS breaker during a test condition. Test signals are then inserted through the Primary or Secondary RSS pump trip logic to open the main HTS breaker. Mechanical interlocks are provided on the bypass breakers to prevent more than one main HTS breaker in any loop from being bypassed at a time. Control interlocks are provided which make the breaker test inputs ineffective unless the bypass breakers are properly installed. Main HTS breaker and test breaker position status is supplied as part of the 57 RSS status display on the main control panel. The RSS subsystems do not directly require the reactor operator or control system to implement a protective action. However, manual control devices to manually initiate each protective function are included in the 57 design of the Plant Protection System. Where signals are extracted from the Plant Protection System, buffers are provided. These buffers are designed to meet the requirements 27l of IEEE-279-1971, RDT Standards C16-IT, Dec. 1960 and C16-3T, Dec. 1971. The buffe.rs prevent the effects of failures on the non-PPS side from affectir.g the performance of the PPS equipment. The buffers are considered part of the PPS and meet all PPS criteria. System Testability Both Reactor Shutdown Systems are designed to provide on-line 571 testing capability. For the Primary RSS, overlapping testing is used. The sensors are checked by comparison with redundant sensor outputs and related measurements. Each instrument channel includes provisions for insertion of a signal on the sensor side of the signal conditioning electronics and test points to measure the performance at the comparator (or calculational unit) input. Where disconnection of the sensor is un-avaidable for test purposes, the comparator is tripped when disconnected. The instrument channel electronics including trip comparators and bypass permissive comparators are tested for ability to change value to beyond 57 the trip point and provide a trip input to the logic. The comparators and logic are tested by the PPS Monitor. A set of pulsed signals are inserted from the Monitor into the ccmparators associated with one sub-system and the logic output is checked by the Monitor to assure that logic trip occurs for the correct combinations of comparator trips. The logic and scram breakers are tested by manually tripping one logic train 57 and observing that the corresponding breakers trip. HTS breakers are 7.2-2a Amend. 57 Nov. 1980
)
i ./ tested by maintaining power to the pump through a bypass circuit breaker and manually inserting a test signal to the pump trip logic. For the Secondary RRS insertion of the test signal into a channel causes the entire train (comparator, logic, and scram solenoid valves) to trip. Testing of the pump breaker trip is identical to that for the 57 Primary RRS. System Instrumentation 57loff-normal plant conditions includes:The instrumentation used by the RSS to detect th e Neutron Flux The Primary RSS uses three compensated ionization chamber power range nuclear sensors evenly spaced around the 57 reactor vessel. The Secondary RSS uses three fission chamber wide range nuclear sensors evenly spaced around the reactor vessel. See Section 7.5.2 for detector details, o Reactor Inlet Plenum Pressure 57l The Primary RSS uses six pressure detectors, two per o HTS primary loop, located as close as practical to the reactor () vessel inlet plenum in the elevated primary cold leg piping. Each set of two detectors comprises an instrument channel. The outputs of the two detectors in each loop are auctioneered. The resultan* output signal is provided to the comparator. See Section 7.5.2 for detector details e Sodium Pump Speed The Primary RSS uses three redundant tachometers per primary 57l and intermediate HTS pump to measure pump speed. See Sec-tion 7.5.2 for detector details, e Sodium Flow The Scuondary RSS uses six permanent magnet flow-57 meters to measure HTS sodium flows. One flowmeter is located in each of the primary and intermediate cold legs. Each flow-meter provides three redundant measurements of loop flow. See Section 7.5.2 for detector details. e Reactor Vessel Sodium Level The Primary RSS uses four sodium level detectors evenly spaced within the reactor vessel. Three of these detec ors provide redundant accive signals to the RSS. The n) ( 57 fourtt detector is used as a spare. See Section 7.5.3 for detector details. Amend. 57 Nov. 1980 7.2-3
e Undervoltaae Relav 57l The Primary RSS uses nine undervoltage relays, three per coolant loop pump bus. The undervoltage relays are located on the HTS pump buses. e Steam Flow The Primary RSS uses three redundant steam mass flow 57l signals per loor. Each steam supply loop has one venturi flow-meter,three dif.erential pressure sensors, three temperature sensors, and three pressure sensors located between the super-57l heater exit and the main steam header. Three redundant steam mass flow signals are generated by pressure and temperature compensation of the venturi flowmeter analog signal. See Section 7.5.2 for detector details. e Feedwater Flow One venturi flowmeter, three differential pressure sensors, and three temperature sensors er steam generator loop supply the r Primary RSS with three redundant temperature-compensated feed-water mass flow signals. See Section 7.5.2 for detector 57 details. e IHX Primary Outlet Sodium Temperature 57l The Primary RSS uses three redundant thermocouples, mounted in three thermowell , per loop to measure the sodium temperature in the primary cold leg. See Section 7.5.2 for detector details. e Steam Drum Level The Secondary RSS uses three redundant reference column level 57 sensors to determine the water level in each steam drum. The level sensor is density compensated. See Section 7.5.2 for detector details. e Evaporator Outlet Sodium Temperature The Secondary RSS uses three redundant thermocouples per loop, 57 mounted in three thermowells. These themocouples are provided to monitor the sodium temperatures in each intermediate cold leg. See Section 7.5.2 for detector details. Amend. 57 Nov. 1980 7.2-4
N )
'"4 7 6 e Sodium-Water Reaction 57l The Secondary RSS uses' three redundant pressure sensors located in the reactyn products vent line inmediately downstream from each rupture u ok to detect if the rupture disks have blown. See Section 7.2.1 for details.
The configuration of the instrumentation in the protective subsystems is described in Section 7.2.1.2. 57l Primary Reactor Shutdown System Logic The Primary RSS logic is implemented using integrated circuits to minimize the scram delay time. Other advantages include minimizing power consumption and space required and maximizing testability, maintainability 57l and reliability. The Primary RSS logic is arr'anged as shown in Figure 7.2-3. 57l In each logic train, twenty-four ?/3 coincidence logic circuits feed a 1/24 module, whose output is coupled to the final actuation logic and rod actuators by a transistorized power amplifier. When only one comparator of (^')
\_ /
any or all protective functions is tripped, the logic signal output remains positive (reset). When any two comparators of a protective function trip and provide a negative logic signal to the protective logic, the output of the corresponding 2/3 module also trips to a negative logic signal. This negative logic signal in turn trips the 1/24 logic module which outputs a negative 5,/ logic signal to the final actuation logic and removes power from the scram breaker undervoltage coil. Light emitting diodes and phototransistors are utilized to provide l complete electrical isolation at strategic points through the Primary RSS l 57 logic. There is no electrical connection betreen the comparetor output and protective logic input. Consequently, an internal electrical fault in a single instrument channel or comparator cannot propagate to the I other channels, protective functions, or logic trains of the protective sys-tem. Each logic train is electrically isolated from the other so that protective action can be initiated regardless of any internal electrical fault in a single logic train. The equipment needed to implement the 24 protective subsystems of the Primary RSS includes the sensors, signal transmitters and amplifiers or 27 equivalent, calculational units, comparators, logic isolators, 2/3 logic modules, 1/24 logic modules, logic drivers, final scram actuation circuitry and breakers, buffers, permissives and bypasses. A three section equipment cabinet is used to 43l house the equipment for each of the three instrumei , channels including culational units, comparators, power supplies and buffers. A two section equipment ( )
'~#
7.2-5 Amend. 57 Nov. 1980
cabir,et is used to house the equipmt:r t for each of the three logic trains and single equipment cabinets hou:e signal conditioning equipment for each channel. This arrangement of equipment within cabinets provides the necessary mechanica 36 separation of redundant equipment.
! Secondary Reactor Shutdown System Logic The Secondary RSS logic consists of the 16 protective subsystems arranged in a general coincidence configuration, as shown in Figure 7.2-4. In this arrangement, the outputs of instrument channel A comparators are directly coupled to a 1/16 logic circuit in logic train A, as are the outputs of instrument channel B with logic train B and the 57, outputs of instrument channel C with logic train C.
When the sensed parameter in an instrument channel exceeds its setpoint (trips), the comparator outputs a zero (trip) signal to the 1/16 logic module, which in turn outputs a zero (trip) signal to the scram latch and scram solenoid valves (see Figure 7.2-2D). The 1/16 logic module outpit voltage changes to zero regardless of the output of the other comparators. The output of the 1/16 logic modules are combined in a 2/3 coincidence by toe scram 3 solenoid valves located within the Secondary RSS rod. Electrical 43 isolation of the logic output to the solenoid drivers and Heat Transport System 57 ure 7.2-20. Redundant iso-shutdown logic lated outputs are(HTS pump provided breakers) from is shownRS each Secondary in Fig'S logic + rain to the Secondary RSS pump trip logic where they are combined in a 2/ logic. Trip signals are provided to the HTS pump breakers when 2/3 of the redtedant 57 Secondary RSS channels are in a tripped condition. I
- The equipment of the Secondary RSS includes sensors, 51 signal conditioning equipment (transmitters), calculational units, com-43 I parators,1/16 logic modules, solenoid drivers, secondary final actuation logic and actuators, buffers, permissives and bypasses. The equipment is designed using hardware which is diverse from that used in the Primary RSS. Since each instrument channel is uniquely associated with a logic train, a four se.ction equipment cabinet houses each of the instrument 43 5_/ channels comparators, logic trains and solenoid drivers. Single equipment cabinets are used to house signal conditioning equipment for each ci innel.
This arrangement of equipment within separate, completely metallically enclosed cabinets provides the necessary mechanical separation between redundant equipment. 4l Channel Output Monitoring Channel output monitoring is included to provide the operators with early indication of anomalous instrumentation performance. This equipment is not safety related. If the output of one channel differs from either of the 57l redundant channels by more than a preset amount, the channel output monitorina circuitry alarms this condition. 7.2.1.2 Design Basis Information The RSS initiates and carries to completion Reactor, 57l Heat Transport and Balance of Plant Shutdown if any of the off-r.ormal plant conditions listed in Table 7.2-2 occur. The table also shows the frequency classification of the postulated fault, and the first Primary and Secondary 57l RSS subsystems which act to terminate the fault. As detailed in Chapter 15, Amend. 57 Nov. 1980 7.2-6
the RSS design describe below provides the performance necessary to appro-Table 7.2-1 shows the r'D 57l priately limit the results of the postulated events. Primary and Secondary RSS subsystems which use the instrumentation described (,) previously to determine the off-normal conditions and trip the plant. 57 7.2.1.2.1 Primary Reactor Shutdown System Subsystems High Flux 57l The High Flux subsystem (Figure 7.2-5) initiates trip for positive reactivity insertions at or near full power. This subsystem assures that sustained operation will not occur with the fuel near incipient center-line melting. As shown on the figure, the subsystem compares +he compensated ion chamber output signal with a fixed setpoint and initiates trip when the signal exceeds the setpoint. Analysis of the performance of this subsystem is based on worst case time response of the instrument of 50 milliseconds and worst case trip point of 115% of full power. This subsystem is never bypassed. Flux-Delayed Flux The Flux-Delayed Flux subsystem (Figure 7.2-5) initiates trip for rapid sustained reactivity disturbances which occur anywhere in the load rance. Two subsystens are provided; one for positive flux rates and one for negative flux rates. These subsystems prevent undesired thermal transients caused by rapid changes in power with flow held constant. As shown on the figure, the flux signal is compared with the output of a long lac circuit whose input is flux. To initiate trip for increasing reactivity (/ 57 disturbances, the flux signal is a negative input to the ctmparator. For decreasing reactivity disturbances, the flux signal is positive and the out-put of the lag circuit is nt y t.ive. The operation of this subsystem is such that the trip point is dependent on the initial condition, rate of power change, and magnitude of r...er changes. For a given initial power, there is a threshold magnitude of power change to cause a trip. A step change of smaller magnitude than the threshold value will not cause a trip. Power changes greater than the threshold value will initiate a trip with a lower total power change than a slower ramp rate. The trip equatien constants are adjusted to provide the necessary protection for the range of normal power conditions without significar.tly impairing the plant operations. Worst case values of the constan+s, instrument response times and repeatabilities are used in analyzing the performance of the subsystem. The positive flux rate subsystem is never bypassed. The negative flux rate subsystem must be bypassed for plant startup. Nuclear flux is used as a per-missive signal. If nuclear flux is less than 20% of full power flux a bypass can be manually instated. The bypass is automatically removed when power is 57 increased above the permissive level. Flux-Pressure 57 The Flux- Pressure
- subsystem (Figure 7.2-5) initiates trip for positive reactivity excursions or reductions in primary flow over the load range. Two pressure sensors are used for each redundant channel of the system. This arrangement assures appropriate redundancy while providing
(] V effective plant operational characteristics since pressure sensor replacement 57 *Formerly referred to as Flux-Vi>ressure subsystem. Amend. 57 7.2-7 Nov. 1980
cannot be carried out on-line. The use of the high auctioneer automatically accommodates failure of a sensor. All six pressure sensor outputs are com-pared in the channel output monitoring circuitry to provide early indication of anomalous performance to the operating personnel. The subsystem perfor-mance is a function of initial operating level and is analyzed using worst case values for the instrumentation and electronics including response time for the pressure instrumentation of 150 milliseconds. This subsystem must be l bypassed for plant startup. Nuclear flux is used as a permissive signal. If if nuclear flux is less than 103 of full power flux a bypass can be manually instated. The bypass is automatically removed when power is increased above 57 the permissive level. Pump Electrics The Pump Electrics Subsystem (Figure 7.2-6) provides protec-40 tion for loss of pumping power for one, two, or three HTS loops. Three subsystems are included, one for each of the coolant pump buses. In each sub-system, three undervoltage relays, one on each phase, are used as redundant channels. If two of tnree are tripped in any subsystem, reactor trip ensues. A time delay is used to allow the plant to continue through momentary power outages. The subsystem is analyzed using worst case values including 500 mil-lisecond total delay time. For two loop operation, a manual bypass is instated 40l under administrative control by changing the hardware configuration. Two loop by-passes are also under permissive control. Nuclear flux must be less than 10% of full power flux at the time of instating and the primary HTS pump speed in the shut-down loop must be less than 152 of the speed producing nominal all flow or the 57 two loop bypass is automatically removei. l Primary-Intermediate Speed Mismatch 57l tiate trip for imbalances in heat removal capability between the primary and intermediate circuits within a heat transport loop. Three subsystems are included, one for each HTS loop. As shown in the figure, the primary and intermediate speed signals are normalized and subtracted. The absolute value of this difference is compared with a fixed bias and a linear ratio of the primary speed to determine trip initiation. The actual trip point is depen-dent on initial conditions. Worst case values are used for analysis includ-ing a 20 millisecond tachometer time constant. These subsystems must be bypassed to start the plant. The permissive signal used is the nuclear flux. If the nuclear flux is less than 10$ full power flux, the subsystem can be bypassed manually. The bypass for the shutdown loop is automatically removed as power is increased. For two loop operation, provisions are made to bypass the function associated with the shutdown loop. A manual bypass is instated under admininstrative control by changing the hardware configuration. Two loop bypasses are also under permissive control. Nuclear flux must be less than 10% of full power flux at the time of instating and the primary HTS pump speed in the shutdown loop must be less than 15% of the speed producing nominal full flow or the u. 57 loop bypass is automatically removed. Reactor Vessel Level The Reactor Vessel Level subsystem (Figure 7.2-5) prevents reactor operation unless the sodium level in the reactor vessel is at least 6 inches above the suppressor plate. The output of the level sensor is compared with a fixed setpoint to determine the need for a reactor trip. Worst case values are used in the analysis of the performance of this subsystem including a sensor time constant of 0.5 second. This subsystem is never bypassed. Amend. 57 7.2-8 Nov. 1980
Steam-Feed Flow Mismatch 57l The Steam-Feed Flow Mismatch subsystem (Figure 7.2-7) initiates reac-tor trip to prevent continued operation with large imbalances between the steam and feedwater flow for each HTS loop. One of these subsystems is
\.s included in each HTS loop. These subsystems protect the steam generators and drums against unacceptable thermal transients. As shown in the figure, each subsystem compares the steam flow and feedwater flow, both of which are mul-tiplied by appropriate constants, in two individual comparators. If the difference between the two values exceeds the setpoint in either of the com-parators, a trip is initiated. Increasing steam flow and decreasing feed-water flow fault events are sensed by the first comparator. The second com-parator senses decreasing steam flow and increasing feedwater flow fault events. Analysis of this function is based upon worst case parameter values.
This subsystem must be bypassed for plant startup. A permissive is included which allows manual bypass of this subsystem for nuclear power less than lot. Two loop bypass provisions are also included for the shutdown loop. Two loop bypasses are also under permissive control. Nuclear flux must be less than 10% of full power flux at the time of instating and the primary HTS pump speed in the shutdown loop must be less than 15% of the speed producing 57 nominal full flow or the two loop bypass is automatically removed. IHX Primary Outlet Temperature The IHX Primary Outlet Temperature subsystem (Figure 7.2-7) compares 4 the sodium temperature in the primary cold leg of each IHX to a fixed set point. A reactor trip is initiated if the sodium temperature exceeds this set point. These subsystems assure that temperature increases in an inter-mediate loop sodium resulting from steam side fault events or intermediate O flow reductions do not increase the reactor coolant temperature. There is C/ one IHX primary outlet temperature subiystem per HTS loop. are never bypassed. These subsystems 7.2.1.2.2 Secondary Reactor Shutdown System Subsystems 57l I Modified Nuclear Rate i The Modified Nuclear Rate subystems (Figure 7.2-8) initiate trip for rapid sustained reactivity disturbances which occur in the load range. Two sub-systems are provided. One for positive flux rates and one for negative flux rates. These subsystems prevent undesired thennal transients caused by rapid changes in power with flow held constant. The reactor trip is based on flux rate measurements from the fission counters. A pennissive is included which allows manual bypass of the negative rate subsystem for nuclear power 57 less than 10% The positive rate subsystem is never bypassed. Flux-Total Flow The Flux-Total Flcw Subsystem (Figure 7.2-8) provides protection against increasing and decreasing flow and power events over the 40 to 100% load range. The primary flows of the three HTS loops are summed and multi-plied by an appropriate gain. A nuclear power signal obtained from the fis-sion counter is subtracted in the comparator from the total flow value and this difference is compared to a fixed set point. If the difference exceeds the set point, then a reactor trip is initiated. Analysis of this subsystem
.) is based on worst case parameter values, including a 500 nsec. time delay for the flow detectors. This subsystem is n m bypassed.
Amend. 57 7 2-9 Nov. 1980
Startup Nuclear The Startup Nuclear subsystem (Figure 7.2-8) obtains a wide range 109 57 channel measurement of nuclear power from the fission counters and compares it to a fixed-set point. If nuclear power is greater than the set point, a reactor trip is initiated. A permissive module is provided which allows manual bypass of this subsystem upon the verification of the operation of 57l tne wide range linear channel. This subsystem provides protection against positive reactivity disturbances occurring during startu;. Primary to Intermediate Flow Ratio The Primary to Intermediate Flow Ratio subsystems (Figure 7.2-8) pro-tect against an imbalance in the heat removal capability of the primary and intermediate loops. The heat removal capability of a particular loop is determined by measurement of the sodium flow within the loop. The Secondary 57l Flow Low.RSS includes In the Primarytwo of High Flow thesesubsystem, subsystems, thePrimary Flow output of the High high and Primary primary flow auctioneer is compared to the summation of the outputs from the low intermediate flow auctioreer and a signal proportional to the total primary flow. When the high primary flow auctioneer signal exceeds the low intermediate flow auctioneer signal by an amount proportional to the total primary flow, a reactor trip is initiated. Similarly in the Primary Flow Low subsystem, a comparison is made between low primary flow and high intermediate flow. When the high intermediate flow auctioneer signal exceeds the low primary flow auctioneer signal by an amount proportional to the total primary flow, a reactor trip is initiated. These subsystems are manually bypassed during plant startup. 41 The permissive signal used is based on reactor power. If reactor power is less than 10%, the subsystems can be manually bypassed. Steam Drum Level 57l The Stea.11 Drum Level subsystem (Figure 7.2-9) measures steam drum water level and compares it to a fixed setpoint. A reactor trip is initiated whenever the drum water level decreases below this fixed setpoint. There are three of these subsystems, one per HTS loop. Analysis of these sub-systems are based upon worst case perimeter values. For two loop operation, a manual bypass is instated under administrative control by changing the hardware configuration. Two loop bypasses are also under permissive control. Nuclear flux must be less than 10% of full power flux at the time of instating and the primary flow in the shutdown loop must be less than 15% of full flow or the two loop bypass is automatically removed. 7.2-10 Amend. 57 Nov. 1980
l Evaporator Outlet Sodium Temperature O The Evaporator Outlet Sodium Temperature subsystems (Figure 7.2-10) compare the sodium temperature at the outlet of the evaporator in each HTS loop to a fixed set point. If this temperature exceeds the set point, a reactor trip is initiated. There are three of these subsystems, one per loop. These subsystems detect a large class of events which impair the heat removal capability of the steam generators. These subsystems are never bypassed. Sodium Water Reaction The Sodium Water Reaction subsystems (Figure 7.2-10) detect the occurrence of a sodium water reaction within a superheater or evaporator module. There are three of these subsystems, one per loop. Each subsystem 571 receives nine signals from the sensors in the reaction products vent lines of a steam generator. These subsystems are never bypassed. 7.2.1.2.3 Essential Performance Requirements In order to implement the required protective functions within the appropriate limits, PPS equipment must meet several essential performance requirements. These essential performance requirements and the PPS equip-ment to which they apply are sunmarized below. The PPS instrumentation will meet the essential performance require-O 57l ments of Table 7.2-3. This table defines the minimum accuracy and time constants which will result in acceptable performance of the PPS. O [ Analysis of worst case PPS functional performance is based on the values given in Table 7.2-3. The maximum delay between the time a protective subsystem indicates the need for a trip and the time the rods are released is 0.200 second. .) l This time includes the delays due to the calculational units, comparators, logic, y scram breakers, and control rod release. The maximum delay between the time a protective subsystem indicates the need for a trip and the time the HTS sodium pumps are tripped is 0.500 second. This time also includes the delays dae to the logic and HTS scram breakers. The PPS is designed to meet these essential performance require-ments over a wide range of environmental conditions and credible single events to assure that environmental effects do not degrade the performance Amend. 57 7.2-11 Nov. 1980
of the pFS. The environmental extremes are documented in Reference 13 of 57 PSAR Section 1.6. Provisions are incorporated within the PPS which provide a defense against the following incidents: e Environmental Changes All electrical equipment is subject to performance degradation due to major changes in the operating environment. Where practical, PPS equipment is designed to minimize the effects of environmental changes; if not, the performance at the environmental ext. emes is used in the analysis. Measures have been taken to assure that the PPS electronics are capable of performing according to their essential performance requirements under variations of temperature. The range of tem-perature environment specified for all the electronic equipment considered here is g eater than is expected to occur during normal or abnormal conditions. Electronics do not fail catastrophically when these limits are exceeded even though this is the assumed failure mode. The detailed design of the circuit boards, board mounting and racks includes free ventilation to minimize hot 57l spots. Ventilation is a result of natural convection air flow. The PPS is designed to operate under or be protected from a wider range of relative humidity than that produced by normal or postu-lated accident conditions. Vibration and shock are potential causes of failure in electronic components. Design measures, including the prudent location of equipment, minimize the vibration and shock experienced by PPS electronics. The equipment is qualified to shock and vibration specifications which exceed all normal and off-normal occurrences. The PPS comparators and protective logic are designed to operate over a power source voltage range of 108 to 132 VAC and a power source frequency range of 57 to 63 HZ. The maximum variation of the source vc tage is expected to be 10% More extreme varia-tions in the power source may result in the affected channel comparator or logic train outputting a trip signal. In addition, testing and monitoring of PPS equipment is used, where appropriate, to warn of impending equipment degradation. Therefore, it is not expected that changes in the environment will cause total failure of an instrument channel or logic train, much less the simultaneous failure of all instrument channels or logic trains. The majority of the PPS electr inics is located in the control building, and is not subjected to a radioactive environment. Any PPS equipment located in the radioactive areas (such as the head access area) will be designed to withstand the level of activity to which it will be subjected, if its function is required. O 7.2-12 Amend. 57 Nov. 1980
e Tornado 7 !") The PPS is protected from the effects of the design basis tornado by locating the equipment within tornado hardened structures. e Local Fires All PPS equipment, including sensors, actuators, signal condi-tioning equipment, wiring, scram breakers, and cabinets housing this equipment is redundant and separated. These characteristics make any credible fire of no consequence to the safety of the plant. The separation of the redundant components increases the time required for fire to cause extensive damage and also allows time for the fire to be brought to the attention of the operator such that corrective action may be initiated. Fire protection systems are also provided as discussed in Section 9.13. e Local Explosions and Missiles All PPS equipment essential for reactor trip is redundant. Physical separation (distance or mechanical barriers) and elec-trica: isolation exists between redundant components. This phys-ical separation of redundant components minimizes the possibility of a local explosion or missile damaging more than one redundant component. Tne remaining redundant components are still capable of performing the required protective functions.
) e Earthquakes All PPS equipment, including sensors, actuators, signal condition-ing equipment, wiring, scram breakers and structures (e.g. ,
cabinets) housing such equipment, is classed as Seismic Category I. As such, all PPS equipment is designed to remain functional under OBE and SSE conditions. The characteristics of the OBE and SSE used for the evaluation of the PPS are found in Section 3.7. 7.2.2 Analysis The Plant Protection System meets the safety related channel per-formance and reliability requirements of the f4RC General Design Criteria, 57 RDT Standard C16-lT, IEEE Standard 279-1971, applicable f4RC Regulatory Guides and other appropriate criteria and standards. General Functional Requirement The Plant Protection System is designed to automatically initiate appropriate protective action to prevent unacceptable plant or component damage or the release or spread of radioactive materials. l (,) Amend. 57 7.2-13 flov. 1980 l l l
l l l Single Failure ; No single failure within the Plant Protection System nor removal from service of any component or channel will prevent protective action when requi red. 57l Two independent, diverse reactor shutdown systens are provided, either of which is capable of terminating all excursions without allowing plant param-eters to exceed specified limits. Each system uses three redunda.nt instru-ment channels and logic trains. The Primary RSS is configured 57 using local coincidence logic while the Secondary RSS uses general coincidence logic. To provide further assurance against potential degradation of protection due to credible single events, functional and/or equipment diversity are included in the hardware design. Bypasses Bypasses for normal operation require manual instating. Bypasses will be automatically removed whenever the subsystem is needed to provide protection. The equipment used to provide this action is part of the PPS. Administrative procedures are used to assure correct uw of bypasses for infrequent operations such as two loop operation. If the protective action of some part of the system has been bypassed or deliberately rendered inoperative, this fact will be continuously indicated in the control room. Multiple Setpoints Where it is necessary to change to a more restrictive setpoint to provide adequate protection for a particular normal mode of operation or set of operating conditions, the PPS design will provide automatic means of assuring that the more restrictive setpoint is used. Administrative proce-dures assure proper setpoints for infrequent operations. For CRERP, power operation on two-loops will be an infrequent occurrence, and will only be initiated from a shutdown condition. While the reactor is shutdown, the PPS equipment will be aligned for two-loop operation which will include set down of the appropriate trip points. Sufficient trip point set down is being designed into the PPS equipnent to adequately cover the possible range (conceptually from 2'l to 100%) of trip point adjustment required. In addition, administrative procedures (specifically the pre-critical checkoff) will te invoked during startup to ensure that the proper PPS trip points have been set. The analysis of plant performance during two-loop operation has not been completed to date. Therefore, the exact trip point settings for two-loop operation cannot be specified at this time. However, the range of trip point settings indicated above is adequate to ensure that trip points appropriate for the anticipated lowest two-loop operating power can be achieved. In sumnary, the design of the PPS equipment trip point adjustments and other features for two-loop operation coupled with the anticipated two-loop operating power level and administrative procedures assure full compliance with Branch Technical Position EICSB 12 and satisfy Section 4.15 of IEEE std 279-1971. .16 7.2-14 Amend. 57 Nov. 1980
Completion of Protective Action 57l The Reactor Shutdown Systems are designed so that, once initiated, a protective action at the system level must go to completion. Return to normal operation requires manual reset by the operator because the Primary RSS 57 scram breakers or Secondary scram latch circuitry must be manually closed following trip. Trip signals must be cleared prior to closure of scram breakers. Manual Initiation The Plant Protection System includes means for manual initiation of each protective action at the system level with no single failure preventing initiation of the protective action. Manual initiation depends upcn the operation of a minimum of equipment because the manual trip directly operates the scram breakers, solenoid scram valve power supply, or equivalent for Shutdown Heat Removal and Containment Isolation System. [') 7.2-14a Amend. 57 Nov. 1980
Access D (V Administrative control of access to all setpoint adjustments, module calibration adjustments, test points and the means for establishing a bypass permissive condition is provided by locking cabinets and other access design 57 features of the control room and the equipment racks. Information Read-Out The trip or reset status of all PPS comparators, 57l logic channels, and power interrupting devices (scram breakers, primary coolant pump breakers, etc. ) is displayed on the Main Control Board. Annunciator for PPS Alarm Trips A visual and audible indication of all alarm conditions within the PPS will be provided in the control room. These alarm conditions include any tripped PPS comparators in the Primary RSS, Secondary RRS, Containment Iso-
~
lation System and Shutdown Heat Removal System. The Plant Data Handlina and Display system alerts the operator to significant deviations between redundant 57 RSS analog instrumentation used to monitor a reactor or plant parameter for the RSS. Control and Protection Sy_ stem Interaction The Plant Protection System and the Plant Control System have been designed to assure stable reactor plant operation and to protect the reactor p plant in the event of worst case postulated Plant Control System failures. i The Plant Protection System is designed to protect the plant regardless of control system action or lack of action. Isolation devices will be used between protection and control functions. Wnere this is done, all equipment common to both the protection and control function is classified as part of the Plant Protection System. Equipment sharing between protection and control is minimized. Where practical, separate equipment (sensors, signal conditioning, cabling penetrations, raceways, cabinets, monitoring etc. ) is provided. The sharing of components does not lead to a situation where a single event both initiates an incident through Plant Control System mal-function and prevents the appropriate Plant Protection System. Periodic Testing The Plant Protection System is designed to permit periodic testing of its functioning including actuation devices during reactor operation. In 57l the Primary RSS, a single instrument channel is tested by insert-ing a test signal at the sensor transmitter and verifying it at the compara-tor output. A logic train is tested by inserting a very short test signal in 2 comparator inputs and verifying that the voltage on the scram breaker trip coils decrease. Because of the time response of the undervoltage relay coils of the scram breakers and very short duration of the test signal, the 571 reactor does not trip. In the Secondary RSS, an instrument /~N U Amend. 57 7.2-15 Nov. 1980
channel can be tested from sensor to scram actuator by inserting a single test signal because of the general coincidence configuration of the 3 redundant channels. The primary and secondary rod actuators cannot be tested during reactor operation since dropping a single control rod will initiate a reactor scram. Scram actuators and control rod drop will be tested and maintained when the plant is shutdown (See Section 7.1-2). When-ever the ability of a protective channel to respond to an accident signal is bypassed such as for testing or maintenance, the channel being tested is placed in the tripped state and its trippec condition is automatically indi-cated in the control room. Failure Modes and Effects Analysis A Failure Modes and Effects Analysis (FMEA) has been conducted to identify, analyze and document the possible failure modes within the Reactor Shutdown System and the effects of such failures on system 40 performance (see Appendix C, Supplement 1). Components of the RSS 41 57 analyzed are: e Reactor Vessel Sodium Level Input e PPS Sodium Flow Input 40 e Pump Electric Power Sersor e Compensated '~ Chamber fluclear Input e Fission Chamber fluclear Input e Primary Loop Inlet Plenum Pressure Input e Sodium Pump Speed (Primary and Intermediate) e Steam Mass Flow Rate Input e Feedwater flass Flow Rate Input e Steam Drum Level Input e Primary Compara tor e Secondary Comparator e Primary Logic Train e Secondary Logic Train e Primary Calculational Unit e Secondary Calculational Uni t Amend. 57 flov. 1980 7.2-16
m (j e Scram Actuator Logic e Heat Transport System (HTS) Shutdown Logic e Control Rod Drive Mechanism (CRDM) Power Train 57 e PPS Isolation Buffer Figures 7.2-3 and 7.2-4 provide assistance in locating the above 57j system level components within the overall RSS. The probability of occurrence of each failure mode is listed 57l in the tables of Appendix C. Supplement 1, in the Probability Column.
- 4) The effects of each potential failure mode have also been categorized in the tables in the Criticality Column. Even though the failure of an individual elerent may result in the inability to initiate channel trip, the pro-vision of redundant independent instrument channels and logic trains assures that single random failures cannot cause loss of either the Primary or 57l Secondary RSS thereby meeting the design requirements of RDT C16-1T and IEEE 279-1971. The high reliability of components, redundant configu' ration, provision for on-line monitoring and on-line periodic testing further assure that random failures will not accumulate to the point that trip 57l initiation by either Primary or Secondary RSS is prevented. All (p') failure effects are therefore categorized as not causing any degradation
'> 41 or failure of a system safety function. The majority of the identified failure modes can be eliminated from consideration based on their low probability of occurrence and the insignificance of their criticality. They are included in the Ff1EA, however, to document their consideration. n Amend. 57 ( ) 7.2-17 Nov. 1980
TABLE 7.2-1 PLANT PROTECTION SYSTEM PROTECTIVE FUNCTIONS Primary Reactor Shutdown System e Flux-Delayed Flux (Positive and Negative) l 57 e Flux Pressure
; e High Flux 57l e Primary to Intermediate Speed Mismatch e Pump Electrics e Reactor Vessel Level e Steam-Feedwater Flow Mismatch e IHX Primary Outlet Temperature Secondary Re;ctor Shutdown System 57 e M dified Nuclear Rate (Positive and Negative) e Flux-Total Flow e Startup Nuclear O e Primary to Intermediate Flow Ratio e Steam Drum Level e Evaporator Outlet Sodium Temperature i
47 ' e Sodium Water Reaction Amend. 57 Hov. 1980 7.2-18 I
(~\ ( [ G \J L TABLE 7.2-2 PPS DESIGN BASIS FAULT EVENTS Fault Events Primary Reactor Shutdown Systen Secogdgy_ Reactor Sh_utdown System 57l I. Anticipated Faults A. Reactivity Disturbances b) 53 l Positive Ramps $5c/sec and Steps $10c Startup Flux-Delayed Flux or Startup Nuclear Flux Pressure 5-40% Power Flux-Delayed Flux or Modified Nuclear Rate or Flux Pressure Flux-Total Flow 57 40-100% Power Flux Pressure Flux-Total Flow Full Power High Flux Flux-Total Flow Negative Ramps and Steps Flux-Delayed Flux Modified Nuclear Rate B. Sodium Flow Disturbances Coastdown of a Single Primary or Primary-Intermediate Primary-Intermediate 57l Intermediate Pump Speed Mismatch Flow Ratio 2 Loss of 1 HTS Loop Pump Electrics P.-imary-Intermediate Sk Flow Ratio EL g- Loss of 2 HTS Loops Pump Electrics Primary-Intermediate 8 F] Flow Ratio Loss of 3 HTS Loops Pump Electrics Flux-Totai Flow
O O O TABLE 7.2-2 (Continued) F uit Events Primary Reactor Shutdown System Secondary Reactor Snutdown System 57 _ C. Steam Side Disturbances Evaporator Module Isolation Valve IHX Primary Outlet Evaporator Outlet Na Closure Temperature Temperature Superheater Module Isolation Valve Steam-Feedwater Flow Evaporator Outlet Na Closure Mismatch Temperature Water Side Isolation and Dump IHX Primary Outlet Evaporator Outlet Na of Single Evaporator Temperature Temperature Water Side Isolation and Dump Steam-Feedwater Flow Evaporator Outlet Na of Single Superheater Mismatch Temperature N Water Side Isolation and Dump of Steam-Feedwater Flow Evaporator Outlet Na [ Both Evaporators and Superheater Mismatch Temperature o Loss of Normal Feedwater Steam-Feedwater Flow Steam Drum Level Mismatch 47l Turbine Trip with Reactor Trip Steam-Feedwater Flow Steam Drum Level (Loss of Main Condenser or Mismatch Similar Problem) Inadvertent Opening of Evaporator Steam-Feedwater Flow Steam Drum Level Outlet Safety Valve Mismatch Inadvertent Opening of Superheater Steam-Feedwater Flow Steam Drum Level Outlet Safety Valve Mismatch oE Inadvertent Opening of Evaporator IHX Primary Outlet ~vaporator Outlet Na
-{ Inlet Dump Valve Temperature Temperature G'
8t3
TABLE 7.2-2 (Continued) 57l Fault Events Primary Reactor Shutdown System Seconjary Reactor Shutdown System II. Unlikely Faults A. Reactivity Disturbances ( 53 Positive Ramps <35c/sec and Steps <60c Startup Flux-Delayed Flux or Startup Nuclear 57l F1 :v Pressure 5-40I Power Flux-Delayed Flux or Modified Nuclear Rate or 57! Flux-Total Flow Flux Pressure y 40-100% Power Flux-Pressure Flux-Total Flow c'o g Full Power High Flux Flux-Total Flow B. Sodium Flow Disturbances Primary Pump Seizure Primary-Intermediate Primary-Intermediate Flow 57I Speed Mismatch Ratio Intermediate Pump Seizure Prima ry-Intermedia te Primary-Intermediate Flow 57l Speed Mismatch Ra tio C. Steam Side Disturbances ( Steam Line Break Steam-Feedwater Flow Evaporator Outlet Na E 3" Mismatch Terrpera ture f$ -." Recirculation Line Break Steam-Feedwater Flow Steam Drum Level $m ou Mismatch Feedwater Line Break Steam-Feedwater Flow Steam Drum Level Mismatch e O O
O O O TABLE 7.2-2 (Continued) 57l Fault Events Primary Reactor Shutdown System A tondary Reactor Shutdown System Failure of Steam Dump System Steam-Feedwater Flow Steam Drum Level Mismatch Sodium Water Reaction in Steam (3) Steam-Feedwater Flow Sodium-Water Reaction Generator Mismatch III. Extremely Unlikely A. Reactivity Disturbances Positive Ramps < $2.0/sec 53 s Startup Flux-Delayed Flux Startup Nuclear h 5-40% Power Flux-Delayed Flux or Modified Nuclear Rate or M Flux-Pressure Flux-Total Flow 57 40-100% Power Flux-Pressure Flux-Total Flow Full Power High Flux Flux-Total Flow (1) The maximum anticipated reactivity fault results from a sin 53 maximuminsertionrateofapproximately4.1centspersecon{efailureofthecontrolsystemwitha , (2) The maximum unlikely reactivity faults result from multiple control, system failures leading to with-drawal of six rods at normal speed or one rod at the maximum mechanical speed. 53l gg (3) The PPS is required to terminate the results of these extremely unlikely events within the umbrella F@ transient specified as emergency for the design o, the major components. F
TABLE 7.2-3 i ESSENTIAL PERFORMANCE REQUIREMENTS FOR PPS INSTRUMENTATION ' ACCJracy Response Time 57l Plant Parameter (%ofspan)__ (msec) Neutron Flux l Primary 21.0 <10 Secondary tl.0 <10 1 Reactor Inlet Plenum Pressure t2.0 <150 l Sodium HTS Pump Speeds 2.0 <20 l Sodium HTS Flow 10.0 <500 ! Reactor Vessel Sodium Level t 5. 0 <500 Undervoltage Relay 3.0 <200 i Steam F1ow r 2. 0 < 500 Feedwater F1ow 12. 5 < 500 Evaporator Outlet Sodium Temperature !2.0 <5000 57 Steam Drum Level + 1. 0 <1000 IHX Primary Outlet Temperature 2.0 <5000 47 l l i i 4 l Amend. 57 Nov. 1980
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a_ 4558-10 Amend. 57 7.2-25 flov. 1980
.- - - .. . - - . . - - . _ - - - - - - - . - _ .-- .- - - - - - ~ . . - . _ - - +.. . - - -n. -.
- . - _ ..- - -.+- . . _ - -
e u t.se (#a i t SE E 40 f t I A U40Env0ttAGE RELAT CHAT l4Et 3 - a 100Pt y SE E N0 f f 1 h .' I A 2 ($t t IGURE 7 2 h C0tePt41AIE0 t04 CMAMGE R CHAbin[? C L OGaC V ARIAtti IIME Of t Av z 5f t 40 f t 1 E E L E CT R04sC1 M MA4UAL ACTU AT104 CHA44EL A , TO LOGIC j TR Att 1 Rit AV $ witch CE5 SEE NOTE I f 10 LOGar 1*O OUT Of IMREE TR4P 14 PUTS RE00fREO LOOP 2 tsEE p3GURE 7 2 2BI y . 10 PRODUCE A 'R9 OUTPUT , CHA44El C Sit hof f 1 30f M T RIP !4PUT1 R10Uant D A 0 TO LOGeC TO PRODUCE A TRt* Ou!PUT, i
. 20 TRAtt 3 Ts)
IRIP COMPARATOR _i
, Sif 40if 1 G
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$f t 401f HA44Et 8 e CM A44t t IS0t AT104
[ 04 LOOP 2 CHA44Et C y RIFERE4CE h t$lt StGURE 7 2 281
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] " j TO 404 PPS d suf f f Rh A40 40TE i E ACM CH A44E t nAs sis 0W4 PE RMissivE COMPARAIOR T YPIC A PE RMIS$1VE It ] h ,, sH0a4 04 t007 3 CmA44f a C } r 10 f L UI COMPAR ATOR$
- n 5N
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ou Figure 7.2-2 A. Typical Primary I'PS Instrument Channel Logic I)iagram > j (Pump I!!ectrics Subsystem Simwn)
4. T-4. r READY TO -=-- B YP ASS EYFAS5 INDICATION IN DIC ATIO N BYPASS PERMISSIVE CONDITIONS SATISFIED I AND
" ^"^ " '
BYPASSED N ro 5a MANUAL BYPASS ' SWITCH ACTIVATED OR BYPASS SEAL-IN EN 58" U Figure 7.2-2 A A. RSS Ilypass Function Illock Diagram O O O
O O
& LOGIC TRAIN 2 "f
(SEE FIG. 7.2-2A) DOR tJ EITHER TRIP INPUT LOGIC TRAIN 3 M OR PRODUCES A TRIP
~'" " OR (SEE FIG. 7.2-2A)
RE ACTOR VESSEL LEVEL V MANUAL HIGH FLUX ,, l M ACTUATION PRIMARY TO LOOP 2 LOOP 3 DR POWER INPUT 34 - U 0 VOLTAGE SPEED C z PUMP ELECTRICS (SEE FIGURE 7.2 2A;
)f LOOP I
LOOP 2 1 - UV l UV LOOP 3 . I( k C0tt NC NC / COIL FLUX - DELAYE0 FLUX l - UV y SPARE LOOP 1 7 STE AM - F E E0 WATER 10CP 2 l UV ' NC NC}- UV C0ll y FLOW TAISM ATCH LOOP 3 OR COIL LOOP 1 IHX PRIMARY LOOP 2 C OUTLET TEMPERATURE LOOP 3 7 POWER TO 9 PRIMARY RODS HTS TRIP INITIATION SPARES < i, TO HTS BREAKERS NOTE 1: PERMIS$1VES AND BYPASSES ARE (SEE FIGURE 7.2-2El ACCOMPLISHE0 IN THE INSTRUMENT (SEE FIGURE 7.2 2A FOR TYPICAL PPS INSTRUMENT oN CetANNEL LOGIC OIAGRAM.) f$ IE m Figure 7.2-211. Primary RSS Final Actuation Logic Diagram
T U S P N T T O O U U P R R I N O O O T I T T A N N E C H A R U O A V W E T C R O T C W T E O I O E T A E C T S S H I W P V TA A S U S D M I S O I S R L E D T N MT N S A R O R N O Y C I A U I R E E E A MP U R I S L P I R M N Q P R P M S E R E O A E O U N E FI R T P C M R T C W G O N A O S R N M T I R A
-- S A L H A L C .
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m _ . . - . _ ..__.__..mm. . __m_ ._ _ . _ _ . . . _ . ___ . _..._ .- _. 1 _ _ . . _ _ _ . _ . . _ . . _ , . ... .. _ . _ . . _ _ . . _ . .. I 4 i Y ,
= A ST ARTOP NurLE AR stagg 1/3 FL Xi TOTAL 0 TOR 003 R0 CONTRot TOR 004 SODIUM WATER RE Acit0N d ---l 850 .10 R00 6 I
t PRIMARY TO O QS TOA00$ p 84TE RME DI ATE FLOW l g Raft 0 9 q J SCRAM SEC040 ARV R001 STE AM ORUM LEVEL _f h LATCH ] 150 TOR 002 VALVf CONTROL d ISO YOR003 i ,N EwAPORATOR DUTLET AI P 0017 y 5001UM TEMPEll ATURE TO R00 5 M 150 I
' $0LE N080 0 RIVE R
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@ AIR PRISSURE SUPPLY M ISO YOR0D4 TO HTS GRE AKERS - 65 0 To ROO 5 (RFFER TO FIG. T2 21 2/3
@ ISOL Afl04 0E VICE A006 - TWO OUT OF THREE TRIP thPUTS 1 g
. RioulRED TO PRODUCE A TRIP OUTPUT NOTE 1. TYPICAL Of TWO REDUNOA4T h
< fo HTS PutIP TRIP LOGIC CIRCUITS a *3 CL G'
Ca en Figure 7.2-2D. Secondary RSS Final Actuation Logic Diagram ) _ _ _ _ . . _ . _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ - _ . - - - . m
SCRAM BREAKER g CH ANN EL TEST INPUT GIC
$ CHANNEL TEST OUTPUT TESTINPUT g $ LOGIC TEST DUTPUT U 3 7 E.LY SENSOR TRANSMITTER CALC. -
N ~ ~ """ - "^ m COMPARATO . EAKER J - LOGIC DRIVER TR ANSMITTE R BUFFER .- :- 1/24 4 . SCRAM BREAKER PTICAL 2/3 LOGIC TEST OUTPUT SE E R TRMITT ER OBUFFER COUPLING D - AL-V - - - CALC.
~ ~ " -
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% MPARATOR
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3E EAKER T COMPARATO _ 5$ TRA TER BUFFER 1/24 2/3 LOGIC g, OPTICAL COUPLING oN Figure 7.2-3. Typical Primary RSS Sutnystem O O O
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TO STEAM FEEDCATER FLOO (FIGURE 7.2-7) TO PRIMARY INTERMEDIATE SPEED (FlGURE 7.2-6) TO NON-PPS + BUFFER - PERMISSIVE TO PUMP ELECTRICS MODULE (FIGURE 7.2-6) COMPENSATED 10N ELECTRONICS -+ Kg CHAMBER
- j + 7,5 FLUX-0ELAYED FLUX
[ POSITIVE K O 1+TS O FLUX DELAYE0 FLUX NEGATIVE F BYPASS _ CIRCUlT HIGH GH FLUX TO NON-PPS
- BUFFER -
PRESSURE FLUX-ELECTRONICS
- a +
P PRESSURE AUCTIONEER HIGH f _ Kp ["" ELECTRONICS * ~ OR RCUIT TO N O N-PPS -*-- BUFFER - TO NON-PPS
- BUFFER <
A TOR LEVEL DETECTOR
- ELECTRONICS - --
k VESSEL LEVEL L Figure 7.2-5. Functional 131ock Diagrams of the Flux-Delayed Flux, liigh Flux. Flux-Pressure, and Reactor Vessel Lesel Protectise Subsystems. One Channel of Three Shown 4558-3 Amend. 57 7.2-33 t.!ov. 1980
. - . ~ _ - - - . . _ _ . - - . . . _ _ _ . . _ . . _ _ - _ _ _ _ _
UNDERV0LTAGE VARIABLE RELAY LOOP 1 TIME DELAY ~,
._, +
0 BYPASS CIRCulTRY UNE iRVO LTAGE VARIABLE Rt. V LOOP 2 TIME DELAY gS i _ + 8YPASS , Cif*CUlTRY UNDERVOLTAGE VARIABLE _
; RELAY LOOP 3 TIME DELAY +
,-( S +
l _ ) j 8YPASS , CIRCUITRY TO NON-PPS r
~
PRIM ARY TACHO-j METER LOOP 1 - l BUFFER K - [ p CIRCUlTRY INTERMEDIATE i ELECTRONICS HP ,_,_ TACHOMETER + ABSOLUTE _ LOOP 1 VALUE
] ELECTRONICS l -
TO NON-PPS m j TO NON-PPS
- l BUFFER H +
J. PRIMARY TACHO
- METER LOOP 2 1 BUFFER l K -
BYPAC , CIRCUlTRY l ELECTRONICS l '
+ ABSOLUTE INTERMEDIATE ~
TACHOMETER VALUE l ELECTRONICS l I LOOP 2 TO NON-PPS 9 BUFFER l _
+
I TO NON-PPS 9 + j l BUFFER K - PRIMARY TACHO- BYPASS METER LOOP 3 CIRCulTRY n M ELECTRONICS l + ABSOLUTE _ INTERMEDIATE - - ^ TACHOMETER ELECTRONICS l PERM N LOOP 3 I ~
~
TO NC'. PPS Figure 7.2-6. Functional Block Diagrams Of The Pump Electrics And Primary To Intermediate Speed Ratio Protective Subsystems. One Channel Of Three Is Shown 4558-6 Amend. 57 7.2-34 Nov. 1980 i _ . -_ _. , . _ . , _ . _ . . - . _ . . . , _ . _ . - ~ . _ - . . _ , . . .
TO NON-PPS
- BUFFER TEMPER AYURE
- ~
DETECVOR ELECTRO N ICS LOOP al + TO NON-PPS
- BUFFER TEMPERATURE DETECTOR ELECTRONICS -
LOOP *2
- l TO NON-PPS
- BurFER IEMPERATURE DETECTOR ELECTRONICS LOOP m3 +
NOTE: TEMPERATURE AND PRESSURE
' COMPENSATION SENSORS AND TO NON-PPS E.ECTRONICS OMITTED FOR ST EAM LARITY FLOWNETER g ~
BUFFEq - - SA LOOP al - p g- CIRCulTRY FEEDWATER - ELECTRONICS WD -
"FA FLOWNETER LOOP el - qIB SYPASS TO NON-PPS ~
ELECTRONICS
~
i lL+ ~ CIPCUtTRY *
~
OU 0 TO NON-PPS
- STEAM FLOWNETER 1 BUFFER h- K SA -
LOOP =2
+
ELECTRONICS HND - FA r- CIRCUlTRY FROM FEEDWATEP
-. PERMISSIVE FLOWNETER LOOP +2 g
pg L + BYPASS + (FIGURE l CIRCulTRY 7.2-5) TO NON-PPS
- BUFFEp SB
+
TO NON-PPS : , S ', E A M - - K BUFFER SA - FLOWNETER + BYPASS LOOP W3 ELECTRONICS HM -
"FA -
FEEDWATER - 0 33
& K FLOWNETER ELECTRONICS FB L_ , CIRCU TRY LOOP =3 -
- E ^
BUFFER SB TO NON-PPS * "" Figure 7.2-7. Functional Block Diagrams of the IHX Primary Outlet Temperature and Steam to Feedwater Flow Mismatch Protective Subsystem, One Channel of Three is Shown 6678-3 7.2-35 Amend. 57 Nov. 1980
TO NON PPS
- SUFFE2 MP t c' R
- E LECTRONeCS Ml FLUM TOTA L G pp l , FLOW LOG lt pp CHANNEL
{ ,g 47ARTUP NUCLEAR LINE Am EY'"II -%i K,5 _ IPOSITIVEl (S.ef p 2 MODeFIE D SE TPOIN T 1 NUCL E AR R ATE , SE TPOINT 2 -
, y d- + (NEGATIVE)
F-b BVPASS CIRCUITRY ggy a y E L E C T RON sCS l PHeMARY , AUCTIONE E R TO NON-PPS SUFFER h HrGH ,
'it s - -
, Low PRIM ARV FLOW- +
METER LOOP NO 2 H E LECTRONICS h pi==
+
- +
TO NON-PPS H BUFFER H PR sM AR Y FLOW ,PL
~ """
AUCTIONE E R PR Rf AR Y F LOW ~ E L E CT RONICS LOW METER LOOP NO 3 ' TO NON-PPS q SUFFER l--- - sNI E RMEDea IE F T FLOWME TE R ELECTRONICS hW i LOOP 40 9 , TO NON-PPS % BUF F E R h --- K 3
" * ~
sN T E RME Ds A T E j F LOWME TE R E LECT RON6CS hg tN T E RME DI AT E L OOP NO 2 l FLOW AUCitONE E R TO NON-PPS % BUFFER M8CM ,
#N TE RME Dd ATE ,
,_ LOW F LOWME TE R d ELECTRO 4sCS h PM' 1
tOOP O i ,m I- _ TO NON -PPS H euFFER h INT E RME DI A T E FLOW " AUCT TONE E R LOW JVP ASS CIRCueTRY BVPASS Cercus T R Y I Figure 7.2-8. Functional Block Diagram of the Flux-Total Flow, Startup Nuclear. Modified Nuclear Rate.and Primary to Intermediate Flow Rate Protective Subsystems. One Channel of Three is Shown 4558-I 7.2-36 Amend. 57 Nov. 1980
O TO NON-PPS + BUFFER
- I SL L -
LEVEL DETECTOR LOOP NO.1 ELECTRONICS [ Sl L TO NON-PPS + BUFFER LEVEL
' {
DETECTOR ELECTRONICS I LOOP NO. 2 SL I TO N O N P.*S + BUFFER LEVEL DETECTOR ELECTRONICS LOOP NO. 3 Figure 7.2-9. Functional Block Diagrams of the Steam Drum Lesel Subsystem. One Channel of Three is Shown. 4558-5 7.2-37 Amend. 57 O l Nov. 1980 l
TO NON-PPS
- BUFFER O
TEMPERATURE DElECTOR - ELECTRONICS - LOOP ai + UT TO NON-PPS + BUFFER T EMPERt.T URE DETECTOR ELECTRONICS LOOP =2 a1 + TO N0h-PPS -*- BUFFER TEMPERATURE DETECTOR - ELECTRONICS k LOOP m3 g -[ EVAPORATOR OUTI.ET TEMPERATURE i . TO SWRPRS e BUFFER (SEE FIG. 7.5-6) SENSOR SUPERHEATER ELECTRONICS TO SWRPRS BUFFEP. (SEE FIG. 7.5-6) + ap. SENSOR
+
AUCTIONEEE - EVAPORATOR hI El ECTRON ICS
~
H1 . BUFFER TO NON-PPS BUFFER ' (SE FG 7.5-6) ' LO'd BUFFER TO NON-PPS EVAP RAT #2 S0DIUM WATER REACTION PROTECTIVE SUBSYSTEM Figure 7.2-10 Functional Block Diagrams of the Evaporator Outlet Sodium 6678-4 Temperature and Sodium Water Reaction Protective Subsystems, 1 One Channel of Three is Shown ! 7.2-38 Amend. 57 Nov. 1980
7.3 ENGINEERED SAFETY FEATURE INSTRUMENTATION AND CONTROL The initiation of containment isolation is the only engineered ' safety feature identified which requires description in this section. Other pd engineered safety features are identified in Section 6.1. Table 6.1-1 lists these features and the sections in which they are discussed. The valves and associated characteristics are described in Section 6.2.4. Accident analyses are presented in Section 15 for the postulated events requiring containment isolation. The material below describes the design basis for the system, the implementation of the instrumentation and controls, and the evaluation of the capability of the system with respect to the General Design Criteria and applicable IEEE and RDT Standards and Regulatory Guides. 7.3.1 Containment Isolation System 7.3.1.1 System Description 57 The Containment Isolation System (CIS) is comprised of redundant instru-mentation which senses the need for closure of certain valves in lines directly connected to the containment atmosphere, logic to initiate :losure of the valves, manual initiation equipment, and the valves which are described in Section 6.2.4. Figure 7.3-1 shows a block diagram of the system. The Containment Isolation System is designed for automatic activation of these valves in lines directly connected to the containment atmosphere and valves which require closure in less than 10 minutes to remain within limits. Where closure time is not required in less than 10 minutes, manual actuation is provided. Radiation sensors are provided in two areas: the exhaust duct of the containment ventilation and the head access area. Three inde-pendent, redundant measurements are provided at each location. The sensor
,-) output is conditioned by the electronics and transmitted to the comparators (d where it is compared with a setpoint. If the signal is greater than the setpoint, a comparator trip is initiated. The logic for the automatic containment isolation is functionally identical to that used for the Primary Reactor Shutdown System. The comparator output is optically 44 coupled to the logic. Within the logic, the two comparator outputs (one from the head access area, the other from the exhaust duct of the containment ventilation) are combined to feed 2/3 coincidence modules.
Within the 2/3 modules, the three independent channels are combined. Any two in the tripped state results in closure of the isolation valves. One of the logic trains drives the in-containment automatic valves. The other logic train drives the ex-containment automatic valves. Figure 7.3-? is a 44 Logic Diagram of the system. Il 3 Amend. 57 y 7.3-1 Nov. 1980
Each detector has a check source which is used for test. In addition to the comparators and electronics, buffers and power supplies are provided for each channel. No permissives or bypasses are provided. The provisions for on-line testing of electrical and mechanical equipment are included in the design. The test source is used to test the 44l instrument channel through the comparator. Signals are inserted prior to the 2/3 module to test the remainder of the logic and the closure of 57 the automatic valves. Since closure of the valves does not require reactor shutdown, this test can be performed during power operation. Sj Channel output meters are included to provide the operators with early indication of anomalous instrumentation performance. This equipment is not safety related. There are no interlocks included in the design nor is there a necessity for sequencing the closure of the valves. Manual initiation capability is provided locally at the CIS breaker cabinets and within the control room on the main control board to close all CIS 57 valves. Capability to close the CIS valves is provided if loss of access to control room is assumed. For all containment isolation valves, position indicating lights are provided in the control room. 7.3.1.2 Design Basis Information The CIS initiates and completes closure of the identified isolation valves to prevent the results of the faults identified in Table 7.3-1 from exceeding the specified limits. Note that these limits apply to the CIS when 57 the containment hatch is closed. Further, the design basis for the CIS is to provide appropriate design margin for postulated events and to assure that the radiological consequences of such events are within the guideline values of 10CFR100. 7.3.1.2.1 Containment Isolation System Subsystems Containment Exhaust Radiation The Containment Exhaust Radiation Subsystem initiates automatic containment isolation valve closure if the radiation level in the contain-42l ment exhaust exceeds preset limits. This subsystem assures that events releasing activity within the containment do not result in exceeding the limits (10CFR20 or 10CFR100 as appropriate) for exposures in unrestricted a reas. As shown in Fiaure 7.3-1, the subsystem includes 3 radiation sensors located in the contar - Laust whose output is compared to a fixed set-point. The subsysta _r bypassed. Worst case values for time response and repeatibilities w111 be used in the final analysis of the performance of this subsystem. Amend. 57 Nov. 1980 7.3-2
Head Access Area Radiation The Head Access Area Radiation Subsystem initiates closure of the containment isolation valves in the event of large radiation releases in the (~j head access area. Three radiation sensors are located in the head access (/ area to provide early initiation and closure of the isolation valves to assure that releases from design basis events do not exceed the guideline values of 10CFR100. 7.3.1.2.2 Essential Performance Requirements To implement the required isolation function within the specified limits, the CIS must meet the functional requirements specified below: The closure time requirement for the inlet and exhaust isolation valves is 4 seconds with a three second or less detection time in the heating and ventilating system. A 10 second transport time from sensing point to the valve exists (3ee Section 15.1.1). The 3 seconds includes 43l sensor time response, comparator and logic time delays. 30 The CIS is designed to meet these requirements for the 57 environmental conditions described in Section 7.2.1. 7.3.2 Analysis The design of the CIS provides the necessary functional performance and design features to meet the requirements of the appropriate standards p specified in 7.1.2 as described below. 7.3.2.1 Functional Performance l The analyses in Sections 15.5 and 15.6 shows the results of the postulated fault conditions. These analyses assumed a closed containment where the events occurred with the containment hatch closed. For the limit-ing event, primary drain tank fire during maintenance, scoping analyses have been performed to determine the required closure time of the containment isolation valves. For the primary drain tank fire, closure within 20 minutes is adequate. Further, analyses to detemine the required closure time under postulated accident conditions have been performed and are discussed in Section 15.1.1. These analyses are used to determine the available design margin. The results of this assumed condition do not exceed the guideline values of 10CFR100 if the main exhaust and inlet valves are closed within 4 seconds assuming the normal air transport time from the detector to the valve is 10 seconds or more, a 14,000 cfm nnrmal ventilation rate. 57
,m (j Amend. 57 7.3-3 Nov. 1980
57l Since the automatic CIS is designed to isolate within the above time response requirements, all of the design basis conditions are teminated within the necessary limits for the present design concept. 7.3.2.2 Design Features 57l The CIS instrumentation, controls and actuators are designed to meet the requirements of IEEE-279-1971 and RDT Standard C16-IT, Dec. 1969. The analyses of compliance with these are summarized below. Single Failure No single failure within the CIS nor removal from service of any component or channel will prevent protective action when required. There are three independent instrument channels for each necessary measurement, 44 l two independent 2/3 logic trains, and two independent actuators provide'd (as 57 shown in Figure 7.3-1). Bypasses No bypasses are provided. Multiple Setpoints fiultiple setpoints are not required. Completion of Protective Action The automatic CIS is designed so that, once initiated, protective action at the system level must go to completion. Return to normal operation requires 57 manual reset of the CIS breakers by the operator. Manual Initiation The CIS includes means for manual initiation of cor,'ainment isolation at the system level. No single failure will prevent manual 57l initiation of the containment isolation action. Control and Protection Interaction There are no shared components between the control system and the CIS. The provisions for access, information read-out, annunciation of trips, and periodic testing are as specified for the Reactor Shutdown System in Section 7.2.2. Amend. 57 Nov. 1980 i 7.3-4
O C) G G v' TAisLE I.3-1 CONTAINMENT ISOLATION SYSTEM DESIGN BASIS Applicable Event Federal Regulation Limit Anticipated Fault 10CFR20 5 105 <2 millirem in any one hour No examples of anticipated faults which lead <100 millirem in any one week to release of activity have been identified. Unlikely Fault 10CFR20 5 403b <5 rem in any two hours No examples are presently identified for the automatic containment isolation system design basis.
" <25 rem in any two hours
. Extremely Unlikely Faults & Design Margin
- 10CFR100 Y
Examples include major sodium fires <300 rem iodine doses in the thyroid in any two hours
<75 rem to the lung 44 <150 rem to the bone d
*The design basis for the CIS includes limiting the results of postulated accidents within the guideline values of 10CFR100. See Section 15.1.1.
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HEAD ACCESS G AMMA DETECTOR 8 E > - e vglg','aa 'x"aus' EITHER TRIP INPUT WILL INSIDE CONTAINMENT h PRODUCE A TRIP OUTP VALVES ELECTRONICS h COMPARATOR REQUIRES 2 TRIP INPUTS , OB OB OB DUT OF THREE TO PRO- i
=- DUCE A TRIP OUTPUT CIS METERS M MANUAL ACTUATION W
O , OUTSIDE Q V ISOLAT10N BUFFER 2/31 CONTAINMENT VALVES Or v u v s - 5w/ t THE CONTAINMENT ISOLATION SYSTEM DOES NOT H AVE PERMISSIVE BYPASSES. 2I3
- 8 8 D Hg @@@ _
g- CIS METERS 8% Figure 7.3-2. Contaiivnent isolation System Logic Diagram
T 7.4.3 Remote Shutdown System (V
A Remote Shutdown System is provided. It consists of the following provisions: Remote lccal control of: o Steam Generator Auxiliary Heat Removal System (SGAHRS) o Diesel Generator System o EVST Cooling Systems Remote Shutdown Panel Instrumentation: o Primary and Intermediate Hot Leg Tenperatures o Primary and I, termediate Cold Leg Temperatures o Primary and Intermediate Pump Speeds o Steam Drum Pressure, Temperature and Level o Steam Flow o Reactor Sodium Level o Diesel Generator Power Output, Current, Voltage Frequency, RPM, Fuel Oil Level The Remote Shutdown System also includes the following features: o The SGAHRS and Diesel Generator System are provided with a local transfer switch which transfers controls from the control room to the local stations o An isolation switch is provided at the local station which transfers signals and isolates power from the control room for all necessary non redundant instrumentation. o Transfer of instrumentation or control to the local control station will not be possible until the local transfer switch is placed in the " local operating" position, resulting in actuating an alarm in the control room. o An independent soundpowered communications network is provided between the local safe shutdown control stations o All PPS signals necessary to remote shutdown control are isolated from the control room. i O) i s The need for remote shutdown (i.e., control room evacuation) is assumed not to occur simultaneously with the recovery from any other con-dition except the loss of off-site pow.r. The Remote Shutdown Panel is 57 Amend. 57 1 located in cell 271 of the SGB, adjacent to the SGAHRS control panels. Nov. 1980 1 7.4-8a
- 7. 5 INSTRUMLtiTATI0riANDf$0fitTORINGSYSTEM
(~ The instrumentation and monitorinq systems included in this Section ( are the Flux Monitoring System, the Heat Transport Instrumentation System, the Reactor and Vessel Instrumentation System, the Fuel failure Monitoring System, the Leak Detection Instrumentation System, and the Sodium-Water Reactor Pres-sure Relief System. Table 7.5-1 lists the measured parameters and instrumen-tation provided by these systems. The instrumentation which is safety related as defined in Section 3.2.1 is identified with an asterisk in column 2 of Table 7.5-1. Instrumentation and monitoring for TLTM parameters not 25, included in the desian basis are also discussed. These include containment hydrogen monitoring and containment vessel temperature and ' pressure rronitorirm. 7.5.1 Flux fionitoring System The objective of the Flux Monitoring System (FMS) is to provide indi-cations and electrical signals proportional to reactor power for reactor plant control and protection. The FMS meets its objective by means of neu-tron measuring instrumentation comprised of sensors and signal conditioning equipment which provide indications and signals for conditions of reactor shutdown, startup and full puwer operation. Neutron sensors located around the periphery of the reactor guard vessel sense thermalized reactor leakage flux which is proportional to the reactor flux and thus to reactor power. Signals from the sensors are con-ditioned and then used to do the following: i 1 e Determine the flux status of the reactor from shutdown through startup and all power levels. e Provide signals to the Plant Protection System (PPS) to initiate . reactor protective trips. l ! e Provide signals to the Plant Control System (PCS) for reactoi anf plant control. o Provide neutron flux information for display, annunciation and recording. A block diagram of the FM System is provided in Figure 7.5-1.
- 7. 5.1.1 Design Description The Neutron Flux Monitoring System.provides three ranges of instru-mentation: Source Range, Wide Range and Power Range. Each range of instru-mentation is provided in three identical channels comprised of a detector, preamplifier (source and wide ranges), junction box (power range) and signal conditioning equipment. The Flux Monitoring System measures neutron flux proportional to reactor power over a span or more than ten decades from shut-down to above full power and provides indications and electrical outputs for plant protection, plant control, data handling and display, recording, and annunciation.
O 7.5-1 Amend. 25 Aug. 1976
The flux detectors will be within the reactor cavity and contained in thimbles positioned approximately 120 apart on the periphery of the guard vessel. Each detector will be separately installed in a dry thimble which extends from the head compartment floor to below the reactor midplane to 57 l position the detectors with the center of their active volumes at or near the core midplane. The thimbles will be surrounded by neutron thermalizing blocks and provided with gamma shielding at the detector positions to provide suitable gamma and thermal neutron environments for the detectors. Preamplifiers and junction boxes for the system will be located in the head access area. The signal processing equipment will be located in the control building and will be contained in equipment racks. Within these racks, there will be drawers containing the electronic signal conditioning equipment packaged such that each channel or range within a channel (in the case of the wide range channels) will occupy one drawer. The drawers will be mounted on slides and will be capable of being withdrawn for alignment and maintenance. The front of each drawer will have meters, switches and adjustments for alignment, tests and monitoring of the channel . The arrangement of drawers in the racks will satisfy the separation require-ments for redundant PPS channels. (See Section 7.1.2 and 7.2.2) Remote meters and switches will be provided to the reactor plant operator to permit him to read the following parameters for each channel: Source Range Channel logarithmic level and rate, logarithmic percent power and rate for Wide Range counting and Mean Square Voltage (MSV) channels, linear percent power for Wide Range and Power Range. Linear and t.ogarithnic Source Range level will be provided on the IVTM control console. Audio Source Range count rate will be provided in the control room and at the IVTM control 57 console. Logarithmic source range count rate will be provided in the Refueling Communication Center. The Flux Monitoring System detectors, cabling and signal condition-ing equipment will be installed so as to preserve the separation require-ments for redundant PPS channels (see Section 7.1.2 and 7.2.2). Figure 7.5-2 presents the flux level coverage of the FMS instrumenta-tion ranges bastd on the neutron flux at the detectors. The instrumentation ranges are shown to overlap so as to provide continuous indication from shutdown to more than full power. 7.5.1.1.1 Source Range The design bases functions and operational requirements as stated ! in Section 4.3.2.1.5 are accomplished by incorporation of the design features described in the following paragraphs. Each of the three Source Range channels will use a high sensitivity, dual 57l section BF 3 filled proportional counter assembly to sense thennal neutron flux ver the range of approximately 0.1 to 104 neutrons per cm2 per second (nv). 54l This corresponds to a range from the fully shutdown fresh core conditions i Amend. 57 7.5-2 Nov. 1980
to reactor low power (a few Kwt) operation. The sensitivity of the 541 BF3 counters will be maintained at a minimum of 40 cps / thermal equivalent O7 nv at a gamma background count rate of approximately I cps. Shielding will be V provided to limit the gamma do",e rate at the detectors due to prompt gannas from the core and from decay of activation isotopes in the sodium coolant and structural materials to less than 100R/hr. This will apply over the operating range of the channel at all times when operation is required. The BF3 counters and associated cables will be designed to operate under normal environmental conditions of 170 F maximum and atmospheric pressure and under emergency conditions of 260 F maximum and 12 psig during con-57 tainment testing. The design life ggl nyt. for the counters is 3 full power years based on a total fluence of 10 In order to achieve this life-time without retracting the counters, the operating voltage will be removed and the anode of the counter shorted to ground through an appropriate resistance when the flux level is above the operating range of the channel. The output pulses from the detector assembly will be amplified by a dual section pre-amplifier mounted in the head access area and routed to 'the signal con-ditioning equipment in the control room racks. This equipment will consist of dual amplifiers with pulse shaping networks to minimize gamma pulse pile-up followed by discriminators to reject amplifier noise and gamma pulses. 57 The summed output of the discriminators will drive a count rate circuit followed by a scaling amplifier which will produce an analog output signal pro-portional to the logarithm of the input pulse rate. The analog signal will be displayed on a log scale countrate level meter calibrated from 1 to 106cps. The accuracy and repeatability of the channel will be + 3% - and i 1%, respectively, of linear equivalent full scale under the 17 worst case environmental conditions of temperature, pressure and input g) t power fluctuations to be encountered while channel operation is required. The response time of the count rate circuits will vary with count rate, being on the order of 30 seconds at the lowest expected average count rate of approximately 4 cps which will occur during refueling beginning of core life conditions. The lo statistical counting error will be approximately 17% 541 at the 4 cps rate. The response time will decrease with increasing count rate to approximately 0.1 seconds at a count rate of 3 x 105 cps produced 57l at a few kilowatts. The analog output signal of the scaling amplifier also drives a differentiating circuit to produce a rate of change of level signal which is displayed on a linear scale meter from
-1 to 0 to +3 decades per minute. Power supplies mounted in the equipment 17 drawers will provide detector excitation high voltage and low voltage instrument power. Individual scaler timers will be provided in each channel which will be driven from a buffered output of the discriminator which precedes the countrate circuit. The scal _er timers will be programmable to count for a short preset time, stop, trans.f_er the accumulated count to tempo-rary storage for PDH&DS readout, immediately restart the count and repeat the l 57 sequence to provide an accurate record of the count rate trace to implement the l inverse kinetics rod drop technique used to estabilsh control rod worth. The l scaler-timers will also provide for counting of signal pulses for longer time 1 57 periods to accurately determine the system calibration constant which relates 17 subcritical reactivity to count rate.
A visual / audio linear count rate circuit will be provided for
, operations at shutdown. This circuit will be provided with a switch f b
Amend. 57 Nov. 1980 7.5-3
through which it can be connected to the summed discriminator output of any of 57 the three Source Range Channels, one at a time. The visual portion provides expanded scale indication of the flux level. The audio portion provides tone bursts whose rate changes (increase and decrease) with increasing and decreasing flux level. Each channel will be provided with built-in counting level and rate of change calibration circuits for channel alignment and pre-operational testing to assure that the instrumentation circuitry is functioning properly. The Source Range Channels are used to monitor the shutdown and startup flux only, no signals are provided to the PPS. Bistable comparators in each channel will activate individual annunciators in the control room to provide alarms if the specified minimum shutdown reactivity is exceeded during re-fueling or if detector excitation or instrument power is lost in any channel. To improve the operational effectiveness of the source range shutdown reacti-vity monitoring function, the reactivity alarm is inhibited during core assembly movement which could cause an erroneous alarm. The inhibit circuit is continuously monitored for proper operation and any malfunction of the inhibit circuit will activate a separate control room annunciator. An alarm will also be provided in the control room via the PDH&DS if any one channel deviates from the other two by a preset amount. The snutdown margin alarm bistables and the PDH&DS channel deviation alarm will operate off of the buffered outputs of the log count rate scaling amplifiers. The loss of detector excitation and instrument power alarm bistables will operate off of H.V. sense signals developed by a resistor divider network at the preamplifier H.V. outputs to the detector assemblies and instrument power supply output voltages, respectively. The operation of the source range channels will be under manual control of the reactor operators. These channels will be in operation continuously during reactor refueling and other shutdown conditions. During reactor startup, the source range channels will be used to monitor the core flux level until a predetermined overlap between the source range channels and wide range log count rate channels is obtained. The high voltage will then be removed from the source range detectors by the operator. He will do this by actuating a momentary contact pushbutton switch on the main control panel. When actuated, this switch will remotely interrupt the input power to the detector high voltage supplies, and ground the detector anodes through an appropriate resistor. The relay control circuit established by the momentary contact of this switch will also illuminate a green indicator light located adjacent to the switch. Upon reactor shutdown, the operator will interrupt the relay control circuit by actuating a second momentary contact pushbutton switch when the wide range log counting channels indications fall to a pre-determined level within the source range channels operating range. This action will remove the ground from the detector anodes, restore the detector excitation voltage, extinguish the green indicator light and illuminate a red indicator light. Inadvertent removal of, or failure to restore the operating voltage when needed will be prevented by procedural control, utilizing the separate on/off switches and with monitoring through the color of the illuminated indi-57 cator light. Amend. 57 Nov. 1980 1 7.5-3a
,p , U' Output of the detector will be processed in the signal conditioning equip-ment to provide linear indication of percent power and linear output signals for plant protection, plant control, data logging and annunciators. This instrumentation operates over the same flux span as the direct current cir-cuitry of the wide range instrumentation to add redundancy and diversity to 57 l the power range measurements. The output of this instrument will be linear to at least 140 percent power and will have no foldover to as high a power level as required by the worst case power overshoot for which protection must be provided. Built-in test circuits and controls will be provided to permit testing and aligning the equipment during plant operation and plant shutdown. 7.5.1.2 Design Analysis The Flux Monitoring System will be a functional subsystem of the Plant Protection System and will meet the safety related channel performance and reliability requirements of the CRBRP General Design Criteria, RDT 27 l Standard C16-IT, Dec.1969, IEEE Standard 279-1971, applicable AEC Regulatory Guides and other appropriate criteria and standards by complying with the applicable design requirements delineated in Section 7.1.2. The FMS meets CRBRP General Design Criterion 21, which is applicable /~N \ to instrumentation for normal and accident conditions, as follows: l w/ e The shutdown flux level will be monitored at all times while fuel is in the core so as to provide safe operational control of the reactor during low power, normal shutdown, refueling and shutdown maintenance operations. e The reactor flux will be continuously monitored during operation from shutdown to full power operation (i.e. , overlap will exist between cascaded channels so that all power levels can be moni-tored without a gap in range). e Reactor power operations will be continuously monitored with linear response to power up to at least 140% full power. Sig-nificant positive response will be provided to as high a power level as required by the worst case power overshoot for which protection must be provided. This positive response will be provided for as long as is required to seal in the scram trip. 57 l e The FMS instrument response times will meet the require-ments of the Reactor Shutdown and Plant Control Systems. e Indication of reactor power level and rate of change of power level will be provided to the operator. One set of meters and a i i selector switch will be provided for each range of instrumenta- ,- tion permitting the operator to select one channel at a time to ! \q j l
.- 1 l
l Amend. 57 Nov. 1980 . 7.5-4 l l l
be displayed on the related meters. Five power level meters, five selector switches and three rate of change of power meters will be (n v) provided for the operator. e The source range level will be indicated in logarithmic counts per second and rate of change of level in decades per minute. Linear count rate will be provided at shutdown at the fueling console and at the FMS system panels in the control roon . Audible count rate indication will be provided in the control room and in con-57 tainment at the refueling console. e The wide ranges will be indicated as follows: Counting channels - Logarithmic percent power level and decades per minute rate of change. MSV channels - Logarithmic percent power level and decades per minute rate of change. DC channels - Linear percent power level. e The power range will be indicated in linear percent power level. Preliminary Failure Mode and Effect Analysis results applicable to the FMS have been datermined in an analysis of possible failure modes and their effects on the Reactor Shutdown System performance and are presented.in l Tables C.S.1-4 and C.S.1-5. p-j 7.5.2 Heat Transport Instrumentation System 7.5.2.1 Description The Heat Transport Instrumentation System provides sensors, associ-ated signal conditioning equipment and controls other than Plant Control, for the Primary Heat Transport, the Intermediate Heat Transport and the Steam Generator. The signals from the sensors are conditioned and then supplied to the Reactor Shutdown System logic, the Plant Control System, the Data Handling and Display System, and the Plant Annunciator System as appropriate. The location of the Heat Transport Instrumentation is provided in Fig-ures 5.1-2 and 5.1-4 (P&ID's). 7.5.2.1.1 Primary and Intermediate Sodium Loops Reactor Inlet Pressure The measurement is made by pressure elements installed in the cold leg of the primary loop piping just before it enters the reactor vessel. NaK filled capillaries from the pressure elements are connected to pressure trans-ducers which develop electrical signals proportional to the pressure. These pressure transducers provide a secondary boundary if the bellows in the pres-sure elements should fail. (n ! Amend. 57 7.5-5
Each pressure transducer consists of a diaphragm whicn moves in response to pressure changes in the NaK filled capillary, a strain gage which converts diaphragm motion to resistance change, and a bridge and amplifier to convert strain gage resistance change to a standard signal. Since pressure element replacement requires plant shutdown, two pressure elements per loop are provided. The signals from the six, two per loop, pressure measurements are transmitted to the control room in three separate isolated PPS channels for use in the Reactor Shutdown System logic. The Reactor Shutdown System sup-plies buffered signals to the DH & DS. Primary and Intermediate Loop Flow The flow measurements are made by a permanent magnet flowmeter located in the cold leg of each primary and intermediate loop (except 49 for intermediate loop 2, which has the PM flowmeter in the hot leg). The magnet assembly is in the shape of an inverted "U" which is suspended around the pipe. The magnet assembly is mounted rigidly to the building structure and is physically separated from the pipe. Type K thermocouples are. installed in the magnet structures to moni-tor the magnet temperatures. This permits temperature corrections to the flowmeter calibration. The signals from these thermocouples are routed to a local panel. Provisions will be made to permit periodic monitoring of the magnetic flux of the flowmeters without disassembly or entrance into HTS cell. This is also accomplished at the local panel. Four independent pairs of 3/8" (approx.) electrodes are attached to the pipe. The electrodes are of the same composition as the pipe so that thennal potentials are not developed. Three pairs of electrodes are con-nected to the conditioning equipment. The fourth pair is available for a portable measuring instrument or as an installed spare. Flexible mica, polyimide and fiberglass insulated cables in separate conduits to meet PPS separation requirements are used to bring the four sig-nals from each flowmeter assembly out of the Heat Transport System cell. The signals are then routed to signal conditioning equipment. From the signal conditioning equipment, the signals are sent to the control room for the Reactor Shutdown System logic which in turn supplies buffered signals to the PCS and the DH & DS. IHX Primary Outlet Temperature i The IHX primary outlet temperature measuremcat is made by three l Chromel/Alumel thermocouples per ' cop installed in thermowells in the ele-vated sectics of the HTS cold leg piping nearest the IHX primary outlet. The thermocouples are 1/8" insw'.ated junction swaged to 1/16" at the tip to l O 7.5-6 Amend. 49 Apr. 1979
! 7.5.3 Reactor and Vessel Instrumentation 7.5.3.1 Description The Reactor and Vessel Instrumentation System includes all in-vessel temperature, sodium level and vibration sensors for instrumenting the reactor parameters required for the Reactor Shutdown System, PCS, surveillance and design verification. It also includes signal conditioning equipment needed to make the sensor signal usable.in the systems receiving the signal. Table 7.5-2 shows the in-vessel instruments provided, their location, their quantity and purpose. 7.5.3.1.1 Sodium Level A total of five sodium level sensors are provided. All of 57 l these sensors are mounted in wells to provide the physical barrier maintaining the integrity of the primary loop closed system. The sensors are induction type probes continuously sensitive over their entire length. Four of the units, located approximately equally spaced on the top of the reactor, are short with a sensing range of from 6 inches above the operating level to 24 inches below. Three of these provide the level signals to the three Reac-tor shutdown system logic and are thus isolated from each other and from non-PPS equipment. The fourth is an installed spare unit providing a means of maintaining the three operating channels without a shutdown in the event of failure of one of them. The fif th level sensor is located close to one of the short units but provides a measuring range from 6 inches above the operating b) D level to below the minimum safe sodium level. It has approximately fifteen feet of sensing length. The signal is supplied to the PCS for conti 21 room indication and is monitored at all times, including refueling. 7.5.3.1.2 Temperature All in-vessel temperatures are sensed by chromel alumel, 40 l ungrounded, stainless steel sheathed thermocouples. Thirty wells are provided for thermocouples located in the sodium at the exit from the core. These thermocouples provide signals to the PCS. Additional wells are provided at the core exit (275), core periphery (2), and on parts of the upper internal structure (6) for ther~mocouples providing signals to the PDH & DS for surveil-41 l lance and desian verification. t 40 57 ' 7.5.3.1.3 Hon-replaceable Instruments 41 40 Within the reactor vessel, four biaxial accelerometers are rmunted on the upper internal structure so that they cannot be replaced. These sensors are not required to fuiction beyond the first 5 six months of operation although they are required to physically withs tand g the sodium environment Fr he life of the reactor. The t ignals 40 ! v Amend. 57 7.5-13 Nov. 1980
provided by these sensors provide design verification information and the location of the sensors and their leads will not affect the safety of the plant. 7.5.3.2 Analysis The in-vessel sensor mounting is designed to be operable during the combined stresses imposed by the reactor coolant velocity, vibration, pres-sure, environment, temperature, thermal shock, and radiation in order to pro-vide operational lifetime that will not significantly affect the reactor availability. The sensors and their lead-out conductors are sufficiently rugged so that they will not be damaged during refueling and maintenance. Thermocouples used for measuring the sodium temperature are mounted to avoid close proximity to the structures so that the temperatures sensed will be that of the coolant and not be influenced by the structures. 57l The sodium level instruments, which are part of the Reactor Shutdown System, will comply with the PPS Design Requirements (see Sections 7.1.2 and
- 7. 2.1 ) . The design analysis for the Reactor Shutdown System applies (Section 7.2.2) and a Failure Mode and Effects analysis is performed as
[ shown in Table C.S.1-1. 41 7.5.4 Fuel Failure Monitoring System The Fuel Failure Monitoring (FFM) System provides: 34l 1. Equipment to detect occurrance of fuel or blanket cladding fail u re .
- 2. Equipment to locate failed fuel assemblies in the reactor core, and, to the extent practicable, failad radial blarket assemblies;
- 3. Equipment to characterize the failed pins as to burn up and other information, to pennit correlation with core and blanket history, thus enhancing location capabilities, in particular for radial blanket.
- 4. Equipment to detect fuel or blanket failures involving exposure 34 to and egress of fissile or fertile material to the sodium coolant.
The FFM System is comprised of several independently functioning parts, each providing information to the plant operations staff. This system does not provide control loop or reactor trip signals, but does supply infor-mation for surveillance, display and alarm purposes. Amend. 57 ! Nov. 1980 l 7.5-14 l l l
40 d 57 l e Pressure is monitored in the Reaction Products Separator Tank to alert the operator to off-normal conditions in the Reaction Products Vent System. 57l 7.5.6.1.6 Sodium Dump Tank Instrumeatation i sodium dump tank is provided in each loop for drainage of sodium from the IHTS. Dump tank level is measured by inductive type level probes. Each tank is provided with two probes to meet the necessary range requirements. Each probe is connected to an excitation-con-ditioning module that provides local indication of sodium levels. I
- The conditioned signal from the wide-range probe is supplied to a trip unit, which provides a high-level alarm signal to the PAS.
, All level measurement channels provide inputs to the PDH&DS.
i Surface thermocouples, which are part of the sodium dump tank trace heating system, are available for display as an indication of level and temperature. A pressure switch actuates an alarm and a rupture 4 disc bursts to vent to atmosphere in case of excessive pressure buildup in the sodium dump tank. 57l 7.5.6.1.7 Water Dump Tank Instrumentation A water dump tank is provided in each loop to accept and store the water from the evaporators when rapid depressurization is required.
- Measured parameters for the water dump tank are level, pressure, and temperature.
Dump tank level is measu: ed by a differential pressure transmitter that senses the difference in pressure caused by the variable height of water in the dump tank. The differential pressure signal is supplied to a local indicator, the PDH&DS, and a high-level alarm circuit that provides a signal to the PAS. Pressure is measured on an appendage off the dump tank by a conven-tional pressure sensor-transmitter. The pressure signal is used for local indication, for the PDH&DS, and for high-pressure alarm to i the PAS. i i Dump tank temperature is monitored by chromel-alumel thermocouples l attached to the surface of the tank. The signals are supplied to 1 the multipoint temperature indicatoc 40 J Amend. 57 7.5-33 Nov. 1980
, - - - e , , , - ~ -, , . r -,- + -- ,----r - - - , .
O 7.5.6.2 Design Analysis Because of the large increase in pressure from the formation of reaction products during a large sodium-water reaction, rupture disc operation is necessary to prevent excessive pressure surges in the Inter-mediate Heat Transport System and possible primary boundary rupture at the Intermediate Heat Exchanger. Reaction products vent line sensors are part of the Reactor Shutdown Syster and as such meet the requirements of the Plant Protection System (see Section 7.1.2 and 7.2.2). The initiation of isolation and dump of the water side of the steam generators, trip of the recirculation pumps and inerting of the steam generators normally follows after rupture disc operation in a large sodium-water reaction and is desir-able from the operational standpoint of minimizing the time to recover from the incident. However, the initiation of these actions after rupture disc operation is not necessary from a safety standpoint to assure protec-tion of the core or the safety of the public. All SWRPRS equipment associated with isolation or dump of evapor-ator or superheater modules is designed to assure that no credible single event can disable more than one of the three redundant decay heat removal paths. All electronic equipment is designed to withstand the Safe Shut-down Earthquake. The mechanical equipment associated with the 3 steam generators is physically separated in 3 different steam generators bays. Electrical and pneumatic supplies are arranged such that a single failure disables at most one decay heat removal path. Where practicable, the preferred failure position for equipment is in a direction to assure the safe operation of the SGAHRS. SWRPRS equipment whose failure could cause loss of decay heat
..noval capability of the SGAHRS is safety related. Any credible single failur e in the SWRPRS can lead to the failure of at most one of the three decay heat removal loops. Since the three decay heat removal loops are redundant and independent, the SGAHRS will meet the single failure criterion and the adequacy of the decay heat removal system following a credible single failure in the SWRPRS is assured.
O 7.5-33a Amend. 40 July 1977
7.5.7 Containment 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. 25 7.5.7.1 Design Description The hydrogen instrumentation consists of two fully redundant and independent analyzer channels. The containment atmosphere is sampled through an entry filter located near the top of the RCB. The air samples are pumped to the analyzer which is located in the SGB and operates on the principle of thermal conductivity. From there signals go to the Cortrol Room where the hydrogen concentration readout is provided. This instru-ment is also required to perform functions for events which lie beyond the design basis for the plant. This instrument is further discussed in 57 this capacity in Sections 2.1 and 3.3 of Reference 10b of PSAR Section 1.6. t 7.5.8 Containment Vessel Temperature Monitoring The objective of Containment Vessel Temperature Monitoring is to provide indication in the Control Room of the containment vessel 25 tempera ture. 7.5.8.1 Design Description " The temperature instrumentation consists of two fully redundant and independent channels. Each channel consists of eight themocouples mour * :d at various locations on the inside of the containment wall, with each thermocouple providing a signal to conditioning instrumentation in the SGB. The instrumentation sends a signal to the Control Room where individual readout is provided. This instrument is also required to perform functions for events which lie beyond the design basis for the Plant. This instru-57 ment is further discussed in this capacity in Sections 2.1 and 2.2 of Reference 10b of PSAR Section 1.6. 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 l corininment above the operating floor. 7.5.9.1 Design Description i The pressure instrumentation consists of a pressure detector inside the containment vessel. Signals will be provided to the display 25 and alarm panel in the Control Room so that continuous readout will be provided to the plant. operator. This instrument is a~lso required to perform functions for events which lie beyond the design basis for the plant. This instrument is further discussed in this capacity in Sections 57 2.1 and 2.2 of Reference 10b of PSAR Section 1.6. b 7.5-33b Amend. 57 Nov. 1980
l l 7.5.10 Containment Atmosphere Temperature The objective of the Containment Atmosphere Temperature Moni-toring System is to provide indication in the Control Room of the atmosphere temperature inside the containment building. 7.5.10.1 Design Description The temperature instrumentation consists of two fully redundant and independent channels. Each channel consists of two thennoccupler mounted on the RCB dome, with each thermocouple providing a signal to conditioning instrumentation in the SGB. The instrumentation sends a signal to the Control Room where individual readout is provided. This instrument is also required to perfonn functions for events which lie beyond the design basis for the plant. This instrument is further discussed in this capacity in Section 2.1 and 2.2 of Reference 10b of 57 PSAR Section 1.6. 57l 7.5.11 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 sic,nals is provided in the Control Room. The instruments which
, serve the Post Accident Monitoring function are included in those discussed "9 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 corresponding to the parameter of Table 7.5-4 50 are described in the following paragraphs.
7.5-33c Amend. 57 Nov. 1930
f-~s U The reactor sodium level is monitored to enable the operator to determine whether manual action is necessary to mitigate conditions which may cause low Reactor Vessel sodium levels. The operator may also use the reactor sodium level to determine conditions in the primary sodium systems such as the volume of primary coolant leakage. The IHX (Intermediate Heat Exchanger) inlet and outlets temperatures are monitored to verify the heat transfer from PHTS to the IHTS. These monitors allow the operator to take manual actions necessary to assure decay heat removal from the reactor is maintained within design limits. The DHR$ cold leg temperature is monitored to assess the systems' decay heat removal performance so the operator may take manual actions necessary to achieve or maintain core temperature at a safe level. Reactor Containment Building pressure and temperature monitors are provided to follow pressure and temperature changes due to an accident, and provide confidence that the accident consequences are within the capability of the Containment Vessel. Also, these monitors can be used to detect conditions beyond the design basis as discussed 57l in Reference 10b of Section 1.6. g-~s The PWST (Protected Water Storage Tank) water level is 5 monitored so the operator is assured of the adequacy of that water supply to remove reactor decay heat for an extended period of time. In addition, the level monitor provides indication of the long term need to refill the PUST or draw auxiliary feeduater from an alternate source. The Auxiliary Feedwater Flow is monitored to inform the operator of normal, abnormal or inadequate flow. He can take manual actions to provide adequate flow and thus maintain adequate decay heat removal through the steam generator system. ) 4 Steam drum level and pressure are monitored to identify an accident and to allow the operator to take manual actions to initiate and control systems required to achieve and maintain decay heat removal via the S team Generator System. EVST sodium hot leg temperature is monitored to enable the operator to make manual 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. 57 Nov. 1980
- 7. 5- 33d w - -. , - -
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] N TABLE 7.5-1 INSTRUMENTATION SYSTEM FUNCTIONS AND
SUMMARY
Measured System Parameters Instrument Purpose Measurement _.L_o ation Flux Monitoring Source Range BF Thimbles on peri 3hery # gured 57l 52l 3 vessel Determines or Provides: 1.. Flux status at shutdown, tilde Range
- Fission Chambers Thimbles on peri:hery of guard startup and power levels vessel
- 2. Signals to PPS logic Power Range
- 8-10, Compensated Thimbles on periphery of guard (except source range)
Ion Chamber vessel
- 3. Signals for reactor and plant control (D.C. linear power rangesi y 57I
- 4. Signals for display, m annunciation and recording b
A Heat Transport Primary / Reactor Inlet Pressure Element Cold leg pri,r.ary loop PPS and display 50l Intemediate Pressure
- PHTS performances Loops Primary and Inter- FM Flowmeter Cold leg of primary and inter- PPS, Plant Control and Display, rOl mediate Flow
- rediate loops (hot leg in inter- PHTS perfor-ance mediate loop 2)
IHX Primcry Outlet Thermocouple Cold leg piping nearest to IHX ,lant Control System (PCS), PPS, Temperature
- primary outlet and Display Primary and Inter- Resistance Primary and Intermediate hot and Surveillance, display and use to r,g mediate Hot and Temperature cold leg calorimetrically calibrate PM Cold Leq Tempera- (RTD) flowmeters ture Primary and Inter- Pressure Elements Drainline from discharge piping of Surveillance, display and mediate Pump the loops's sodium pump monitor differential pressure Disharge Pressure between primary & intermediate 50I loops PHTS performance Intermediate Pressure Elements Intermediate between IHX & Surveillance, display & monitor IHX Cutlet Superheater differential pressure between r;q l Pressure intermediate loops 5$
2
TABLE 7.5-1 (Continued)
..easure: l j Syste- Para-ate : I Instrrent '
Measu e ent Lecatier Purpose Heat Transport Inte- edie:e Pu p Pressure Ele ents Pipes between evaporator and Wp Tri aryj Inlet Fre35 re inlet Display pu p performance I-te--ediate L;;;; (cont'd) Inte cediate Expan- Level Pr:be Inter ediate expansion tank Display-inte mediate loop sodium sien Tank 5:diu-Le.el invr 'ory and alam Evaccrator 5:dium Themoccuple Downstream where the two evaporator Cutlet Tem;e ature* PPS and display outlets join into the header j 5:di e P eps Seci r Level Level Probe Pump Tank Display-used for sodium inventory and purp protection (alarm) Prir sry/:r ter- e- Tachometer diate P ep Speed
- Main Shaft of each pu p PPS, display, pump speed control, perfomance 49 Por.y P' tor hr.ning Speed Switch Pumps Display, performance Dia;restic Instru- Various Purps cntatien Display, purp performance Steam Generator Sodium Ficw Venturi _
M Superheater sodium outlet (1 loop) Display & superheater & evapo-rator performance 5:dium Te perature Themoccuple Superheater evaporator outlet Display & steam generator (31 cops) perforrance evaluation Sodium Pressure Pressure Element I loop-superheater inlet, outlet Display & steam generator 49 (bothlegs)andevaporatoroutlet perfornance evaluation (one leg) feedwater Flow
- Venturi Inlet line to steam drun F . display & steam generator (feedwater) formance evaluation Superheat Steam Venturi Dutlet of each superheater Pt5. display & steam generator Flew +
(ste3m) rerformance evaluation Steam Drum Drain Orifice stean Drum Drain line for each F1ew steamdrum Per'omance evaluation E< ater 'nlet Venturi Inlet to one evaporator (1 loop) F '. Perfomance evaluation 4
- 2. 8
?
m
@D N@
e O O
i l ) TABLE 7.5-1 (Continued)
~ ~
i Measured 4 System Parameters Instrument Measured Location l Purpose I Steam Generator Feedwater Temp.* RTD Steam drum inlet (feedwater) PPS, display & steam generator j (cont'd) perfcrmance evaluation Recirc. Water Temp. Thermocouple Recirculation pump discharge Performance evaluation lj header i Steam Temp. Thermocouple Outlet header from steam drum Performance evaluation i (steam) I I Superheat Steam RTD Superheater outlet line (steam) PPS and Cisplay { Temp.* i \ Evap. & Superheater RTD Inlet & outlet nozzles for 1 Display & performance evaluation j Inlet-Outlet Temp. evaporator & superheater (1 loop) j Blowdown Temp. Thermocouple Blowdown line for each steam drum Perforr.ance evaluation ! u Feedwater Pressure Pressure element Inlet line to each steam drum Display and steam ger.erator j g performance i i y Steam Drum Pressure
- Pressure element Appendage from steam drum PPS and display Recirc. Pump Oat- Pressure element Appendage from recirculation l let & Inlet pres- pump discharge and suction Performance evaluation i
sure. header. Superheat Steam
- Pressure element Locp output steam line 29
{ PPS and display j Pressure ! Evaporator and Pressure element Inlet nozzle for 1 evaporator i Performance evaluation Superheater Inlet and superheater (1 loop) Pressure Evaporator and Pressure element Outlet nozzle for evaporator SWRPS and performance evaluation ! Superheater Outlet and superheater in each loop j Pressure Steam Drum Level
- Differential Differential pressure across PPS and display Pressure Element steam drum 2
ON 4
.8".
.c I @
i NN l Ch O l i I I i, _ _ _ _ _ _ _ _. _- -m -- _ _ _ - __ __
TABLE 7.5-1 (Continued)
~~~
Measured ' Sy s tem Pa ram ters Instrument Measured Location Purpose Reactor and Core Sodium Exit Thermocouples 51 vessel Ins tru- Temperature Selected fuel and blanket Display and control - core out-mentation assewalies. let temperature Core Peripheral Thermocouples Core periphery - 2 locations lemperature Display - Design verification t Upper Internals Thermocouples Temperature Parts of upper interaal structure Dispiay - design verification. 6 locations predict stress on various 40 components Sodium level above Level Probe Reactor vessel plerm PPS and Coatrol, Display Core
- N.
on 40 2, 57, 41l Upper Internals Vibration Element 4 Biaxial on parts of Display - measure vibr tions j Movement appropriate structure induced by sodium flow Fuel Failure Cover Gas Gamma Gamma Spectroneter Sampling in RSB Monitoring Activity Detect each instance of fuel clad failure and char.icterize failure Delayed Neutron BF3 Counter Shielded moderator assembly adjacent Detect fuel in PHTS Monitoring to each of the PHTS hot leg pipes Tag Gas Isotopic Mass Spectrometer Gas tag sampling traps in RSB Locate failed fuel Composition yg Leak Detection liquid Petal Contact detectors various locations in sodium identify locaticn of licuid
< o to Gas leab cable detectors circuits metal to gas leols for 3
aerosol monitors c on t i rar..e .cru i11once of g .C1 liquid retel syste s t,ow.Nrirs g, .- _.. .__ 28 ou l l 9 O O
A r Y N,g) w) TABLE 7.5-1 (Continued)
~
Measured System Paraneters Instrument Measurement location Purpose Leak Detection Intermediate to Level probe IHTS cxpansion tank Detect leak in IHX (cont'd) Primary Leak l Steam Generator Hydrogen and 0xygen Sodiun exiting either or both Leaks detectors superhester outlets Steam Generator. Hydrogen e d Oxygen, Sodium filled vent line from Detect small water-sodium and Leaks detectors ' superheater steam-sodium <aak in steam generator, ider.?if- leaking Hydrogen and 0xygen Sodium filled vent lines from module, provide signal for ' Steam Generator either combined evaoorator Leaks detectors vents or evaporator A vent perator action i S dium exiting combined Steam Generator Hydrogen and evaporator cutlets or evap-Leaks Oxygen Detectors 47 _ orator A outlet. Sodium-Water Rupture discs Sensors Downstream of rupture discs- FPS and SWRPS initiation Reaction Pressure Operation
- before reaction products N Relief separation tank Downstream of rupture discs - SURPRS initiation T Rupture discs 5*" * ' Cf 'C 5 di"" d"*P t^"k M 43 Operation Rupture discs Pressure Element Gas space between rupture discs Surveillance of discs integrity SWRPS equipment Thermocouples Surface temperatures of reactor Surveillance l Temperature products separation tank,
# centrifugal separator, drain l 6 ) tank and hydrogen igniter -
3 ( D Separation Tank iressure Reactor products separation tank Surveillance 29 M Pressure Elements , 3s :. Evaporator Water and sodium dump Level, Pressure, Temperature Evaporator water and sodium dump tanks Surveillance 2 tanks level. Elements i
$ pressure and cc y temperature
.__
- Safety Related.
r
l TABLE 7.5-2 REACTOR AND VESSEL INSTRUMENTATION Measured Instrument Parameter Location Purpose Thermocouple Core Exit Sodium One at each of 30 Control and surveil-Temperature selected fuel and lance - Core outlet blanket assemblies temp. 275 additi.onal Surveillance and Diagnostic - 57 locations Distribution of at selected fuel temperature across and blanket assem- the core blies. 51 Thermocouple Core Peripheral Two spaced loca- Design Verification - Temperature tions on the core Distribution temp. periphery around the core O 40 Upper Internals Six on parts Design Verification - Temperature of the upper Distribution of temp. internal structure to predict stress on various components Sodium Level Sodium Level Four short units Protection and Con-Detector above the core distributed trol - Measures the equally around operating level of periphery the sodium in the reactor One long unit Monitoring - Measure near one of the the sodium level from four short ones operating level down to minimum safe level 41l Vibration Upper Internals Four biaxial on Design Verification - Detector Vibration parts of appro- Measure vibrations priate structure induced by sodium flow Amend. 57 Nov. 1980 l 7.5-39
R e REFUELING
,$ CONSOLE SOURCE RANGE DISPLAY (TYPICAL OF 3 EXVESSEL DET SIGNAL PA BUFFERS CHANNELS) CONDITION ANNUNCIATION DATA HANDLING (T M C DET PA --
MS.V. SH TDOWN ANNEL ! LOGIC
, DISPLAY LINEAR U w -
BUFFERS ANNUNCIATION 6
- COUNTING -
DATA H ANDLING POWER RANGE (TYPICAL OF 3 EXVESSEL SI Al DET JB SHUTDOWN CHANNELS) CONDITIONER LOGIC II DISPLAY DET - DETECTOR BUFFERS PA - PREAMP MSV - MEAN SOUARE VOLTAGE JB - JUNCTION BOX ANNUNCIATION CONTROL SYSTEM oN 58 DATA HANDLING P.
$ Figure 7. 1. CRBRP Flux Monitoring System Block Diagram
s i SOURCE RANGE - LOG CPS PROPORTIONAL COUNTER ~ 40 CPS /NV W10E RANGE - LOG COUNT RANGE - LOG % POWER FISSION CHAMBER WIDE RANGE - LOG MSV - LOG % POWER 'PO ER W10E R ANGE - LINE AR PWR. COMPENSATED 10N CH AMBER 4 PWR R ANGE - 1 IN. PWR. v. b NEUTRONS PER SQ. CENTIMETER PER SECOND (NV) AT DETECTOR 2 4 5 10'l 100 101 10 10 3 10 10 10 6 10 7 108 109 1010 1011 l t i t il i i t il i i i ti i t ill i t iil i i til i i til s i al i i iil i s iil i t iil i i iil j i i i ng i l ig i iij i s iij i s is j i isig i isig i i i ij i i s ig i s isj i i sij i i iij 10-9 10-8 10-7 10 6 10-5 10'4 10 3 10 2 10-1 10 0 10 1 10 2 10 3 PERCENT REACTOR POWER oN < ro
- i Figure 7.5-2. CRilRP Flus Monitoring System Instrument Range Coverage 5'
8 El O O O
m 7.7.1.3 Primary and Secondary CRDM (Control Rod Drive Mechanism) Controller I T and Rod Position Indication V The Primary Control Rod Drive Mechanism Control System transforms the bulk 3 phase power into the pulsed DC necessary to operate the Control Rod Drive Mechanism in response to input commands from the Reactor Control System. Interlocks and permissives are provided to prevent operating sequer.ces of the control rods which would damage the equipment, and assure that the rods are maintained in the banked configuration required to maximize core performance. Rod Position Indication is provided redundantly for each rod to permit the operator to verify the reactivity status and operation of the control sys-tem. The Secondary CRDM Controller and Rod Position Indicators are described in Section 4.2.3. 7.7.1.3.1 Primary CRDM ontrol The control rod drive mechanism is actuated by a 4 pole, 6 winding reluctance stepping motor. The mechanism lead screw has a thread pitch of 0.6 inch, and moves 0.025 inches for each pulse to the drive stator. A block diagram of the drive system is shown in Figure 7.7-4. Driving power is supplied from the site power system through redundant motor-generator sets, Reactor Shutdown System scram breakers, a 3 phase to 6 phase transformer, and banks of silicon controlled rectifiers (SCR's) in the individual controllers to the stator windings of the CRDM. The primary rods are divided into 2 groups. One group of 3 startup rods responds only to single rod manual control and while operating during (V ) reactor operation is normally fully withdrawn. A group of 6 control rods respond to manual control or to an analog signal from the Reactor Control 57 System. Rod speed demand limits are included in the reactor controller as well e rod speed limits in the individual Primary CRDM Controllers. Rod block interlocks are included in the "OUT" demand input as shown on Figure 7.7-5. When the Reactor Shutdown System initiates a scram, the " Scram Breakers" open and interrupt the power to the Primary CRCM stator coils; the rotor collapses and disengages the rollers from the lead screw; and the C90M drive train falls under the force of gravity and the scram assist spring to insert the control rods into the core. Failures within the seq-uence and controller units cannot prevent removal of the power required to hold the CRDM's in the withdrawn position. The componen*s are described below. 19 Motor-Generator Set Dual M-G sets provide the 3 phase power for CRDM operations. 57 When a latch signal is received at the voltage control, th'e output voltage of a generator is increased. The M-G sets for the primary rods use a 200 Hp motor and a 150 Kw generator. p 7.7-4 Amend. 57 Nov. 1980
Mechanism loads are shared by the. two M-G sets; however, either M-G set has the capacity to power the entire load of the primary system. Controls are provided to synchronize the two M-G sets. The rootor-generator sets are designed to provide sufficient inertia and volt-age control to prevent rods dropping in the event of power dips of 0.3 seconds or less. Generator output circuit breakers provide the necessary electrical
- 57) protection for the generators and for system maintenance Power Supplies and Transformers 571 Since the CRDM controllers use 6 phase AC power, one 3 phase to 6 phase transformer is provided for the primary rods. The 'rans-fonner includes appropriate secondary side surge protection.
O l l l O 7.7-4a Amend. 57 Nov. 1980
Each CRDM controller requires contt01 power to operate the interface circuitry, programmer, gate drives, internal interlocks and display (3 equipment. As shown on Figure 7.7-4, redundant AC power sources Li 57 energize redundant DC logic power supplies whose outputs are auctioneered. This design prevents failure of a power supply from causing a rod to drop. The power supplies are sized to provide sufficient capacity for all of the CRDM controllers in the primary group. Transforme, isolation, including grounded Faraday shields, is used to prevent failures from propagating into the controller electronics. CRDM Motor Controller The CRDM Motor requires DC energization of coils (in the pro-per sequence to develop the required setpoint motion. The sequence of coil energization for rod motion is in a two coil-three coil 57 sequence. Thus a forward step is produced each tire a leading coil is energized and also when a trailing coil is de-energized. To reverse the motion, the sequence is reversed. The CRDM Controller uses six SCR's for each stator coil to half wave s rectify the 6 phase AC input power and supply DC output to a stator coil. All six SCR's for a stator coil are turned on by one gate drive unit. The Controller incorporates the logic necessary to correctly sequence the gate drive units on and off, thereby sequencing the coils in appropriate order. Separate controllers (3 f
) provided for each individual mechanism.
input or output logic errors are detected. Holds are provided when are 57l In Single Rod Control Mode, the input circuitry to each controller accepts on-off inputs for IN, OUT, and HOLD commands and provides the sequencer with an IN pulse train, OUT pulse train, or HOLD DC output. The IN command steps a single rod down in the core at a predetermined rate. The OUT command steps a single rod up out of the core at a predetermined rate (not necessarily the same as the IN rate) and the HOLD command maintains the rod in its present position (no motion). The input circuitry also incorporates adjustable speed settings for the IN, OUT, and LATCH modes of CRDM operation and assures that an IN command takes precedence over an OUT command. In addition to the adjustable speed settings, the controller provides an independent speed limitation which has a separate clock and power supply from that used by the input circuitry. If the input circuitry called for a speed greater i 57l than 10% above 9 inches per minute due to a postulated failure, the ! speed limiter circuit will place the rod in the Hold Mode. ! (o) Amend. 57 7.7-5 Nov. 1980
In any automatic control mode, or in Group flanual mode, the mechanism controllers are operated in sequence one step at a time to keep the rod bank in required alignment. The sequence rate and direction are determined respectively by analog and digital signals from the reactor control system. If the selector sequence rate is higher than a predeter-mined trip paint, an overspeed detector will alarm and place the controllers 37 in HOLD. A funstional block diagram of the control is shown in Figure 7.7-5. Hold Bus A Hold Bus Power Supply and transfer select circuitry are provided to allow any controller to be replaced without a plant shutdown. In the event of a controller failure, the mechanism controller in 57 question can be switched out and transferred to a Hold Bus. Power to the Hold Bus Power Supply is provided downstream from the scram breakers. This ensures that if a scram is initiated, a rod on the Hold Bus will also scram. 7.7.1.3.2 Rod Position Indication System Two independent Rod Position Indicating Systems are provided for each control rod: An Absolute Position Indication System (ARPI) and a 57 Relative Position Indication System (RRPI). These systems assure that the plant operators can continuously determine the position of the control rods. The ARPI provides a direct measurement of rod position at any time and, unlike the RRPI, does not require re-zeroing a,;er a scram or temporary loss of power. The system is solid state, utilizing ultrasonics and magnetics to provide a D.C. output indicative of rod position. The sensor for this system consists of a tube extending down from the top of the motoi tube and into the inside diameter of the PCRDM lead screw. A nickel-cadmium wire is stretched axially through the tube. As the lead screw translates, the flux from a torroidal magnet mounted on top of the lead screw intersects the wire at a point indicative of the rod position. Electrical pulses sent down the wire generate magnetic fields which, when they intersect the flux of the lead screw magnet, causes a torsional strain creating a sonic pulse which travels from the point of flux intersection upward. The sonic pulse is detected at the top of the wire, and the time of propogation is measured electronically. This propogation time is converted to a D.C. signal which is analagous to rod position. This signal is read out on the main control panel by rod top and rod "7 bottom indicator lights and a vertical bar graph indicator. It is also used to operate the rod out of alignment alarm and rod control interlocks. Amend. 57 7.7-6 flov. 1980
\
G The Relative od Position Indication System provides a digital rod posi-57 t on indication on a CRT at the Main Control Board. Two pairs of magnetic coil pick-ups are mounted within each stator jacket above the stator and on opposite sides. A 6 pole magnetic section is attached to the mechanism rotor and rotates in the plane of the pick-up coils. Voltage pulses caused by the move-5/l ment of the poles in the proximity of the pick-up coils are sent to the PDH&DS. The resolution of the indicator is +0.1 inch. Unlike the Absolute Position
~
Indication System, this system must be reset after each scram and in the event of a powe, failure reset af ter power is restored. The pulses are also 57 counted by an odometer type readout in the rod control equioment room. l.7.1.4 S_od_ium Flow Control System The Sodium Flow Control System consists of six controllers used to drive the three primary and three intermediate sodium pumps. Each controller consists of a cascade system with an inner loop using speed as the feedback signal and an outer loop based on a flow feedback signal. The flow control
; range is 30 to 100% of rated flow. The flow setpoints are generated either 41 manually or by the Supervisory Control.
Figure 7.7-7 is a block diagram of the flow / speed control loop which is typical of the six controllers in the system. The Speed Control System h O is an inner loop and ; sed pump speed, which is sensed via a pump shaf t mounted tachometer, as the feedback variable. The Speed Control System is limited internally by the torque limit circuit which sets both the accelerating and decelerating torque of the variable speed pump drive. The demand to the Speed Controller is set by the FLOW / SPEED Mode Select Switch. In the Speed Mode, pump speed is set by a manually adjusted 44l potentiometer; in the Flow The Flow Controller usesMode, pump median the filtered, speed isselect set by sianal the Flow Controller. of three available redundant flow meter buffered PPS outputs as the feedback signal . This signal, along with the flow demand, is used to generate the error signal which is com-pensated through the Control Compensation Network and then limited by the High Speed Limit Circuit prior to being used as the speed demand signal. The demand to the Flow Controller is set by the MAN / AUTO Select Switch. In the automatic mode, the demand comes from the supervisory control, while in the Manual Mode, the demand comes from a manually adjusted potentiometer on the control panel. 41, / Amend. 57 (]) Nov. 1980
7.7.1.5 Steam Generator Feedwater Flow Control System The Feedwater Control System provides for the controlled supply of feedwater to the Steam Generator System steam drums. Flow of water into the steam drums is automatically controlled to maintain water in the drum within specified limits. The range of water level is based upon the ability of the steam separators to function properly. Primary control is effected by mod-ulating a Feedwater Flow Control Valve at the inlet to the steam drum. In addition, Feedwater pump speed is varied to maintain the pressure drop across the Feedwater Flow Control Valves at a value that does not change excessively with load. This reduces the operating requirements of the Flow Control Valve and improves feedwater pump efficiency. Control logic for the Feedwater Flow Control System is shown in Figure 7.7-1. 7.7.1.5.1 Feedwater Flow Control Valve Control System The Feedwater Flow Control Valve is positioned via a conventional three-element cascade control system which consists of drum level and feed-water flow controllers. The drum level controller compares measured drum level with a pre-determined setpoint value. The resulting error signal is summed with a sig-nal representing drum steam flow and constitutes the input (i.e., flow set-point) to the feedwater flow controller. The output of the feedwater flow controller is the position demand signal to the Feedwater Flow Control Valve. Instrumentation required by this control system is obtained as e Steam Drum Level - Water level is measured by a differential pres-sure transmitter which senses the difference between the pressure resulting from a constant reference column of water and the pres-sure resulting from the variable height of water in the steam drum. The measurement is density compensated. e _ Steam Flow - Steam flow is sensed at a flow element in the outlet line from the superheater by a differential pressure transmitter. The differential pressure signal is compensated for temperature and pressure variations and linearized to provide a mass flow signal. e Feedwater Flow - Feedwater flow is sensed at a flow element in the i iiet line to the steam drum by a differential pressure transmitter. The differential pressure signal is corrected for temperature vari-ations and linearized to provide a mass flow signal . 7.7-8 O
, - , ,-~
< \ ,/Y %/
Temp. Temp. High Magnitude
+
+ Flux Flux Dead
- m Dead Comp.
~
Flux Dead Comp. Zone And to CRDM Zone Net. Limit Zone pFOS ][- _ Net. Saturation Control Supervisory + Up Direction Control an Set o M n/ Auto
+ Down j
N Average N . 27 Core
/ Median / 3 Primary 7 and . Exit Flux Signals y Reject . Thermocouples \ Select Ny N
SE seo. G- Figure 7.7-3 Block Diagram of Reactor Control System 8$0
7 5
"A" SOURCE OTOR ST ARTE R -
O-480 VAC GE N SET f HOLDBUS I % POWER 3e TO 60 SUPPLY
- VOLT l T R ANSF ORME R REGULATOR & +
"+ CONTROLLER !
L ATCH CONTROL - GEN K R.' 4 = CRDM BKRS CONT 1 ITC ING
-+ VOLT j "
34 , REGUL ATOR & + l g CONTROLLER I TYPICAL OF 9 I I l 8 N - 480 VAC GE N SET +
- e CONT 9 h 128 LOGIC SUPPLY _
)l, VAC A V ~
PROGRAM
& CONTROL 120, ,
128 LOGIC SUPPLY VAC B V ACi l 1T ME CH ANISM INPUT COMMANDS.. H F AUT0 ENABLE FROM RE ACTOR SE LE CTOR CONTROLLER EF < n>
- 3
~. e 8 kl Figure 7.7-4. CRDM Controller and Power Train for Primary Rods e O O
) ,
) )
%,i 's.,/ L.J S ERROR B'J F F E R VOLTAGE ALARM CONTROLLEO RESET h AMPLIFIER AMPLIFIER OSCILL ATO R REACTOR R1 _
DVERSPEED CONTROL
'Q ' - -
' VC0 '
ERROR SIG N A L [ t-
'[ PROTECTOR
_l R2 v R00 SELECT R ATE SPEED ENABLE ENABLE j GATE
# +
C 2 CLOCK FREQ. R00 : IN0lV. ' B TO IN0lV100AL r O SELECT 4 fAECHANISM CLOCK : DNE : , TO CONTROL A 5 CONTROLLERS ROOM GROUP r ; SHOTS ~ 2 THRU 6 M00t SWITCH 6 - #
- I n o REACTOR IN WITH0 R AWAL CONTROL STOPINTERLOCKS SIGNALS OUT oN 52 F
$ i.n ON Figure 7.7-5. lilock 1)iagram of I'rimary Rod Group Control
l l O' 57 FIGURE 7.7-6 HAS BEEll DELETED O I \ Amend. 57 7.7-25 floy- 1980
7.8 PLANT DATA HANDLING AND DISPLAY SYSTEM O 7.8.1 Design Description 57l The Plant Data Handling and Display System PDH&DS supports plant operations and performance by monitoring, limit checking, trending, and dis-playing plant information. It supplements other monitoring and displaying 571 systems including Plant Annunciation and Plant Control. The PDH&DS performs only diagnostic and informative functions and its operation is not a require-ment for startup, operation, or shutdown of the plant. However, additional 571 requirements may be placed on the plant operations should the PDH&DS be inoperative. Specific functions of the PDH&DS include: e Monitoring of plant variables and alerting the operator when 57 selected variables exceed predetermined limits. e Recording of the operating history of the plant. e Performance parameter calculations including: plant and equipment calorimetries 57 plant protection system channel output monitoring shutdown margins (] control rod worth V core assembly exit temperatures 37l reactivity calculations sodium inventory calculations 57 component efficiencies e Providing Cathode Ray Tube (CRT) display units in the control room to present pertinent plant data for surveillance of plant pro-tection and control systems, auxiliary systems and balance of plant. 1 e Supporting of refueling operations by providing capacity for long 57 term core component inventory. 57 l e Displaying group annunciation of measurements to reduce the number of trips by plant personnel to read local indicators. e Providing a mechanism to forewarn the operator of potential harmful conditions. Examples of these include high bearing temperature, detection of small sodium leaks and radiation levels. If the condition deteriorates further, the operator will be warned by the annunciator system. U Amend. 57 7.8-1 Nov. 1980
57 O e Providing pre and post trip information for review. 57 l e Providing for acquisition of data for design verification of plant components. 57 7.8.2 Design Analysis The PDH&DS is inherently designed for high system availability. The sensor measurements being monitored are divided into two groups according to their importance to plant operation. Significant information (Group 1) will have the capability to be processed, recorded, and displayed by more than one piece of equipmerit. The less important beneficial information (Group 2) may be processed, recorded and displayed by only one piece of equipment. When one piece of equipment fails, significant information is n t affected but the failure may limit the amount of beneficial infonnation 57 that can be processed, recorded and displayed. The main processing part of this system is located adjacent to the control room in the computer room. Information generated by the PDH&DS is presented in the control room. The data acquisition components of the system are located near sensor local panels. The data acquisition components inultiplex sensor signals to reduce the number of control room panels and associated cabling. Although the PDH&DS is designed for high availability, it should be emphasized that the system performs no direct safety or protection function and is not essential for plant operation. Operating procedures of systems which normally use PDH&DS capabilities are written te 111ow operation 57 of the plant with manual data recording and calculations. O Amend. 57 7.8-2 Nov. 1980
/9 \v/ 7.9 OPERATING CONTROL STATIONS 7.9.1 Design Basis A control room is provided from which action can be taken to operate the nuclear power station safely under normal operation and off-normal condi-tions and to maintain it in a safe condition under postulated accident con-ditions. Adequate radiation protection is provided to permit access and occupancy under Ertremely Unlikely Fault conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the postulated accident. The control room provides protection from substances such as sodium oxide which might be released to the local environment under Extremely Unlikely Fault conditions. The basic cri te ria for inclusion of displays or controls in the control room shall be: e The displays or controls necessary to support all normal plant operating conditions; e The displays and controls necessary to respond to off-normal or casualty conditions which impact on power operations capability; e The displays or controls necessary to prevent potential radiological ('] hazards to offsite personnel; V e The displays necessary to the operator for detection of fire hazards; or e The display and controls necessary to prevent potential damage to the plant. The control boards are arranged functionally based on normal and off-normal operational considerations to minimize the number of operators required and to enhance the capability of the operational personnel to monitor and assure the safe status of the plant during all operations. Remote control stations are provided outside of the control room to shut the plant down and maintain it in a safe condition assuming loss of control 57lroomhabitability. Access to the control room is controlled by card key to assure that only qualified persornel use the equipment provided to monitor and maneuver the reactor plant. 7.9.2 Control Room 7.9.2.1. General Description The Control Room is located in the Control Building at approximately grade level . The Control Room proper occupies approximately a space of p) Q Amend. 57 7.9-1 Nov. 1980
57l 70' x 75'. The remainder of the floor is devoted to auxiliary services for the control room operat ' personnel . The Control Building provides the necessary structural ar._ _-mospheric protection to allow continued habit-57l ability that collectively satisfy CRBRP General Design Criterion 17 These features are summarized in Section 3.A.3. The indicators, annunciators, and controls included in the Control Room provide the capability to operate the plant through all normal opera-tional sequences and to respond to off-normal or emergency conditions without continuously manned remote stations. 7.9.2.2. Control Room Arrangement The control room is arranged to provide an effective interface between the plant and the operating personnel (refer to Figure 7.9-1). Fre-quently used safety related instrumentation and controls are located on the
- 57) Main Control Panel. This equipment is grouped by operational category to assure that determination of plant condition and action to correct the condi-tion are in close proximity. Less frequently used equipment and certain electronic equipment for which access control is desired are located in the rear panel area.
7.9.2.3 Main Control Board Arrangement As shown in Figure 7.9-1, an open U-shaped main control panel is provided. The main control panel and the area it encloses form the central operating area for the plant. Equipment on the main control panel is arranged functionally and according to the power generation flow path. From left to right, the main control panel sections are as follows: emergency systems, plant protection system and engineered safety features, plant control and primary heat transport systems, intermediate heat transport and steam generator systems, steam generator auxiliary systems, turbine system and generator and switchyard. The equipment is arranged with annunciators at the top and display, controls, and switches in functional groups on the vertical and sloping bench sections. The size and arrangement of 57 equipment is based on the following guides: e Displays, annunciation, switches, and control necessary to operate the plant without continuously manned remote stations are located on the main control panel or displayed through the 57 Plant Digital Data Handling System and the cathode ray tube (CRT) displays. e Graphic or mimic disp 10ys are provided where warranted to enhance the operator / plant control interface and minimize the chances for inappropriate operator action. e Physical separation of redundant safety related instrumentation equipment is incorporated. Amend. 57 Nov. 1980 7.9-2
e Physical, color, and geometric differentiation of displays and ,, controls mounted on the board is provided to assure ease of (V ; recognition by operating personnel and minimize the chances for inappropriate actions, o Arrangement and design of displays and controls is specified to provide arrays which permit determination of proper alignment at a glance, where practical. e Modular design of switches, controls, and indicators is used to permit ease of maintenance and minimum inter ference with operation. The equipment included on the main control board is summarized below (refer to Figure 7.9-1 and Table 7.9-1). The arrangement of the instrumentation and control devices on the main control panel is as follows: Sections 1 & 2 - Emergency Systems o Emergency Chilled Water o Emergency Plant Service Water Sr-tion 3 - Plant Protection System and Engineered Safety Features o Reactor Shutdown (") v o Containment Isolation o Steam Generator Auxiliary Heat Removal System Status o Sodium Water Reaction Pressure Relief System Status o Sodium Dump o Control Room Heating, Ventilating, and Air Conditioning o Containment Instrumentation o Flux Monitoring Section 4 - Plant Control and Primary Heat Transport Systems o Primary and Secondary Manual Scram Switches o Supervisory and Reactor Control o Reactor Instrumentation o Rod Control and Rod Position Indication Primary Heat Transport p 57 V Amend. 57 7.9-3 Nov. 1980
Section 5 - Intermediate Heat Transport and Steam Generator Systems o Intermediate Heat Transport o Steam Generator o Feedwater Section 6 - Steam Generator Auxiliary Systems o Condensate o Auxiliary Feedwater o Protected Air Cooled Condenser Section 7 - Turbine System o Turbine Control Panels o Turbine Instrumentation o Turbine Steam Bypass
; c Circulating Water I
o Secondary Plant Service Water o Secondary Manual Scram Switch Sactions 8 & 9 - Generator and Switchyard o Grar hic Arrangement of High Voltage AC o AC Bus Circuit Breaker Control o Generator Syncroscope The Following instrumeltation and crntrcl panels, while not a part of the Main Control Panel, demand rapid opere;or response and have been arranged to permit operator scanning fron tha Main Cantrol Panel: o Failed Fuel Monitoring o Sodium Leak Detection o Sodium Fire Detection 57 o Non-Sodium Fire Detection O Amend. 57 7.9-4 Nov. 1980
g o Control Building Fire Detection o Emergency Diesel Generators o Switchyard and Station Electrical Distribution o Di: ect Heat Removal Service The layout of Section 3 of the main control panel is designed to minimize the time required for the operator to evaluate system perfonnance under accident conditions. Deviations from predetermined conditions are alarmed and the status of automatic safety systems is alarmed and/or indicated so that corrective action may be taken by the operator. The control room also includes control and instrumen'ation equip-ment that is used infrequently or for which controlled access is desirable. Included in this control room back panel area are power distribution, chilled water, containment instrumentation, recirculating gas, heat transport, steam generator, heat ventilation and air conditioning, annunciator electronics, turbine, balance of plant, plant control, plant data handling and display system multiplexers, flux monitoring, radiation monitoring, reactor shutdown 57 and containment isolation panels. n (J \ 7.9.2.4 Main Control Panel Design 57 The Main Control Panel is an open U-shaped, stand up vertical panel as shown in Figures 7.9-1 (plan view) and 7.9-2 (side view). There are 3 significant features of the control board mechanical design: seismic capabil-ity; separation of redundant safety related equipment and wiring; and modular construction of switch, indicator and control equipment. 1 (V Amend. 57 l flov. 1980 l 7.9-5 l
57l Since the Main Control Panel includes safety related equipment, the 52 l sections including this equipment are designed to Seismic Category I. Struc-tures, wiring, wireways, and connectors are designed and installed to ensure that safety related equipment on the control panel remains operational during 57 and after the SSE. The Main Control Panel is constructeJ of heavy gauge steel within appropriate supports to provide the requisite stiffness. l Within the boundaries of the Main Control Panel Sections, modules 5/;l are arranged according to control functions. Fire retardant wire is used. Mooular train wiring is formed into wire bundles and carried to metal wire ways (gutters). Gutters are run into metal vertical wireways (risers). The risers are the interface between external wire trays feeding the panel and 57l Main Control Panel wiring. Risers are arranged to maintain the separated routing of the external wire trays. (See Figures 7.9-3 and 7.9-4). Mutually redundant safety train wiring is routed so as to maintain a minimum of six inches air separation between wires associated with different trains. Where such air sepration is not available, mechanical barriers are provided in lieu of air space. 67 7.9.3 L_ocal _ Control Stations Local control panels are provided for systems and components which do not require full time operator attendance and a' e not used on a continuous basis. In these cases, however, appropriate alarmi are activated it, the Control Room to alert the operator of an equipment malfunction or approach to an off-57 normal condi tion. 7.9.4 Communications Communications are provided between the Control Room and all operating or manned areas of the plant. In addition to ?ublic address and interplant comma-41 g aications and the private automatic exchange (ued for in-plant and external communications) vided. a sound powered maintenance communic: tion jacking system is pro-57 I and other TVA nenerating plants is also provided. Redundant and separate meth 7.9.5 Design Evaluation Safe and continuous occupancy of the Control Room during normal and off-normal conditions is provided for in the design of the Control Building. The probability of the Control Room becoming uninhabitable due to fire or other cause is considered extremely remote. However, in the event the Control Room must be vacated temporarily, the reactor plant can be brought to and main-tained in a safe shutdown condition for an extended period of time from local V stations outside the control room. These local stations have been disco w d n Section 7.4.3 as the Remote Shutdown System. O Amend. 57 7.9-6 Nov. 1980 1
- - - . . -- _ . . - - = - . . - - . . - - - - . - - - . . _ . .
J l O 1 l l 57l PAGES 7.9-7 and 7.9-7a HAVE BEEN DELETED : i 1 1 1 1 l i
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1 i i i i
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\ i e Amend. 57 7.9-7 ' Nov. 1980 . I
Table 7.9-1 CONTROL ROOM ARRANGEMENT U _Itam 41 l i . Reactor Support Systems
- 2. Reactor Support Systems
- 3. Engineered Safety Systenis
- 4. Reactor Primary Heat Transport Systems 41
- 5. Intermediate Sodium Heat Transport Systems
- 6. Steam Generator and Associated Systems 7 Turbine Systems
- 8. Generator Systems
- 9. Synchronization and Main Unit Control 41 j 10. Switch Yard and Station Electrical DistributLa
- 11. Emergency Diesel Generator Panels
- 12. Reactor Operator
- 13. Sodium Leak Detection Panel
- 14. Desk
- 15. Chair
- 16. Seismic Instrument Panel
- 17. Flux Monitoring
- 18. Process Radiation Monitoring
- 19. Area Radiation Monitor'ng
- 20. Weather Station Space
- 21. Phones (Station Comunications)
- 22. HVAC System Recirculating Air Status Cabinei
- 23. HVAC Control Rack 41
- 24. Generator and Transformer Protection tO Amend. 41 7.9-8 Oct. 1977
Table 7.9-1 (Continued) Item
- 25. Turbine / Generator Supervisory Panel
- 26. Turbine (Electro Hydraulic Control) Equipment Bays
- 27. BOP Auxiliaries 57l 28. Non-Sodium Fire Protection Rack
- 29. Sodium Fire Protection Zone Indicating Panel
- 30. Heat Removal and Conditioning Logic Rack 31 Steam Generator Logic Rack
- 32. Chilled Water Control Cabinet 6 33. Auxiliary Liquid Metal Control Cabinet
- 34. Remote Annunciator Cabinets 6 l 35. Steam Plant Conditioning Rack
- 36. Steam Plant t.ogic Rack 57 l 37. Not Used
- 38. Termination Racks
- 39. PPS Containment Isolation Instrumentation Racks 57 l 40. Not Used
- 41. Primary PPS Buffers
- 42. Primary PPS Termination Cabinet
;3. Primary PPS Comparator Panels
- 44. Primary PPS Logic Racks
- 45. Primary PPS Isolation Racks
- 46. Secondary PPS Buffers
- 47. Secondary PPS Termination Cabinets
! 48. Secondary PPS Comparator Panels 1
41 7.9-9 Amend. 57 Nov. 1980
_s Table 7.9-1 (Continued) Item
- 49. Secondary PPS Logic Racks
- 50. PPS Monitor Rack
- 51. Not Used 57 52. Not Used
- 53. Desk Top Radio
- 54. Reactor Control Rack
- 55. ilant Supervisory Rack
- 56. Sodium Flow Logic / Control Interface Rack
- 57. Plant Switching Logic Racks
- 58. Failed Fuel Readout Panel
- 59. Computer Typewriters
- 60. Computer Line Printer / Roller l 61. Cathode Ray Tube Display & Keyboard
- 62. Cathode Ray Tube Display & Keyboard
- 63. Cathode Ray Tube Display & Keyboard
- 64. Cathode Ray Tube Display & Keyboard
- 65. Cathode Ray Tube Display & Keyboard
- 66. PDH & L Remote Data Acquisition Terminal
- 67. Recirculation Gas Control Cabinet
- 68. Not Used 57 69. Not Used
- 70. Security Surveillance Station 41
- 71. Containment Instrumentation
() 7.9-10 Amend. 57 Nov. 1980
O TABLE 7.9-1 (Continued) Item
- 72. Plant Control System Switching Logic
- 73. Auxiliary Equipment Isolation Logic
- 74. Portable Radio Communications
- 75. Secondary Rod Control Cabinet
- 76. Heat Removal & Conditioning Logic Rack
- 77. PDH&DS Remote Data Acg. Term
- 78. Remote Annunciator Cabinets
- 79. Not Used
- 30. CRT Display and Keyboard
- 81. Not Used
- 82. Building Fire Protection Panel
- 83. Plant Security
- 84. Patch Panel 57 85. Containment Instrumentation O
7.9-10a Amend. 57 Nov. 1980
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11.6 0FFSITE RADIOLOGICAL MONITORING PROGRAM O ( The preoperational environmental monitoring progrm has the objec-tive of establishing a baseline of data on the distribution of background radio-activity in the environment near the plent site. With this background Informa-tion, it will then be possible to determine any statistically significant changes in the radioactivity levels. The preoperational environmental radiological moni-toring program will be initiated approximately two years prior to plant opera-l tion. The progr m outlined herein is based on current regulatory guidelines and 57: m nit ring philosophy. At such time that any monitoring progrm is implemented, it may be revised to reflect changes in regulatory requirements and monitoring philosophy. Evaluations after plant startup will be made on the basis of the baselines established in the preoperational progrm , considering geography and the time of the year where these f actors are applicable, and by comparisons to control stations where the concentrations of station ef fluents are expected to be negligible. In those cases where statistically significant increase in the radioactivity level is seen in a particular sampling vector but not in the con-trol station, meteorology and specific nuclide analysis will be used to identify the source of the increase. The planned sampling f requencies will ensure that significant changes in the environmental radioactivity can be detected. The vectors which would first Indicate Increases in radioactivity are sampled most frequently. Those wn!ch are less effected by transient changes but show Icng-term accumula-tions are sampled less frequently. However, specific sampling dates are not O'4 crucial and adverse weather conditions or equipment f ailure may en occasion prevent collection of specific samples. 4 The capability of the environmental monitoring program to detect design-level releases f rom plant ef fluents is uncertain because of the small quantitles which are expected to be released. The program will however provide the capability of detecting any significant buildup of radioactive material in the environment above and beyond that which is a! ready present. Those vectors which are most sensitive to reconcentration of specific isotopes are sampled. If any increase in radioactivity levels is detected in these vectors, the program will be evaluated and broadened if deemed necessary. From the data obtained from the radioanalytical and radiochemical analyses of the vectors sampled, dose estimates can be made for an Individual or the population living near the plant site. 11.6.1 Expected Backgrot;nd For e number of years measurements of background radiation have been made at various locations ';hroughout the Tennessee Valley region. Environ-mental monitoring programs have been conducted in the vicinity of Oak Ridge, Watts Bar, and Chattanooga, Tennessee, and Decatur, Alabama. Over periods of not less than two years, the measuromonts made in these areas have Indicated only very siight variations from location to location. The measurements obtained utilizing film badges or thermoluminescent dosimeters have revealed the following background radiation levels: Oak Ridge 78 mR/ year, Chattanooga 71 mR/ year, Watts d Bar 68 mR/ year, and Decatur 71 mR/ year, it is estimated that the expected Amend. 57 11.6-1 Nov. 1980 l
l background levels in the vicinity of the Clinch River Breeder Reacter Plant CRBRP> wili he between 60 and 90 mR/ year. Measurements f or the period 1977-1978 57j I (indicate the following yearly variations: Oak Ridge 73-83 mR, Chattanooga 68-74
; mR, Watts Bar 68-69 mR, and Deca fur 70-73 rrR.
Measurements will be made in the immediate vicinity of the CRBRP site and will provide baseline data necessary for comparison of background radia-tion levels prior to and after startup of the plant. 11.6.2 Critical Pathways to Msn Although the anounts of radioactivity added to the environment from plant operations are small, critical exposure pathways to man have been identified in order to estimate the maximum dose to the Individual and to estab-lish % sampling requirements for the environmental radioactivity monitoring
- sgram. The six principal pathways which can result in radiation exposure to man are as follows:
- a. External exposures and inhalation of gaseous releases.
- b. Drinking water from the Clinch River and f rom wel ls in the immediate vicinity of the plant.
- c. Swimming, boating, and fishing tr the Clinch River.
- d. Eating fish from the Clinch River.
- e. Consuming animal flesh and other animal products which may be af fected by plant operations.
- f. Eating foods gr'.en in areas adjacent to the plant site af fected by plant releases.
The er.vironmental monitoring program as outlined, provides sampiIng necessary to evaluate the dose received thrcugh the critical pathways in items a. through f. above. The following items indicate the samples collected in order to make the critical pathway-dose correlations:
- a. Data from readdings of the thermoluminescent dosimeters will be utilized to 57 estimate external exposures and data from of f site air monitorc will be used to estimate contributing internal exposures.
- b. Analysis of water samples collected will be used to estimate the dose that might be received from drinking water fran the Clinch River or from wells in the vicinity of the plant.
- c. Analysis of water samples will also be used to estimate the dose an indl-vidual might receive while swimming, boating, or fishing on the lake in the vicinity of the plant.
57l d. Analysis of samples of river water, sediment, and fish will be correlated to estimate the dose that might be received by an Individual who eats fish f rom the Clinch River. Amend. 57 flov. 1980 11.6-2
e&f. Analysis of samples of air, particul ate matter, soll, vegetation, food crops, and milk will be used to estimate tho dose to the surrounding popu-57! lation through the consumption of food or dairy products. The environmental monitoring program to be conducted throughout operation of the plant provides the necessary means of evaluating the dose to man through critical exposure pathways. Environmental concentrations of radioactivliy due to plant releases to unrestricted areas may be so Icw as to be unmeasurable with present techniques. Therefore, methods to calculate the potential exposure to man have been derived for both gaseous and liquid releases. 11.6.2.1 Dosos freg Gaseous Effluents The following doses to humans living in the vicinity of the CRBRP will be calculated for the releases of radioactive gases:
- a. External beta- and gamma-air doses from airborne radioactivity 57 b. Total body and skin doses f rom direct radiation due to ground contamination
- c. Internal doses f rom Inhalation
- d. Internal doses from ingestion The basic assumptions and calculational .-thods that will be used i in computing these doses are similar to that described in the appendix to Section l 57 Review of the data resulting f rcm the of fiste monitoring progran and reevaluations of the adequacy of the dose models will verify that the actual doses received by individuals and the population as a whole remain within the applicable Federal Regulations and as low as reasonably achievable. i l
11.6.2.2 Internal Doses frem Liould Effluents i The fo!!cwing doses will be calcula1ed for exposures to radio-. nuclides routinely released in liquid effluents: l
- a. Internal doses frca the Ingestion of water
- b. Internal doses from the consumpfion of fish 57lc. External snd internal doses from water sports
- d. External doses frca shcrolIne activities The basic assumptions and calculational methods that will be used in computing these doses are similer to that described in the appendix to Section 11.2.
The dose models that are employed will be reevaluated in light of l V the data resulting f rom the of fsite monitoring progran to ensure that all signi-Amend. 57 Ilov. 1980 11.6-3
57l ficant pathways are included in the calculations and to verify that the actual doses recolved by Individuals and the population as a whole remain within the 38l applicable Federal Regulations and as low as reasonably achievable. 11.6.3 Samoling Media. Locations. and Frecuencies The sampl ing media, the locations from which the samples are collected, and the f requency with which the samples are collected are presented in Table 11.6-1. Tentative sampling locations are shown in Figures 11.6-1 and 57 11.6-2. The final selection of sampling locations will be made approximately one year prior to implementation of the program. The media selected were chosen on two bases: First, those vectors which would readily Indicate significant increases in radioactivity levels, and secondly, those vectors which would Indi-cate long-term buildup of radioactivity. Consideration was also given to the pathways which would result in exposure to man, such as milk and f ood crops. Locations for sampling stations were chosen after considering meteorological fac-tors and population density around the site. Frequencies for sampling the various vectors were established so that seasonal variations in radioactivity levels might be determined. In addition, sanples are collected during the season in which the major growth occurs to ascertain radioactivity upteke by the vectors during their most susceptible period of growth. 11.6.4 Analvtical Sensitivity Samples wil l be collected routinely following established proco-dures so that unif ormity in sampling methods will always be assured. The samples 57 will be transported to a laboratory f acility for preparation and processing. All the radioanalytical and radiochemical analyses will be conducted in that labora-tory. The following types of equipment will be utilized in perf orming the analyses: Pulse height analyzers with solid and well Nal detectors and Ge(LI) detectors; low background beta counters; liquid scintillation counters; GM detec-57 l tors; and Internal proportional counters. Data will be coded and stored in com-puterized data base. The detection capabilities for environmental sample analyses will be presented in the PSAR. The nominal lower limit of detection (LLD) for the various analytical techniques will be based on the method discussed in HASL-300 57 l (ref. 1). The nominal LLD values are expected to approximate the values recom-mended in Regulatory Guide 4.2: However, the LLDs will vary depending on the 38 activities of the various components in the samples. 11.6.5 Data Analvsis and Presentation A quality control program has been established with the Tennessee Departmeni of Public Health Radiological Laboratory. Sanples of air particu-57 l ates, water, and milk will be collected and forwarded to this laboratory for analysis. The results will be exchangud f or comparison to aid the laboratories in evaluating their analytical systems and minimizing errors in data production. Data collection around the opei c+1ng plant will be compared to data f rom control stations and f rom the preoperational program to identify the earliest possible Indications t' the accumulation or buildup of rad!onuclides in the environment. During the lite of the plant, this accumulation should exist in 38 no more than trace anounts, with only minor impact on the environment. Amend. 57 flov. 1980 11.6-4
Reports describing the results of the environmental radiological monitoring activities will be submitted routinely as required by the Technical N 38 Specifications. The reports will follow the format used in reporting environ ' mental radiological data from TVA's other nuclear power facilities. 11.6.6 Program Statistical Sonsttivity As previously noted, because of the expected small quantitles of radioactive material which may be released to the environment from the CRBRP, it is uncertain as to what extent the results from the environmental monitoring pro-gram will be used to estimate the probable radiation exposure to man. Only if the radioactive waste releases from the plant cause statistically measurable incresess of radiation in the environment can dose correlations be made. Results f rom the analysis of of fluent samples, which contain higher concentrations of radionuclides than environmental samples, will be used in the TVA models similar to those given in the appendices to Sections 11.2 and 11.3 to estimate the possible exposure to man. Because of the conservative assumptions applied in these models, the estimated dose to the population should be higher than that actually received. However, TVA, even using the conservative assumptions, will control the releases of radioactive materials to the environ-57 ment such that the releases ranain within the applicable Federal Regulations and as low as reasonably achievable. 57 The statistical sensitivity of the monitoring during accident con-ditions will not be dif ferent f rom those during normal plant operation and is p expected to detect concentrations well below 10CFR Part 20 limits. L) i O ' b Amend. 57 Nov. 1980 11.6-5
l REFERENCES 57
- 1. HASL-300,, HASL Procedure Manual, J. H. Harley, Ed., Rev.
O\ August 1974. I 1 l l I l l i l O l l l l Amend. 57 Nov. 1980 11.6-6
_ _ _ _ _ _ _ _ _ _ ___m __ _ _ - - . O O O TABLE 11.6-1 ENVIRONMEf4TAL RADIOLOGICAL SURVEILLANCE PROGRAM Collection Criteria and Sampling Locations Frequency Analysis / Counting I. Atmospheric A. Air b
- 1. Particulate filter paper at 12-15 locations Weekly Gross beta, gross alpha, (gama scan monthly, Pu and U quarterly),
(895r, f0Sr)c b 131
- 2. Radioiodine Charcoal filter at 12-15 locations Weekly 3 3
- 3. Moisture Sampling at 3-8 locations Weekly H h B. Fallout Gummed acetate at 12-15 locations Monthly Gross beta, gross alpha T
C. Rainwater Rainwater collection trays at 12-15 locations Monthly Gross beta, gama scan 895r, 90Str 3;, II. Reservoir A. Water
- 1. Municipal All public water supply intakes within 10 Monthly # Gross beta, gross alpha, (Publicsupplies) miles upstream and downstream of the plant gama scan, JH monthly, Pu quarterly
- 2. River Continuous samples from 4-8 locations Analyzed Monthly Gross beta,3 gross alpha, gaira scan H monthly, E@ U9S r, 90Sr. Pu, U quar-5g terly a.
G* a. All public water supplies within 10 miles downstream of the plant will be collected automatically and analyzed g$ monthly.
- b. Continuous sampling. 137 Cs
- c. Radiostrontium composite analyses if is found in the gama scan.
57
TABLE 11.6-1 (Continued) Collection Criteria and Sarpling Locations Frequency Analysis / Counting B. Aquatic Biota
- 1. Fish Two locations Semiannually gross Gross beta,EYSr,' 9gSr, gart a scan lpha, Pu
- 2. Shellfish Four to six locations Semiannually (ifavailable) Grossbeta,grossafpha, gama scan (69 S r, 9 Sr Pu shells only)
C. Sediment Four to six locations Semiannually Gross gammabeta, scan, cross 695r, gpha, Sr Pu III. Terrestrial A. Soil Atmospheric monitoring locations Annually Gross beta, gross alpha, gant.a scan, Pu, U - B. Vegetation F 1. Pasturage Selected dairy farms within 10-mile Quarterly T cc grass radius of plant Gross beta, gross gpha, gaura scan, b95r, bSr. Pu
- 2. Grass Collected at atmospheric monitoring stations Quarterly Same as pastarage grass
- 3. Food crops Within 10-mile radius of plant Annually Gross beta, goss gpha gaar.a scan, Sr. Sr. Pu C. Milk Selected dairy farms within 10-mile radius Monthly d
Ganna scan, 89Sr,90Sr,1317 of plant D. Well water Selected farms within 5 niles of plant and Monthly Gross beta, gross alpha, 1-5 wells on site gaana scan monthly, Pu quarterly E. Direct radiation TLDs on site and at atmospheric conitors Qua rterly Dose determination
- d. Af ter the plant begins operation milk samples will be taken at least biweekly for 1311 analysis wtan cows are on 57 pasture.
5N $e e G G
" O O LOCAL AIR MONITOR (L AM) t<oeris Dam e PERIMETER AIR MONITOR (PAM) 9 REMOTE AIR MONITOR (RAM)
) ORGOP -0AK P.10GE GASEOUS OlFFUSION PLANT U ORIP -0AK RIDGE INDUSTRIAL PARK ORNL -0AK RIDGE NATIONAL LABORATORY Chnton 04k Ridge y O
/
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- O 10.0 f 'Ok White Oak p# 7 15.0 '
Rock wood 15.9* CRBHP 8 946 O l18Melton Hill Kingsto @ p,, 24.0 E %,, (v,'/) 6 5 Mile, Leno r Caty 10 Miles Louden 15 M tes NOTE: THE FOLLOWING SAMPLES ARE COLLECTED AT EACH MONITORING SITE: AIR PARTICL . ATE R AINWAT E R RA01010 Dir- S0lt
. E AVY PARTICLE YEGETATION FALLOUT Figme i 1.6-1. Almospherie and Terrestrial \1onitoring Netw ork Clincli Iliser lireeder Itenelor 6713-1
( Amend. 57 Nov. 1980 ( 11.6-9
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12.3 HEALTH PHYSICS PROGRAM v 12.3.1 Program Objectives The health physics staff is a unit of the Radiological Hygiene Branch and is responsible for the health physics activities at the plant. It applies radiation standards and procedures; reviews proposed methods of plant operation; participates in development of plant documents; and assists in the plant training program, providing specialized training in radiation protection. During preoperational tests and af ter plant startup it provides health physics coverage for all operations including maintenance, fuel handling, waste disposal, and decontamination. It is responsible for personnel and inplant radiation monitoring, and maintains continuing records of personnel exposures, plant radiation, and conthmi-nation levels. Through implementation of the described program, plant personnel exposures will be maintained as low as is reasonably achiev-able (ALARA). The health physics staff is under the administrative super-vision of the Chief, Radiological Hygiene Branch, in t'e TVA Division 57l of Occupational Health and Safety. The plant health physicist is responsible for direction of an adequate program of health surveillance for all plant operations involving potential radiation hazardo. "e keeps the plant superintendent informed at all times of radiation hazards and conditions related to potential \' exposure, contamination of plant and equipment, or contamination of site and environs. His duties include training 3nd supervising health physics technicians; planning and scheduling monitoring and surveillance ser-vices; scheduling technicians to ensure around-the-clock shift coverage as reouired; maintaining current, data files on radiation and contami-nation levels, pei sonnel exposures, and work restrictions; and ensuring that operations are carried out within the provisions of developed radio-logical hygiene standards and procedures. He provides assistance and advice to the plant superintendent during radiological emergencies. In addition, off-site staff from the Radiological Hygiene Branch is responsible for conducting a comprehensive environmental moni-toring program prior to, during, and after plant startup. The Division 57 of Occupational Health and Safety also advises on potentially harmful factors in the working environment other than radiation, all in relation to identified and approved standards as given in the TVA Hazard Control Manual. Health physics personnel that are assigned to the plant staff, will meet the provisions of Regulatory Gus T 8 and 8.10 or Regulatory Guide 1.8 and Section 4 of ANSI N18.1, lit eve- is 'pplicable at the time of appointment. As TVA has a F'% y .ali fimi aealth physics staff, I there should be no problem in meet.m . ;avisions of the applicable ! 49 guides. ' h G 12.3-1 Amend. 57 Nov. 1930
TVA has established a formal program to ensure that occupa-tional radiation exposures to employees are kept as low as is reasonably achievable (ALARA) unich will be applied to CRBRP. The program consists of: (1) full manaaement commitment to the overall objectives of ALARA; (2) issaance of specific administrative documents and procedures to the TVA design and operating groups that emphasize the importance of ALARA throughout the design, testing, startup, operation, and maintenance phases of TVA nuclear plants; (3) continual appraisal of radiological conditions in the operating nuclear plants by an on-site health physics staf f, and (4) a 4-member corporate ALARA committee consisting of repre-sentatives from ae TVA design, operations, and radiation protection groups, whose purpose is to review and appraise the effectiveness of the ALARA program on a plant-by-plant basis including CRBRP. This committee consists of key management and technical staff who have extensive backgrounds in inplant radiation control, including such areas as plant layout, shielding, personnel access control, ventilation, '^ste management, area and personnel monitoring, plant operations, and pl<nt maintenance. The committee periodically evaluates TVA's overall ALARA program by assessing trends in occupational exposures or other radiation control problems, reviewing plant operating reports and radiation exposure profiles, and conducting onsite audits of each plant's ALARA efforts. Specific responsibilities of the ALARA committee include the following: (a) Determines that an effective ALARA program is established at each TVA nuclear power plant that appropriately integrates TVA manage-ment philosophy and NRC regulatory requirements; (b) Determines that the Al ARA program is implemented from initial planning through decommissioning of the plant; (c) Reviews plant design features, operating procedures, and maintenance practices and audits the onsite radiation control program at least annally to assure that the objectives of the ALARA program are attained; (d) Determines that infomation and data pertaining to radiation ex-posure of personnel from other operating LWR power plants are reflected in the design and operation of new TVA plants; (e) Determines that experience gained during the operation of nuclear power plants relative to inplant radiation control is factored into revisions of operating procedures, where necessary, to assure that the procedures indeed do meet the objectives of the ALARA program; (f) Determines that all maintenance activities are planned and accom-plished in accordance with the objectives of the ALARA program; and ! WJ ' i 12.3-2 Amend. 49 Apr. 1979
(g) Determines trends in the exposure of station personnel in order to permit actions to be taken to correct adverse trends. Reports of the findings of the ALARA committee are promptly conveyed to top-level management staff along with appropriate recommen-dations for improvements in the design of new plants or corrections in operating plants. 12.3.2 Facilities and Equipment The health physics facilities necessary to monitor and control the routine radiological condition of the plant is shown on the Plant Service Building General Arrangement drawing in Section 1.2. The focal point of control is the assembly area. All personnel entering or leaving the controlled (restricted) area must pass through this area. All other 52 doors are for emergency use only. The general entry control requirements will be coordinated with industrial security. The control point is equipped with the following: a) A health physics work station for routine counting and assignment of required equipment to workers entering the restricted area. b) Mask-Protective clothes issue and storage area. c) Male / female lockers, toilet facilities shower station d) Mask cleaning station e) Combined laboratory f) Chemical storage room g) Counting room h) Hot instrument shop The main personnel monitoring station is at the Reactor Service Buildin; entrance from the Plant Service Building. This entrance / exit door will be equipped with all necessary monitoring equipment. Contamination control, within the Plant Service Building control area, such as entering and exiting from counting room, hot instrument shop, mask cleaning area, laboratory area, radwaste area, will be accomplished by use of local survey equipment located at the accesses to these areas. Contaminated laundry will be bagged, surveyed, and shipped 49 off-site for laundering. O V Amend. 52 12.3-3 Oct. 1979 ' I
For radioloqical purposes, all other areas, i.e. , outside the RCB, RSB, Intermediate Bay and the cells in the PSB are uncontrolled (unrestricted). Portable and laboratory equipment located in the health physics work station will allow the health physics personnel to measure dose rates and contan: . tion levels throughout the plant in all routine and emergency si tuations. The portable health physics survey instrumen-tation is sted in Table 12.3-1 with the operational characteristics for each -trument. The fixed health physics laboratory counting sys-tems are cescribed in Table 12.3-2. All potentially contaminated liquid drains in this area will be routed to the radwaste system and all potentially contaminated gaseous exhausts will be HEPA filtered. Portable survey instrumentation will be checked and calibrated 57l reatinely with standard radioactive sources by the TVA Branch Laboratory in Muscle Shoals, Alabama. Accurate records on the performance of each instrument during each calibration will be maintained at this laboratory. Calibration and maintenance procedures specific for each instrument are written and routinely used. Each laboratory counting system is checked at regular intervals with standard radioactive sources for proper counting efficiencies, background count rates, and high voltage settings by health physics personnel at the plant. TVA will provide protective clothing for use in radiation areas. Clothing required for a particular instance shall be prescribed by the Health Physics Staff based upon the actual or potential radio-logical conditions. Protective clothing available for use are:
- a. Coveralls
- b. Lab coats
- c. Gloves - plastic and/or latex in light and heavy weights
- d. Gloves - cotton in heavy weights and light weights
- e. Head covers - skull caps and hoods
- f. Foot covers - shoe rubbers and plastic booties 9 Plastic suits Tape will be provided so that openings in clothing and between 49 pieces can be sealed.
12.3-4 Amend. 57 Nov. 1980
O TABLE 13.3-4 i PROJECTED MAXIMUM RESIDENT AND TRANSIENT POPULATION
- Il cVACUATION l 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 6220 6295 O
- Transient population is based on current available infonnation 50 See Figures 13.3-5 and 13.3-6 Amend. 50 13.3-15 June 1979
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1 8./ m . g 15 25 I a _f i 1 3 4 5 i 4 5 11 7 25 3 4 3 6 7 s t C 0.1 1.0 10.0 %.t Distance from $,0f Containment Building, Miles Amend. 57 oe o q. . Nov. 1980 13.3-19 m od o JJ.. _a
l l 1 40 15.A.1 INTRODUCTION In accordance with Title 10 Code of Federal Regulations Part 50 (10CFR50), the CRBRP Project has submitted an Environmental Report (ER) and a Preliminary Safety Analysis Report (PSAR) to support an application 40l for a license to construct the CRBRP. These reports include an evaluation of a spectrum of postulated accidents. For each accident an analysis of the potential consequences to the health and safety of the public is presented. Consistent with the intent of Regulatory Guide 4.2, " Pre-paration of Environmental Reports for Nuclear Power Plants", the accident evaluations presented in the ER are based on realistic accident analyses and analytical assumptions. The evaluations presented in the PSAR are based on conservative accident analyses and analytical assumptions. The spectrum of accidents considered in Chapter 15 of the PSA'1 and Chapter 7 of the ER encompass Class 1 through Class 8 events. Tnit spectrum con-stitutes the accidents included in the design base for tae plant. Class 9 events are of such low p,obability that they can be excluded from the design bases. /~s U 29 In accordance with Title 10 Code of Federal Regulations Part 100 (10CFR100), a major fission product release from the core has been hypo-thesized for the purpose of detennining the suitability of the selected site for the construction and operation of the CRBRP. In compliance with 10CFR100, the potential hazards resulting from this hypothesized release are not exceeded by those from any design basis accident analyzed in Chapter 15 of the PSAR. The radiological source term associated with this hypo-thetical release is specified in terms of percentages of fission products and fuel material released from the core to the Reactor Containment Building. The source term used for site suitability assessment is as follows: 100% Noble Gas Inventory 50% Halo,qen Inventory (25% Airborne) 1% Solid Fission Product Inventory 1% Plutonium The applicant has utilized this source term in compliance with specific direction from the Nuclear Regulatory Commission (Ref.1). How-ever, while accepting this source term and committing to design features to assure acceptable consequences as a result of it, the applicant considers 40 this source term to be overly conservative. /\ I J. Amend. 40 15.A-1 July 1977
O The source tenn specified by NRC not only envelopes all design basis accidents considered in Chapter 15, but further envelopes a wide range of conservatively hypothesized core-related events. Evidence, both analytical and experimental, supports the Applicant's position that compliance with the requirements of 10CFR100 could be demonstrated with a less stringent source term. 40 The potential radiological consequences of the above source term are conservatively calculated and compared to the guideline values of 10CFR100, thus providing the basis for conducting an assessment of the l29 1 site suitability.
- 15. A.2 SITE SUITABILITY SOURCE TERM 15.A.2.1 Source Tenn "
The source term is identified in terms of percentages of fission products and fuel material ra1 cased from the core to the Reactor Contain-ment Building. The source term is itemized in Table 15. A-1. The indicated percentages of these materials are assumed instantly released to and uniformly distributed in the RCB. For the halogens, 50% of the halogens released to the RCB are assumed to immediately plateout on surfaces (con-sistent with LWR practice), thus being removed from the airborne source term available for leakage from the RCB, with the not result that 25% of the initial halogen inventory is assumed airborne in the RCB. The initial core fission product inventcries are based on end-of-cycle equilibrium core conditions for power opeistion et 975 megawatts-57 541 thermal. The specific isotopes included in each fission product category, as identified in Table 15. A-1, are as follows: Noble Gases: Xe, Kr Halogens: Br, I 40 Solids: All remaining fission products 15.A-2 Amend. 57 Nov. 1980 9
L The mass- associated with the non-gaseous portion of the source term, initially airborne in the RCB, is shown in Table 15.A-4. The resultant initial airborne concentration is also provided. 57 l The quantity of fuel (62.4 kg) included in the source term aerosol analysis was selected to represent 1% of the total (core plus blanket) 51 plutonium-oxide mass plus 1% of the core uranium-oxide mass. The mass of the uranium-oxide in the blanket was not included in the aerosol analysis. This approach is conservative since including the uranium blanket mass would result in a much higher initial airborne concentra-tion and subsequently more rapid aerosol depletion. Even though the uranium blanket mass has been excit.ded from the aerosol analysis, it has been conservatively assumed that the radioactive inventory of the blanket is included in the source tom. Table 15. A-5 presents the important input parameters to the HAA-3 code, used to compute the concentration-time behavior of the source term aerosol and resultant aerosol depletion factors. The time-depen-dent depletion factors computed by HAA-3 for the source term aerosol and used in the COMRADEX radiological analysis are itemized in Table 15.A-6. 15.A.2.3 Containment Modeling A complete description of the reactor containment / confinement fi N_/ system and the engineered .;afeguards associated with it is presented in Chapter 6 of the PSAR. For the radiological analysis, it is conservatively assumed th:st all leakage (except bypass) from the RCB to the annulus is directly to the intake of the filter system. This assumptior neglects any credit for delay time in the annulus. The recirculation flow was assumed to mix in 50% of the annulus volume. Only one-half of the annulus volume 40 is used to be consistent with the 50% mixing assumption specified in Standard Review Plan Section 6.5.3. I i l Amend. 57 15.A-5 Nov. 1980 f
, _ _ - __ yr. , , - - - - - - , - _ - ,
- y _
Leakage of airborne radioactivity from the RCB was assumed to occur at the containment design leak rate, 0.1% Vol/ Day for the duration of the evaluation. The RCB is designed to limit leakage to 0.1% Vol/ Day at a containment overpressure of 10 psig. The use of the containment design leak rate (0.1% Vol/ Day) for the duration of the site suitability source term evaluation is conservative, since assuming a constant 10 psig containment overpressure for the duration of the site suitability source term evaluation is conservative. A portion of the leakage from the RCB may bypass the confinement annulus. Chapter 6 of the PSAR identifies the individual containment penetrations contributing to bypass leakage; the majority of the bypass leakage is associated with the containment airlocks. The containment / confinement system is being designed to achieve a bypass leakage value of less than 1% of the RCB design leak rate, i.e. ,1% x 0.1% Vol/ Day = 0.001% Vol/ Day. Sixty percent of this bypass leakage escapes directly to the outside atmosphere and the remainin9 forty percent escapes to the Reactor Service Building (RSB). The treatment of leakage to the RSB depends upon the status of the railroad door in the RSB. When the railroad door is closed, the RSB atmosphere is maintained at a negative pressure with respect to the outside atmosphere. When the railroad door is open, maintenance of a negative pressure in the RSB is not assured. If the RSB railroad door is open, both doors of the equipment hatch airlock are secured and the airlock atmosphere is vented to the containment / confinement Annulus Filtration System. In this mode, essentially all (96.4%) the bypass leakage fror, the RCB to the RSB (40% of total bypass) is vented from the equipment hatch airlock directly to the Annulus Filtration System, t.here it is subject to filtration and recirculation prior to release to the environment. The re'nainder of leakage into the RSB (3.6%) escapes directly to the atmosphere. 4 hen the railroad door is closed, the airlock vent to the Annulus Filtration System is closed and the airlock atmosphere is iso-lated from the containment / confinement annulus. In this mode, all bypass leakage frora the RCB to the RSB (40% of total bypass) escapes directly to the RSB whece it is subject to recirculation and filtration prior to release to the envir1nment. Airlock operation with the railroad door open (i.e. , with the airlock atmosphere vented to the annulus) results in larger potential off-site exposures for the site suitability source term analysis than operation with the railroad door closed and the radiological consequences are there-fore presented when the railroad door is assumed open. Confirmation that this does result in more limiting exposures is given below. When '.he railroad door is closed and all byoass leakage to the airlock escapes to the RSB, this leakage is filtered prior to ultimate release to the environment. Considering the efficiencies (99% particulate and 95% iodine) of the RSB filters and the recirculation flow pattern (1700 cfm exhausted per 14300 cfm recirculated), the net filtration efficiency of 40 the RSB system is greater than 99% for both particulates and halogens. Conse-quently, non-gaseous releases (which are controlling with respect to off-site 15.A-6 Amend. 57 Nov. 1980
l
,f O O V kJ TABLE 15.A-2 HEAVY METAL
- MASS (KG) INVENTORY IN THE CRBRP (Entc)
Inner Radial Lower Axial Unper Axial Fuel Blanket (a) Blanket (a) Blanket Blanket End-of-Fourth-Cycle Pu-239 1216. 206.8 285.6 34.9 21.2 Pu-240 273.5 8.0 11.3 0.9 0.3 Pu-241 32.7 --- --- --- --- _. Pu-242 5.2 --- --- --- --- ( U-235 5.4 11.6 21.3 3.8 4.0
$ U-238 3421. 7381. 12936. 2149. 2165.
Fission Products 414.2 55.2 55.7 4.4 2.4 Total Heavy Metal 5368.0 7662.6 13309.9 2193.0 2192.9
- Heavy metal excludes oxygen.
51 5N 58 P
l TABLE 15.A-3 CRBRP TRANSURANIC INV^4 TORY (E0EC) Isotope Hal f-Li fe Mass (gms) Curies O 0 Np237 2.14 x 106 Y 3.38 x 10 3 2.38 x 10 Mp238 2.1 D 1.50 x 10 0 3.93 x 10 5 Mp239 2.35 D 4.08 x 10 9.48 x 10 Am241 458 Y 7.33 x 10 3 2.51 x 10 Am242* 152 Y 1.62 x 102
- 1.57 x 10 3, 0 6 Am242 16 H 3.81 x 10 3.08 x 10 Am243 7650 Y 2.17 x 10 4.18 x 10 I 4
Am244 10 H 8.69 x 10~ _.58 x 10 Cm242 163 0 6.23 x 10 2.06 x 10 0 Cm243 32 Y 2.27 x 10 I 1.04 x 10 3 Cm24', 18.1 Y 8.70 x 10 0 7.05 x 10 2
-l Cm245 9320 Y 1.45 x 10 2.57 x 10~
-4 Cm246 5480 Y 2.19 x 10~ 6.77 x 10 1.67 x 107 Y -b Cm247 1.92 x 10 1.69 x 10 ~9 1.78 x 10 -7 7.29 x 10 -10 5
Cm248 4.7 x 10 Y 57 40 Cf252 2.55 Y 4.15 x 10 2.22 x 10 -13 51
- Estimated Value 15.A-12 Amend. 57 Nov. 1980
i i l0 { TABLE 15.A-4 i MASS OF SOURCE TERMS INITIALLY AIRBORNE IN rcd l l Isotope Mass l Class (kg) j Noble Gases
- 74.34 Halogens ** 1.59 i
Solid Fission Product 5.55 Fuel 62.45 2 Total j 51 Non-Gaseous 69.59 i I initial RCB Concentration (ugm/cc) ! 57 0.68 1 i
- Mass of Noble Gases excluded from aerosol analysis.
**25% of E0EC Inventory.
40 i 1 } { l e i l i O Amend. 57
- 15. A-13 Nov. 1980
TABLE 15.A-5 IIAA-3 INPUT PARAMETERS USED FOR SOURCE TERM ANA! Y"TS_ Pa rameter Initial Concentration, Particles /cc 1.34 x 10 Count Mean Particle Radius, um 0.1 Geometric Mean Deviation, um 2.0 Aerosol Material Density, gm/cc 10.55 Stokes Correction Factor, a 0.1 Gravitational Collision Efficiency, c 1.0 RCB Volume, cm 3 1.02 x 10 ll RCB Leak Rate, fraction /sec 1.16 x 10 -8 57 Plating Constant, A 4 x 10 -5 40 15.A-14 Amend. 57 Nov. 1980 l l
l l TABLE 15.A-6 l AEROSOL DEPLETION FACTORS USED FOR SOURCE TERM Depletion Factor Time (Sec) (Fraction /Sec) 6.50-3 9.55-5 4.60-2 9.49-5 I 2.63-1 9.17-5 1
; 1.25+0 8.21-5 3.84+0 6.92-5 9.22+0 5.79-5 1.95+1 4.87-5 3.94+1 4.10-5 7.73+1 3.45-5 1.47+2 2.91-5 5.20+2 2.04-5 j 1.90+3 1.37-5 J
6.42+3 9.43-6 i 1.98+4 7.06-6 1.09+5 6.33-6 4.21+5 4.56-6 i j 1.09+6 2.24-6 j . 57 2.60+6 1.40-6 i 40 3 i I I Amend. 57
- 15. A-15 Nov. 1980 i
i _ _ _ _._ _ ___.___._._ .__ _. ._ _._ _. _. - . _ - _ _ _ _ . . . . . _ _ _ _ _ . ._._.________...___._.-___.---.._I
TABLE 15.A-7 CONTAINMENT / CONFINEMENT PARAMETERS USED FOR SOURCE TERM ANALYSIS RCB Leakage to Annulus 0.1% Volume / Day (Direct to Annulus Filter Intake) Annulus Flow Rates Filtered Exhaust 3000 CFM Filtered Recirculation 3500 CFM per 1000 CFM Exhausted Time Delay from Source Term Release No Delay to Initiation of Annulus Filtration Time Delay from Source Term Release <10 Seconds to Initiat#on of Annulus Recirculation Total Bypass Leakage 0.001% Volume / Day (1% of RCB Leakage) Bypass Leakage Direct to Environment 0.0006% Volume / Day (60% of Total Bypass) Bypass Leakage to the RSB (40% of Total Bypass) 0.0004% Volume / Day O Sources of Bypass Leakage 96.4% Personnel and Airlock to the RSB Equipment 3.6% All other sources Gamma Shielding 1.5" Steel (RCB) Plus 4' Concrete Filter Ef ficiencies Iodine 95% Particulate 99% 40 Noble Gases 0 Am"nd. 40 O 15.A-1 6 July 1977
l l n V TABLE 15. A-8 ! METEOROLOGICAL PARAMETERS USED FOR SITE SUITABILITY ASSESSMENT 3 Exclusion Boundary (0.42 Miles) X/Q (sec/m ) 0-2 Hours 3.12 x 10-3, Low Population Zone (2.5 Miles) 0-2 Hours 8.55 x 10-4* 2-8 Hours 2.85 x 10-4 8-24 Hours 2.60 x 10-5 ,q 1-4 Days 1.30 x 10-5 (s' 4-30 Days 8.70 x 10'6 40
- 0-2 Hour X/Q's based on single-hour 95% X/Q Values.
O %)
- 15. A-17 Amend. 40 July 1977
TABLE 15.A-9 0FF-SITE EXPOSURE
SUMMARY
(Power Level = 975 Megawatts-Thermal) Dose (Rem) 2-Hour 30-Day Site Boundary Low Population Zone Organ 10CFR100 (0.42 Miles) (2.5 Miles) Bone 150* 7,2 4.1 Lung 75* 1.6 .9 Thyroid 300 23.1 12.6 57 51 Whole Body ** 25 3.7 1.7
- Equivalent to 10CFR100 guideline values; see Reference 4.
** Includes inhalation, external gama cloud, and direct gamma shine exposures.
40
'^~
Amend. 57 flov. 1980 l l
i 1 0 i i i i i I 1 i i. i I l Pages 16.6-11 and 16.6-12 HAVE BEEN DELETED 57 1 I O i i l 16.6-11 (Next Page is 16.6-13) Amend. 57 Nov. 1980
i t I f j Amendmend 57 t i j LIST OF RESPONSES TO NRC QUESTIONS l l l t 1 i l There are no new NRC Questions in Amendment 57. 1 1 1 i l 4 O : i b 1 [ t i 1 ! ! f
^
l l i I e i l l , l l I j i i ; ' i I i ! l 4 ! Q-i i
_uestion Q 7 Provide an overview of the methods used to evaluate the structural integrity of the fuel assembly including a description of all analytical methods used (e.g., PECT2) and all applicable data. The report should be in the form of a summary addressing all calculational limits (e.g., stress and deflection).
Response
The information requested concerning the CRBRP Fuel Assembly was provided under separate cover in the following topical report:
"CRBRP Fuel Assembly Structural Analysis in Support of the Final Design Review", CRBRP-ARD-0204 Additional information concerning the CRBRP Fuel Rod will be provided in a topical report at a later date. A Table of Contents for this report 57 was provided to the NRC in December, 1976.
O O Amend. 57 Q7-1 Nov. 1980 _ _ - -}}