ML20040F047
| ML20040F047 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 01/31/1982 |
| From: | ENERGY, DEPT. OF |
| To: | |
| Shared Package | |
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| References | |
| NUDOCS 8202080255 | |
| Download: ML20040F047 (800) | |
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{{#Wiki_filter:._ _ .. _ _ - _ _ - . l PAGE REPLACEMENT GUIDE FOR AMENDMENT 64 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT 1 (DOCKET NO. 50-537) l Transmitted herein is Amendment 64 to Clinch River -Breeder O Reactor Plant Preliminary Safety Analysis Report,-Docket 50-537. Amendment 64 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 64 pages and instructions for their incorporation into the Preliminary Dfety Analysis Report. I i O . 8202000255 820129 PDR ADOCK 05000537 B PDR
AMEN 0 MENT 64 PAGE REPLACEMENT GUIDE REMOVE THESE PAGES INSERT THESE PAGES Chapter 1 1.1-5, 6 1.1-5, 6 1.1-11 thru 22 1.1-11 thru 27 1.2-1 thru 6 1.2-1 thru 6 1.2-32 thru 104 1.2-32 thru 104 1.3-11 thru 14 1.3-11 thru 14 Chapter 3 3.8-22, 22a 3.8-22, 22a 3.8-B11 3.8-Bii 3.8-B.1 thru 6 3.8-B.1 thru 4, 4a, 5, 6 3.8-B.11, 12 3.8-B.11, 12 3.8-B.14a, 15 thru 18 3.8-B.14a, 15, 15a, 16, 17, 18 3.8-B.29 3.8-B.29 thru 32
- 3.8-C.i, ii,1 thru 18 (Insert following 3.8-B.32)
D 3A.1-1, 2 3A.1-1, 2 (d 3A.8-1, la 3A.8-1,-la Chapter 4 4.1-3 thru 6 4.1-3 thru 6 4.2-42, 43 4.2-42, 43 4.2-181a 4.2-181a 4.2-216 thru 219 4.2-216 thru 219 4.2-330, 330a 4.2-330, 330a 4.2-342, 343 4.2-342, 343 4.2-474, 475 4.2-474, 475 4.2-500, 501 4.2-500, 501 4.2-504, 505 4.2-504, 505 4.2-524, 525 4.2-524, 525 4.2-548, 549 4.2-548, 549 l 4.2-576, 577 4.2-576, 577 4.3-1 thru 6 4.3-1 thru 6 4.3-17 thru 22 4.3-17 thru 22 4.3-41, 42 4.3-41, 42 4.3-92, 93 4.3-92, 93 i 4.3-98 thru 105 4.3-98 thru 105 4.3-108, 109 4.3-108, 109 4.3-144 thru 147 4.3-144 thru 147 > 4.3-150 thru 153 4.3-150 thru 153 , 4.3-180, 181 4.3-180,.181 4.3-198, 199 l l]n 4.3-198, 199 l A i __.____A
REMOVE THESE PAGES INSERT THESE PAGES Chapter 4 (Cont'd.) 4.4-87 thru 90 4.4-87 thru 90 4.4-127 thru 130 4.4-127 thru 130 4.4-135, 136 4.4-135, 136 4.4-141 thru 144 4.4-141 thru 144 4.4-151, 152 4.4-151, 152 4.4-157 thru 162 4.4-157 thru 162 4.4-171 thru 178 4.4-171 thru 178 4.4-191, 192 4.4-191, 192 Chapter 5 l 5.1-20, 21 5.1-20, 21, 22 5.3-75b, 75c 5.3-75b, 75c Chapter 6 6.1-1, 2 6.1-1, 2 , Remove Entire Section 6.2 Insert Entire Section 6.2 (Last page is 6.2-47) 6.4-1 (following 6.3-18) 6.5-1 (following 6.4-1) Chapter 7 2 7.1-9, 10 7.1-9, 10 7.4-3 thru 6 7.4-3 thru 6 7.5-24 thru 27 7.5-24 thru 27 Chapter 9 9-iii, iv 9-111, iv 9.1-29, 30 9.1-29, 30 Remove Entire Section 9.2 Insert Entire Section 9.2 (Last page 9.2-5) 9.3-2 thru 5 9.3-2 thru 5 9.5-1 thru 6, 6a 9.5-1 thru 6, 6a 9.5-7, 8, 8a, 9, 9a, 10, 11, 12 9.5-7 thru 12 9.5-15, 15a, 16, 17, 18, 18a, 9.5-15, 15a, 16, 17, 18, 18a, 18b, 18c 18b, 18c 9.5-19 thru 41, 41a, 42, 43 9.5-19 thru 41, 41a, 42, 43 9.8-1 thru 6 9.8-1 thru 6 9.13-13 thru 17, 17a thru 17d 9.13-13 thru 17,.17a thru 17d 9.13-45c thru 45h 9.13-45c thru 45h 9.13-47, 48 9.13-47, 48 Chapter 11 i Remove Entire Section 11.3 Insert Entire Section 11.3 1_ m) (Last page 11.3-49) (Last page 11.3-48) B
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! REMOVE THESE PAGES INSERT THESE PAGES !
!O Chapter 12 l
.12.1-11 thru 17, 17a 12.1-11 thru 17,17a ;
12.1-20, 20a 12.1-20, 20a 12.1-32, 33 12.1-32, 33
- f 12.1-58, 59 12.1-58, 59 l 4
l Chapter 13 13.2-1 thru 4 13.2-1 thru 4 Chapter 15 ,. l Remove Entire Section 15.6 Insert Entire Section 15.6 - i (Last page is 15.6-45) 15.7-11 thru 15 15.7-11 thru 15 15.7-17d thru 17h 15.7-17d tiru 17j Appendix A t A-1 thru iv A-1, ii, iia, iii, iv _
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. A-139, 140 . A-138a , 139, 140
, A-239 thru 242 A-239, 239a,.240, 241, 242 -
A-247, 248 A-247, 247a, 248 j 4 , i i I I L ? 1 , i i l l s !O C l ; I u_...____.____
< t ( AMENDMENT 64 QUESTION / RESPONSE SUPPLEMENT l This Question / Response Supplement contains an Amendment 64 tab i sheet to be inserted following Qi page, Amendment 63. December 1981. Page Qi Amendment 64, is to follow the Amendment 64 tab. Replacement pages for the Question / Response Supplement are listed below: 1 l REPLACEMENT PAGES 4 REMOVE THESE PAGES INSERT THESE PAGES () Q040.8-1 Q130.82-1 Q040.8-1 Q130.82-1 Q130.92-1 Q130.92-1 I i l I
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However, to provide extra margin, because of the first-of-a-kind nature of the plant, certain hypothetical core disruptive accidents have been analyzed, As a unique feature of the design of this plant, additional Design Margins are provided such that even the consequences of postulated HCDAs will also be adequately accommodated. These include both structural and thermal margins. Structural Margin Beyond the Design Basis (SEDB) assure that extra margins exist to accommodate structural loadings on the reactor vessel system and the PHTS components f rom postulated HCDAs. Thermal Margins Beyond the Design Basis (TEDB) assure that radiological consequences of HCDAs will be mitigated to acceptable levels. Preliminary details are provided in Reference 10 of PSAR Section 1.6. PIoing InteprItv Considerations The Integrity of the sodium piping in containment is of special concern because of the f unction of the sodium heat transport systems in maintaining the reactor core in a safe condition. A multifacetted program has been undertaken by the Project to assure and document the high integrity of the in-containment sodium piping. The details of the piping integrity program are discussed in Ref erence 2 of PSAR Section 1.6. The program will show that with the appropriate matarials surveillance, in-service inspection and leak detection provisions, double-ended pipe ruptures of the sodium piping in containment can be excluded from the CRBRP design basis. 1.1.3 Aoolleability of Regulate,v Guldes l The NRC Regulatory Guides provide Regulatory guidance as to how the requirements of NRC regulations may be satisfied. These requirements are set forth in Appendix A to 10CFR Part 50 for design of nuclear power plants and in various parts of 10 CFR for design, construction, operation, and quality assurance. Many of the detailed requirements address directly the light-water-cooled nuclear power plants. Consequently, a number of the requirements of the existing Regulatory Guides may or may not apply to the CRBRP, due to the dif ferences in designs between the LWBR plants and the LWR plants. The Regulatory Guides have been reviewed for applicability to CRBRP and the NRC guidance is followed as appropriate to an L WBR. A summary of this applicability to CRBRP is provided in Table I to this Section. The Regulatory Guides have been designated: applicable in total, applicable in principle and Intent, or application to CRBRP is inappropriate. Table I will be revised in a f uture amendment to add ref erences to PSAR Sections where specific applicability of the Regulatory Guide is discussed. References to Section 1.1
- 1. Letter from R. P. Denise (NRC) to L. W. Caffey (Director of CRBRP Project), May 6, 1976.
- 2. Final Environmental Statement for the Clinch River Breeder Reactor Plant (Docket No. 50-537), NUREG-0139, February 1977.
1.1-5 Amend. 64 Jan. 1982 i
TABLE 1.1-1
SUMMARY
OF DESIGN SAFETY APPR0ACH FOR THE CRBRP This table represents the CRBRP Project Design Safety Approach. 41l 1. The following three levels of safety approach used in CRBRP design is generally consistent with the three levels of safety concept used by Regulatory to evaluate the adequacy, for licensing purposes, of nuclear power reactors. However, beyond these three levels, the CRBRP approach includes structural and thermal margins beyond the 41 design base as indicated in reference 10 of Section 1.6.
- a. The first level focuses on the reliability of operation and prevention of accidents through the intrinsic features of the design, construction, and operation of the plant, including quality assurance, redundancy, testability, inspectability, maintainability, and failsafe features of the components and systems of the entire plant.
- b. The second level focuses on the protection against Anticipated' Faults and Unlikely Faults (as defined in Table 1.1-1 A) which might occur despite the care taken in design, construction, and operation of the plant set forth in Level One above. This protection will ensure that the plant is placed in a safe g condition following one of these faults. T
- c. The third level focuses on the determination of events 42l to be classified as Extremely Unlikely Faults (as defined in Table 1.1-1A) and their inclusion in the design basis. Table 1.1-3 contains a list of such
" Extremely Unlikely Faults". These faults are of low probability and no such events are expected to occur during the plant lifetime. Even though they represent extreme and unlikely cases of failures, they have been analyzed using the same conservative assumptions as those employed in consideration of second level events. ,
41 1.1-6 Amend. 42 O Nov. 1977
}
v b] v TABLE I EVALUATION OF APPLICA81LITIES OF EXISTING NRC REGULATORY ",UIDES TO THE CLINCH RIVER BREEDER REACTOR PLANT TITLE ffPLtCABiLITY .1USTIF1 CAT lON1EXPLANATlON E NuleER 1.1 Net Positive Suction Head The Regulatory Guide has no CRBRP design precludes the need for for Emergency Core Cooling commonality with CRBRP design; an ECCS or other safety relateu and Containment Heat therefore, application to CRBRP systems with pumps whose performance Removal System Pumps is inappropriate. depends on containment pressure. 1.2 Thermal Shock to Reactor The Regulatory Guide has no The CRBRP Reactor Vessel is not sub-Pressure Vessels commonality with CRBRP design; jected to thermal shocks with signif f-theref ore, application to CRBRP cant potential to cause brittle is Inappropriate. fracture even af ter a 30 year accumu-
- lation of Irradiation damage.
2 1.3 Assumptions Used for The Regulatory Guide has no Refer to Regulatory Guide 1.4. The Evaluating the Potential commonality with CRBRP design; gulde Is appIIcable to BwRs. Rediological Consequences theref ore, application to OtBRP of a Loss of Coolant Acci- Is Inappropriate.
~ dont for Bolling Water
- Reactors.
~
l
- 2 1.4 Assumptions Used for Evalu- Applicable in principle and Accident conditions and source terms
[ ating the Potential Radio- Intent. delineated in Regulatory Position C.1 logical Consequences of a are specific to RfR technology. Loss of Coolant Accident to Positions C.2a through C.2f are Pressurized Water Reactors. applicable to OtBRP. 1.5 Assumptions used for Evalu- The Regulatory Guide has no These regulatory requirements are for ating the Potential Radlo- commonality with OtBRP design; BWR plants. Reg. Guides 1.109 and logical Consequences of a therefore, appIIcation to CRBRP 1.145 Identify the assumptions to be Steam Line Break Accident is inappropriate. used for radiological analysis of for Bolling Water Reactors. acc! dents. 1.6 Independence Between Redun- Applicable in Total. dant Standby (Onsite) Power Sources and Between Their Distribution Systems. 1.7 Control of Combustible Gas The Regulatory Guide has no CRBRP is not subject to a design Concentrations in Contain- commonality with CRBRP designs basis accidents with potential to ment Following a Loss of theref ore, appilcation to CRBRP produce significant amounts of @g Coolant Accident is inappropriate. hydrogen In the containment.
- s en
.ii.
e- - 1.8 Personnel Selection and Applicable in Total. 8o N4 Training
TABLE i (Continu:d) fly NUEER TITLF [PPL it'AB il I TY .f uSTIF ICAT ION / EXPLANAT I ON 2 1.9 Selection, Design and Applicable in Total. Qualification of Diesel-Generator Units Used as Standby (Onsite) Electric Power Systems at Nuclear Power Plants. 1 1.10 Mechanical (Cadweld) Spilces Applicable In Principle Superceded by R.G. 1.136, 1.142, 1.94 In Reinforcing Bars of Cate- and intent gory 1 Concrete Structures 1.11 Instrument Lines Pene- Applicable in total. trating Primary Reactor Containment Supplment to Safety Guide Appl! cable in total. 11, Back fitting Considera-tions
? .12 Instrumentation for Earth- Applicable in principle and The location restrictions spect-7 1 quakes Intent. fled for LWRs are not responsive 8 to the structural and equipment N diiferences in an LWBR. Appro-priate locations for CRBRP are identitled.
I 1.13 Spent Fuel Storage Applicable in principle ar'd Specific elements of the method of Facility Design Basis Intent. Implementation as described in this Reg. Guide are not compatible ulth a liquid metal coolant design. I 1.14 Reactor Coolant Pump Fly- The Regulatory Guide has no Coolant pumps in CRBRP are not wheel integrity commonality with mBRP design; provided with flywheels. theref ore, application to @BRP is Inappropriate. I 1.15 Testing of Reinforcing Bars Applicable In principle Superceded by R.G. 1.136, 1.142, 1.94 for Category 1 Concrete and Intent Structures 4 1.16 Reporting of Operat!ng Applicable in Total. Information - Appendix A Technical SpectiIcation Qg 1 1.17 Protection of Nuclear Plant Applicable in Total. pg Against industrial Sabotage
~ .cl. I 1.18 Structural Acceptance Test Applicable In Principle Superceded by R. G.1.136 $m NA for Concrete Primary Reactor and Intent.
Contalnments. O O O
)
L/ \ V TABLE I (Continued) fiL NLDGER J.11LE APPtICAB!L1TY 41SJJFfCATfON/ EXPLANATION 1 1.19 Nondestructive Examination Applicable in Principle Superceded by R.G.1.136 of Primary Containment Liner and Intent
. Welds 2 1.20 Comprehensive Vibration Applicable In principle and The high temperature sodlum environ-Assessment Program for Intent, mont in CRBRP precludes in place Reactor Internals During Inspection and the use of Preoperational and initial Instruments to measure UlS vibration.
Startup Testing. Extra emphasis is being placed on the vibration analysis program and the testing of scale models to compensate for this limitation. I 1.21 Noasuring, Evaluating, and Applicable in prir.sple and Evaluation and reporting of the Reporting Radioactivity in Intent quantity of radioactive offluents Solid Wastes and Releases pertaining to LWBR technology are of Radioactive materials in applied. Operating Information
-a Liquid and Gaseous Ef fluents is covered under R.G. 1.16.
from Light Water-Cooled 1 Nuclear Power Plants 1.22 Periodic Testing of Pro- Applicable in total. tection System Actuation Functions. 1.23 Onsite Noteorological Applicable in total. Programs. 1.24 Assumptions Used for Applicable in principle and Gas storage tanks for noble gas decay Evaluating the Potential Intent, are located inside controlled venti-Radiological Consequences lation protective structures. Off-of a Pressurized Water site radiological consequences Reactor Gas Storage Tank account for these protective Fallure, structures. 1.25 Assumptions Used for Evalu- The Regulatory Guide has no This R.G. discusses pool water absorp-ating the Potential Radio- commonality with CRBRP designs tion parag.eters and is primarily con-logical Consequences of a therefore, application to CFERP corned with 'odine-based thyrold doses Fuel Handling Accident in is inappropriate. to offsite Individuals. CRBRP does the Fuel Handling and Stor- not include a fuel storage (water age Facility for Bolling and pool) nor any other significant Iodine gg Pressurized Water Reactors, source.
?8 3 i.26 ovainty Group Ciassifications Appilcable In principle and CRBRP meets the Intent of this Guide g" and Standards for Water, Intent. as detailed in the response to NRC e Steam-and Radioactive- Question 001.274.
O$ Waste-Containing Ca ponents of Nucisar Power Plants. -
TABLE I fContinued) ELL NLM1FR TITL E APPLICABILtTY JUST !F ICAT f 0fiMXPL AP.AT I ON 2 1.27 Ultimate Heat Sink for Applicable in principle and CRBRP uttilres both water and air Nuclear Power Plants Intent. as ultimate heat sinks. 2 1.28 Quality Assurance Program Applicable in principle and See note 1. Requirements (Design and Intent. Construction) 3 1.29 Seismic Design Classifl- Applicable la principle and The Reg. Guide ref ers to LWRs and cation intent. and their technology. CRBRP utilizes LWBR equivalents. 1.30 Quality Assurance Require- Applicable in principle and See note 1. ments for the installation, intent. Inspection, and Testing of instrumentation and Electric Equipment. 3 1.31 Control of Ferrite Content Applicable In principle and The Intent is met by assuring a
- in Stainless Steel Weld Metal. Intent range of 5 to 9% delta ferrite.
2 1.32 Criteria for Saf ety-Rel ated Applicable in Total. e* Electric Power Systems for
# Nuclear Power Plants.
2 1.33 Quality Assurance Progran Applicable in principle and See note 1. Requirements (Operation). nient. 1.34 Control of Electroslag The Regulatory Guide has no This weld technique is not used on Weld Properties commonality with CRBRP design; saf ety rel ated equipment, however therefore, application to CRBRP If electroslag welding is used on is Inappropriate. saf ety related equipment the requirements of this R.G. will be met. 2 1.35 Inservice Inspection of The Regulatory Guide has no OBRP containment is not a pre-Ungrouted Tendons In commonality with CRBRP design; stressed concrete containment. Prestressed Concrete Con- therefore, application to (RBRP tainment Structures. Is inappropriate. 1.36 Normetallic Thermal Insuia- Applicable in Total. tion for Austenttic Stain-less Steel. C-. > - 1.37 Quality Assurance Require- Applicable in principle and
- m ments for Cleaning of Fluid Systems and Associated Intent. See note 1.
w ." Components of Water-Cooled y Nuclear Power Plants. N4 O O O
O O O TABLE I (Continued) E MDEER TITLE APPLICABillTY .!UST!FICATION/EXPLAMATION 2 1.38 Quality Assurance Require- Applicable in principle and See note 1 monts for Packaging, Intent. Shipping, Receiving, Storage, and Handling of items for Water-Cooled Nuclear Power Plants. 2 1.39 Housekeeping Requirements Applicable in principle and See note I for Water-Cooled Nuclear Intent. Power Plants. 1.40 Quellfication Tests of Applicable in principle and Applicable except accicent Continuous-Duty Motors Intent. quellfication environments will be Installed inside the based on CRBRP design basis events. Containment of Water-Cooled Nuclear Power Plants. 1.41 Prooperational Testing of Applicable In Total.
- Redundant Onsite Electric Power Systems to verify e Proper Load Group Assign- $ monts.
1.42 Interim Licensing Policy on Withdrawn by NRC As Low As Practicable for Gaseous Radlolodine Releases from Light-Water-Cooled Nuclear Power Reactors. 1.43 Control of Stainless Steel The Regulatory Guide has no The Regulatory Guide is concerned Wold Cladding of Low-Alloy commonality with CRBRP design; with SS Welds of low-alloy steel Steel Components. therefore, application to CRBRP components under water at high is inappropriate. pressures. The situations of concern as described in the Regulatory Guide do not exist in CRBRP 1.44 Control of the use of Appilcable in principle and Due to the high temperature of opera-Sensitized Stainless Steel. Intent. tion, scnsitization cannot be evolded. Low Internal system pressures and so-Icction of operating environments, including Insulation, will minimize the potential for stress corrosion cracking. $gN - 1.45 Reactor Coolant Pressure Applicable in principle and Leekage characteristics for CRBRP { Boundary Leakage Detection Intent. primary sodium coolant are not the
. Systems. same as for LWRs. Detectors and 8m N4 detection methods are different.
TABLE I (Continued) TITt F APPLICABILITY JUST IF I CAT f DN/ EXPLANAT I ON 11L Nun 43ER 1.46 Protection Against Pipe Whip Applicable in principle and CRBRP meets the Intent of this Reg-Inside Containment Intent. ulatory Gulce, however, the primary coolant system is a low pressur e system and not subject to pipe whip. 1.47 Bypassed and inoperable Applicable In Total. Status Indication for Nuclear Power Plant Safety Systems.
- 1.48 Design Limits and Loading Applicable In principle and The loadings associated with design Combinations for Seismic Intent. basis events In CRBRP are dif f erent Category i Fluid Systems f rom those associated with design Components. basis events in LWRs. Leakage from IHTS and SGS la the only leakage with significant potential to produce dynamic system loadings.
1 1.49 Power Levels of Nuclear Applicable In Total. 7 Power Piants. Y ~ - 1.50 Control of Preheat Tempera- Applicable In Total.
- ture for Welding of Low-Alloy Steel.
1.51 Inservice inspection of AS E Withdrawn by NRC Code Class 2 and 3 Nuclear Power Plant Components. 2 1.52 Design, Testing, and Mainte- Applicable in Total. nonce Criteria for Post Accident Engineered-Saf ety-Feature Atmosphere Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants.
- 1.53 Application of the Single- Applicable In Total.
Failure Criterion to Nuclear Power Plant Protection Systems.
'g 3ig" - 1.54 Quality Assurance Require- Applicable in principle and See note 1. .s ments for Protective Intent.
gP Coatings Applled to Water-e Cooled Nuclear Power Plants. Co cn r0 A O O O
O O O TABLE I (Continued) EfL useER TITl F APPLfCABIl_iTY .1USTfF1 CATION /EXPLANAttON 1.55 Concrete Placement in Applicable In principle Superceded by R.G. 1.136, 1.142, 1.94 Category l Structures and Intent. 0 1.56 Maintenance of Water Purity Applicable in pr!nciple CRBRP uses Sodlum Technology to in Bolling water Reacters and Intent, establish purity limits for the reactor coolant. 1.57 Design Limits and Loading Applicable in principle and Design limits and load combinations Combinations for Metal Intent. have been applied for CRBRP design Primary Reactor Contain- basis evants. ment System Components. I 1.58 Quellfication of Nuclear Applicable in principle and See note 1. Power Plant inspection, Intent. Examination, and Testing Personnel. 2 1.59 Design Basis Floods for Applicable In Total. Nuclear Power Plants. - Errata to Regulatory Applicable in Total. 1 N Guide 1.59. 1 1.60 Design Response Spectra App!Icable in Total. for Seismic Design of Nuclear Power Plants. 1.61 Damping Values of Selsmic Applicable In Total. Design of Nuclear Power Plants. 1.62 Manual Initiation of Pro- Applicable in Total. tective Actions. 2 1.63 Electric Penetration Assem- Applicable in Total. blies in Containment Struc-tures for Light-Water-Cooled Nuclear Power Plants. 2 1.64 Quality Assurance Require- Applicable In principle and See note 1. monts for the Design of Intent. Nuclear Power Plants, c3 1.65 Materials and Inspection for The Regulatory Guide has , The (RBRP reactor vessel closure su 3 Reactor Vessel Closure no commonality with CRBRP design; head Is not subjected to high
?8Q. Studs :..oref ore, application to CRBRP Internal pressures for normal Is Inappropriate. operation or design basis events.
e yQ - 1.66 Nondestructive Examination withdrawn by NRC of Tubular Products.
TABLE I (Continued) f1L NUMlER TITLE APf1 t CAB I L ITY _fuSTIFICATION/ EXPLANATION 1.67 Installation of Overpressure Applicable in Total. Protective Devices. 2 1.68 Initial Test Programs for Applicable in principle and CRBRP will use applicable portions Water-Cooled Nuclear Power Intent. of the Reg. Guides in developing an Plants. Initial startup test program tailored to an LWBR type plant. 1 1.68.1 Preoperational and initial The Regulatory Guide has no This guide is pertinent to BWR's only. Startup Testing of Feedwater commonality with CRBRP design; and Condensate Systems for theref ore, application to CRBRP Bolling Water Reactor is inappropriate. Plants. I 1.68.2 Initial Startup Test Program Applicable In principle and CRBRP will use applicable portions of to Omonstrate Romote Shut- Intent. the Reg. Guide in developing an down Capability for Water- Initial startup test program tailored Cooled Nuclear Power Plants. to an LWBR type plant. I - 1.69 Concrete Radiation Shields Applicable In Total. Y for Nuclear Power Plants. 3 1.70 Standard Format and Content Applicable in Principle and The CRBR PSAR used " Standard Format of Saf ety Analysis Reports Intent. and Content of Safety Analyst s Report for Nuclear Power Plants, for Nuclear Power Plants: "L WBR Edition" (Issued by NRC 1974). 1.71 Welder Qualification for Applicable in principle and See note 1. Areas of Limited Accessi- Intent, bility. 2 1.72 Spray Pond Piping Made Frm The Regulatory Guide has no There Ir. no spray pond in CRBRP. Fiberglass-Reinforced commonality with CRBRP design; Thermosetting Resin, therefore, appl: cation to CRBRP is inappropriata. 1.73 Quallfication Tests of Applicable in principle and Due to the low pressure sodlum coolant Electric Valve Operators Intent. systems located in inerted cells, Installed inside the Con- CRBRP is not subject to a LOCA In-cp tainment of Nuclear Power volving high containment pressure, yQ Plants. temperature, humidity and radiation.
. :s Other appropriate design basis events
." have been Identitled for CRBRP g equipment qualification, tu 4 O O O
TABLE I (Continued) P1L lameER TITLE APPLICABfLITY JusTIFICAT10fvEXPLANAT10N 1.74 Quality Assurance Terms and Appilcable in principle and See note 1. DefinltIons. Intent. 2 1.75 Physical independence of Applicable in Total. Electric Systems. 1.76 Design Basis Tornado for Applicable in Total. Nuclear Power Plants. 1.77 Assumptions Used for Evalu- AppI! cable In principle and Appropriate reactivity transients and ating a Control Rod Ejection Intent. analysis assumptions have been Accident for Pressurized specified for CRBRP. Water Reactors. 1.78 Assumptions for Evaluating Applicable In Total. the Habitability of a Nuclear Power Plant Control Roca During a Postulated Hazardors Chemical Release. w
- 1 1.79 Preoperational Testing of The Regulatory Guide has no The (RBRF 6esign precludes the need i Emergency Core Cooling commonality with OlBRP design; for an ECCS.
to Systems for Pressurized therefore, application to CRBRP Water, Reactors. Is inappropriate. 1.80 Prooperational Testing of Applicable In Total. Instrument Air Systems. 1 1.81 Shared Emergency and Shut- The Regulatory Guide has no (RBRP is a single unit plant. down Electric Systems for commonality with OtBRP design; Multi-Unit Nuclear Power therefore, application to CRBRP Plants. Is Inappropriate. 1.82 Sumps for Emergency Core The Regulatory Guide has no CRBRP design precludes the need for Cooling and Containment commonality with CRBRP design; either ECCS or a containment spray Spray Systems therefore, appI! cation to system or other safety related CfERP ls inappropriate. systems taking suction on the containment sump (s). I 1.83 Inservice inspection of Applicable In principle and The Steam Generators operate at gp Pressurized Water Reactor Intent. high temperatures and In liquid
- s 75 Steen Generator Tubes. metal environments. TSey have en
{ Inservice Inspection progr a e- . designed for an LWBR. m-
TABLE I (Continued) F1L NLDEER TITtF f?PLICABILITY .f usT IF ICAT I ON/ EXDL ANAT ION 18 1.84 Design and Fabrication Code Applicable in principle See components are built to a prev-Case Acceptability AS E and Intent. lous Regulatory Gulce revision. Section lli Division 1. 18 1.85 Materials Code Case Accepta- Applicable In principle Some components are bul d t to a prev-bility - ASE Section 111 and Intent. lous Regulatory Guide revision. Division 1. 1.86 Termination of Operat!n) Appilcable In Total. Licenses for Nuclear Reactors. 1 1.87 Guidance for Construction of Applicable in Total. Class 1 Components in Ele-wated-Temperature Reactors (Supplement to ASE Section Ill Code Classes 1592, 1593, 1594,1595, and 15%) . I 2 1.88 Collection, Storage, and AppIIcahls In principle and See note 1. 7 N Maintenance of Nuclear Power Intent. g Plant Quality Assurance Records. 1.89 Qualification of Class IE Applicable in principle and Due to the low pressure primary Equipment for Nuclear Power Intent. coolant system located in inerted Plants. cells, CRBRP is not subject to a LOCA Involving high containment pressures, temperatures, humidity and radiation levels. Other appro-priate design basis events have been Identlfled for (RBRP equipment qualification. I 1.90 Inservice inspection of Pre- The Regulatory guide has no (RBRP containment is not a prestressed stressed Containment Struc- commonality with (RBRP design; containment. tures with Grou ted Tendons. theref ore, application to CERP is inappropriate. 0 1.91 Evaluations of Explosions Applicable In Total. Postulated to Occur on Transportation Routes Near Nuclear Power Plants. @g 1 1.92 Combinir.g Moda; Responses Applicable In Total.
- 3 m and Spatial Components In R Selsmic Response Analysis.
cn - 1.93 Avallability of Electric Applicable in Total. NA Power Sources. O O O
v/ v \ . TABLE I (Continued) E KEEER I.11LE APPL 1CABlLfTY _fuKT1FfCATlON/EXPLAkATlON 1 1.94 Quality Assurance Require- Applicable in principle and See note 1. monts for Installation, Intent. Inspection, and Testing of - Structural Concrete and Structural Steel During the Construction Phase cf Nuclear Power Plants. 1 1.95 Protection of Nuclear Power Appilcable in principle Liquid or gaseous chlorine will not be Plant Control Roan Operators and Intent. used for treatment.of circulating Against an Accidental water and will not to stored un site. Chlorine Release Possible offsiite release control requirements elli be as per Reg. Guide 1.78. 1 1.% Desfon of Main 5 team iso- The Regulatory Guide has no GiBRP design precludes tt.e need for lation Valve Leakage Control ccamonality with OERP designs a main steam Iso ation valve leakage Systems for Bolling Water therefore, application to CRBRP control system or a similar . ~ Reactor Nuclear Power Plants. Is Inappropriate, system as part of the containment system. B N 2 1.97 Instrumentation for Light- Applicable In principle and The permeters to be monitored for - Water-Cooled Nuclear Power Intent. accidents in OERP are dlf ferent thes " ' Plants to Assess Plant and those for LWRs due to the dif ference - Environs Condiffons During in technology. Parameters appropriate and Following an Accident. for an LWBR will be monitored.
~
1.98 Assumptions Used for Evalu- Issued for comment. ating the Potential Radio-logical Consequences of a Radioactive Offgas System Failure In a Bolling Water Reactor. I 1.99 Effects of Residual Elements The Regulatory Guide has no The OERP reactor vessel does not on Predicted Radiation Damage commonality with DERP design contain ferrltic materials subject to Reactor Vessel Materials. therefore, applica+1on to OERP to significant Irradiation damage. is Inappropriate. Reactor vesset material In the belt-line region is austinitic stelnless steel not subject to brittle fracture under any normal or accident gg conditions. 3 to
- 3 1 1.100 Seismic Qualification of Applicable In Total.
~P Electric Equipment for y, Nuclear Power Plants. N4
TABLE I (Continued) TITLF APPLlCAB1LITY JUSTIFiCATfON/EXPtANATfON ELY NUM1E.R 2 1.101 Emergency Planning for Applicable in Total. Nuclear Power Plants. 1 1.102 Flood Protection for Applicable In Total. Nuclear Power Plants. Post-Tensioned Prestressing The Regulatory Gusde has no CRBRP does not utIIIzo a post-1 1.103 Systems for Concrete Reactor commonality with CRBRP design, tenstoned prestressing systm. Vessels and Contalnments. theref ore, application to OBRP is inappropriate. 1.104 Overhead Crane Handling Withdrawn by NRC Systems for Nuclear Power Plents. 1 1.105 Instrument Setpoints Applicable in Total. I 1.106 Thermal Overload Protection Applicable in Total.
.* for Electric Motors on L Motor-Operated Valves.
1 .1 07 Qualificatloas for Cement The Regulatory Guide hus no Cement grouting for pre-stressing m 1 commonality with CRBRP design; tendons in containment structures is Grouting for Prestressing Tendons In Conistnment therefore, application *o GBRP not used in CRBRP. Structures, is inappropriate. 1 1.108 Periodic Testing of Diesel Applicable in Total. Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants. Errata to Regulatory Guide 1.108. This Guide is met in practice and 1 1.109 Calculation of Annual Doses Applicable In principle and to Man From Routine Releases Intent. Intent. More advanced assessment of Reactor Effluents for the techniques are used as appropriate. Purpose of Evaluating Com-pliance with 10 CFR Part 50, Appendix l. S$ s ra - 1.110 Cost-Benefit Analysis for issued for comment Radweste Systems f or Light-
*d H. Water-Cooled Nuclear
$en Power Reactors.
N4 9 9 e
O O O TABLE I (Continued) EEL NinerR TiTtE ieeticasitity .iusviriciriouf oxrsixxvion 1 1.111 Methods for Estimating Applicable In Total. Atmospheric Transport and Dispersion of Gaseous Efflu-ents in Routine Releases from Light-Water Cooled Reactors. 1.112 Calculation of Releases of Appil:able In Total. Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors. I 1.113 Estimating Aquatic Disper- Applicable in Total. slon of Effluents from Accidental and Routine Reactor Releases for the Pur-pose of implementing Appendix
- 1. -
7 N 1 1.114 Guidance on Being Operator Applicable in Total.
" at the Controls of a Nuclear Power Plant.
1 1.115 Protection Against tow- Applicable in Total. Trajectory Turbine Missiles. l - 1.116 Quality Assurance Require- Applicable In principle and See note 1. ments for Installation, Intent. j inspection, and Testing of Mechanical Equipment and Systems. , 1 1.117 Tornado Design Classifi- Applicable in Total. cation. ! 2 1.118 Periodic Testing of Elec- Applicable in Total, j trir. Power and Prot 9ction Systems. l 1.119 Surveillance Program for Withdra.n by NRC i New Fuel Assembly Designs. cg
$o 1 1.120 Fire Protection GuideIInes issued for Comment.
! - m for Nuclear Power Plants. g, - 1.121 Bases for Plugging Degraded issued for Comment. ma PWR Steam Generator Tubes.
TABLE I (Continued) EEL 3)tBEB TITLE fffL ICAB il lTY JUSTIF ICATION/DPJEATl0N 1 1.122 Development of Floor Design Appilcable in Total. Response Spectra for Seismic Design of Floor-Supported Equipment or Conponents. 1 1.123 Quality Assurance RNulre- Applicable In prirciple and See note 1. ments for Control of Pro- Intent. curement of items and Services for Nuclear Power Plants. I 1.124 Service Limits and Loading Appitrable in principle and Stress limitations specified in the Combinations for Class 1 Inteni. R.G. do not apply to high temper-Linear Type Ocaponent eture operation and are appropriately Supports. adjusted. 1 1.125 Physical Models for Design f.pplicable In Total, and Operation of Hydraulic
~ Structures and Systems for Nuclear Power Plants.
I y 1 1.126 An Acceptable Model and The Regulatcry Guide has no For LWBR technology, experimental Relcted Statistical Methods commonality with GBRP design; results show negative densification. for the Analysis of Fuel theref ore, appilcation to CRBRP DensIfIcation. Is inapproprlate. 1 1.127 Inspection of Water-Contrct 9plicable In Total. Structures Associated =Ith Nuclear Power Plants. I 1.128 Ir.stellation Design and Applicable In Total. Installation of Large Lead Storage Batterles for Nuclear Power Plants. I 1.129 Maintenance, Testing, and Applicable in Total. Replacement of Large Lead Storage Batteries for Nuclear Power Plants. Ig 1 1.130 Service Limits and Loading Applicable In Total. s (D Ccubinations for Class 1 ~* 5 Flate-and-Shell-Type Ccun-ponent Supports. N$ 0 O e
[ ( t t U^ TABLE I (Continued) Ef1 luELER TITlF APPLtCAB1L1TY JUSTIFICATION / EXPLANATION 1.131 QualIfIcatton Tests of issued for Comment Electric Cables, Field Splices, and Connections
- for Light-Water-Cooled.
) Nuclear Power Plants.
1 1 1.132 Site Investigations for Applicable in Total. I foundations of Nuclear 4 Power Plants. I 1.133 Loose-Part Detection Program 'J. der Project Rev iew for the Primary System of Light-Water-Cooled Reactors. 1 1.134 Medical Evaluation of Applicable In Total. Nuclear Power Plant Per-sonnel Requiring Operator License.
~
L' - 1.135 Normal Water Level and issued for Comment Discharge at Nuclear
$ Powor P1 ants.
2 1.136 Meterial for Concrete Con- Under Project Review tainments (Article CC-2000 of the " Code for Concrete Reactor Vessels and Con-i tainments. I 1.137 Fuel-Oll Systems for Standby Applicable in Total. 4 Diesel Generators, l - 1.138 Laboratory investigations of issued for Comment. Solis for Engineering 4 Analysis and Design of Nuclear Power Plants. i 0 1.139 Guidance for Residual Heat issued for Comment. Removal. I 1.140 -Design, Testing, and Main- Applicable in principle and CRBRP has incorporated require-tenance Criteria for Intent. monts of Reg. Guide 1.52 gg _s o Lbrmal Ventilation Exhaust Srstem Air Filtration and w. { A> sorption Units of Light-Water-Cooled Nuclear Power 8m N4 PIents. l 4
TABLE I (Continued) EEY. NUMEER TITl F APPLICABILITY JUSTIFICATION /EXPLAriATION 1.141 Containment isolation issued for Comment. Prov!slons for Fluid Systems. I 1.142 Saf ety-Related Concrete Under Project Review. Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments). 1 1 .1 43 Design Guidance f or Radio- Applicable in Principle Calculated off-site releases of CRBRP, active Waste Management and Intent, during any accident conditions, are Systems, Structures, and well below 10 CFR 20 L imits. Components Installed in Light-Water-Cooled Nuclear Power Plants. I 1.144 Auditing of Quality Assur- Applicable in principle and See note 1. ance Programs for Nuclear intent. Power Plants. w 1.145 Atmospheric Dispersion issued for Comment. 8 Models for Potential Ei Accident Consequence Assess-ments at Nuclear Power Plants. 1 .1 46 Qualifications of Quality Applicable in principle and See note 1. Assurance Program Audit intent. Personnel for Nuclear Power Plants. 1.147 Inservice inspection Code Applicable in principle and Those portions of af f ected systems Case Acceptability-ASME Intent. that f unction at high temperatures Section XI DIvlsion 1 and in IIquid metal environments have the Inservice inspection programs designed for an LMFBR. 1.148 Functional Specification Under Project Review for Active Valve Assemblies In Systems important to Safety in Nuclear Power Plants. 1.149 Nuclear Power Plant Simu- Under Project Review cp lators for Use in Operating
@g Training. .m , f' -
1.150 Ultrasonic Testing of Under Project Review ua Reactor Vessel Welding E3C[ During Pre-Service and in-Service Examinations. O O O
o Note la The CABRP Quality Assurance Program complies wIth the Quality Assurance criterla of the Federal Regulations as defined in 10 Appendix B, "Quellfly Assurance Requirements for Nuclear Power Plants and Fuel Reprocessing Plants " The Project's method of complying with with RDT F2-2, 10CFR50,
" Quality Appendix Assurance B Is Program through the implementation of Its overall Quality Assurance progra developed in accordance Requirements."
The Progra also complies with supporting standards RDT F1-2,
" Preparation Progr of System Design Descriptions," RDT F1-3, " Preparation of Unusual Occurrence Reports," and RDT F3-2, " Calibrat e Requirments."
The various NRC Regulatory Guldes are not established as requirments for the CRBRP Quality Assurance Progre. However, the w progrm will accept requirements of RDTpractices that comply with these guldes as f ulfilling the Ilke requirements of RDT F2-2 Insof ar as the F2-2 are met. e and other nationally recogn! zed codes and standards as part of the implementation of the CRBRP
%rR om
- R.
CU O Nb
O V 1.2 GENERAL PLANT DESCRIPTION The Clinch River Breeder Reactor Plant is a liquid sodium cooled fast breeder reactor nuclear power plant. The major systems of the plant are the Reactor, Heat Transport and related systems, Steam Generator and related systems, Turbine / Generator and related systems, Fuel Handling System, Power Transmission and Plant Electrical System, Auxiliary Systems, and Instrumentation and Control System. The major plant operational parameters are given below. Thermal Output 975 MW (rated) Electrical Output 380 Mll Gross (rated) 350 MW Net Steam Production 3.34 x 10 6 lbs. p Steam Temperature at Turbine 900 F Steam Pressure at Turbine 1450 psi Total Primary Coolant Flow Rate 41.5 x 10 6fby Breeding Ratio (initial c,ycle) 1.29 (equilibriumcore) 1.24 Average Burnup (initial core) 50,000 MWD /T O (futurecore) 80,000 MWD /T 51 The thermal hydraulic design parameters for the plant are based on a thermal power rating of 975 MW. The recommended design margins create a possibility that the plant may attain a rated thermal power of 1121 MWt without exceeding the system design basis values. Accordingly, structural analysis for all non-replaceable systems is based on the 1121 MWt rating. The plant will initially be licensed for 975 MWt. 1.2.1 Site The Clinch River site is on a peninsula bounded on the south by tb Clinch River and on the north by the AEC Oak Ridge Reservation and within the city limits of Oak Ridge, Tennessee, The 1364 acre site is owned by the United States Government and is in the custody of the Tennessee Valley Authority. The site is 12 miles southwest of oowntown Oak Ridge and 25 miles west of Knoxville, Tennessee. The ground-level on which the CRBRP will be situated is 815 feet above sea level, 74 feet above the mean Clinch River water level of 741 feet. Figure 1.2-1 is an artist conception of the plant location. Amend. 51 Sept. 1979
~J 1.2-1 i
Tha Reactor Confinement / Containment Buildings are centrally located in a complex of major plant buildings. The complex is enclosed by a paved road. Tha cooling tower and circulating water pumphouse are located approximately 700 f eet east of the reacto- building. The power transmission f acilities are located approximately 900 feet northeast of the reactor building. The riverwater pumphouse is on the Clinch River northeast of the plant cunplex. Highway and railroad access will be f rom the North. 1.2.2 Enaineered Safety Features Th2 CRBRP design includes engineered safety features that are provided to mitigate the consequences of postulated accidents. These features are discussed in Chapter 6.0. Examples of such f eatures are: Containment / Conf inement Systems, Reactor Guard Vessel, Guard Vessels f or Primary Heat Transport System major components, Steam Generator Auxiliary Heat Removal System (SGAHRS) and Habitability Systems. 1.2.3 Reactor. Heat Tr3.nsoort and Related Systems A system of three identically configured piped circuits transport heat from the reacior, through primary and Intermodlate sodium loops, to steam generator modules which produce steam for the turbine. The three loops are Independent with only the Primary Heat Transport System (PHTS) having common flow paths through the reactor (See Figure 1.2-2) . The PHTS removes the heat generated in the fuel assembl ies, blanket assembl ies, control rod assembl ies, and structural el ements. Each of the three independent intermediate Heat Transport Systems (IHTS) receive heat f rom ihe PHTS through an Intermediate H:at Exchanger. The IHTS transfers heat outside containment with non-raJioactive sodium. The Intermediate Heat Exchanger acts as a barrier for the transfer of radioactive materials between the PHTS and lHTS. The IHTS is maintained at higher pressure than the PHTS to inhibit leakage from the radioactive PHTS into the non-radioactive IHTS. Each primary loop contains a hot leg pump, an interme.iate heat exchanger, a cold leg check valve and interconnecting piping between the above-mentioned components and the reactor vessel inlet and outlet nozzles. Each intermediate loop has a cold leg pump, Intermediate sodium expansion tank and interconnecting piping to transport the sodium from the tube side of the IHX to the superheater inlet and from the evaporator outlets back to the IHX tube side inlet (Reference Chapter 5 for datails of the Heat Transport System). The Primary Heat Transport System (PHTS) piping and components are located in calls within the containment building. The components and piping for each loop are located within three vaults (cells) in the containment building: (1) en HTS cell which contains all of the major loop components; (2) an lHTS pipeway; and (3) the reactor cavity which houses the reactor vessel and the r,ssociated primary loop piping. The cells are separated from each other by 9 1.2-2 Amend. 41 Oct. 1977
i (N (,,) concrete shielding walls and are inerted with nitrogen which is circulated for cooling. Those parts of the PHTS equipment which come in contact with sodium are located in a nitrogen atmosphere below the level of the containment 7 building operating floor. Each HTS cell has a separate atmosphere and the reactor cavity ano the HTS pipeways have a common atmosphere. The pump drive systems (motors, speed controllers, and heating and seal assemblies) are located in an air environment above the operating floor. Separation of the equipment cells provides the capability of deinerting Individual vaults for independent access for maintenance or inspections. l The reactor has 156 fuel assembiles, 82 inner blanket assemblies,126 radial blanket assemblies and 15 control assemblies (9 primary assemblies and 6 secondary assemblies). The reactor f uel assemblies are about 14 feet long, with an active core height of 36 Inches, upper and lower axial blankets of 14 inches each, and a fission gas plenum of 48 Inches. Each fuel assembly contains 217 stainless steel clad fuel pins. Each blanket assembly contains 61 stainless steel clad f uel pins. The fuel in the active core is mixed oxides of plutonium and uranium (Pu07/UO2 ). The blanket rocs are 116.5 inches loag with 64 Inches of depleted uranTum oxide pellets and a 48 inch long plenum. The control rod absorber material is enriched boron carbide (B4C). Each primary control assembly contains 37 absorber pins and each secondary assembly contains 31 pins. The core is designed for annual refueling. The cool ant f l ow is upwards through the core. The free sodium surface in the upper plenum is covered by argon. The reactor is located in a stainless steel reactor vessel of nominal inside O diameter 20'3", and 58'8" high from the bottom of the vessel to the top of the support ring. The vessel is provided wiih a closure head designed to accommodate through-the-head ref uel ing (See Figure 1.2-3) . The reactor vessel, lHX, and primary sodium pumps are enclosed by free-standing, structurally Independent guard vessels. Chapter 4 of this PSAR provides a thorough description of the reactor. 1.2.4 Steam Generator - Turbine and Related Systems The Steam Generator System provides independent steam gene ation capability for each of the three reactor heat transport loops. Steam, combined from all three loops, supplies the single turbine-generator. The Steam Generator System is of a modular forced recirculation configuration. The recirculation ratio is 2 to 1. Each of the three independent loops consists of the following: Steam Generator Evaporator /Superheater Modules Feedwater System Sodium-Water Reaction Pressure Relief System bv 1.2-3 Amend. 64 Jan. 1982
Water Isolation and Dump System Leak Detection System The steam generator evaporator /superheater is of the shell and tube type. Sodium flow is on the shell side, counter-flow to the water / steam flow on the tube side. The Feedwater System supplles feedwater to the steam drum where it mixes wIth and subcools the saturated water from the evaporators. The subcooled water flows through the tube side of the two evaporators where it is partially vaporized by the higher temperature sodium flowing on the shell side. The steam / water mixture then flows to the steam drum where the steam is separated end the water continues to recirculate. Entrained moistbre in the steam is removed by dryers and separators, Internal to the drum. The dry saturatei steam is superheated, to the desired temperature, in the tube side of the superheaters. The superheated steam flows to main steam headers and ther to the turbine. The Sodium-Water Reactor Pressure Ret lef System is a passive system whir n only becomes operational in the event of a steam tube leak within an evaporator or superheater module large enough to cause a rapid pressure rise due to the sodium / water reaction, in the event of a large sodium-water reaction, the system protects the sodium side of the evaporatcr and superheater modules, the IHTS and the IHX from over-pressure by the use of rupture discs on the piping adjacent to the modules. Sodium and/or sodium / water reaction products expelled through the rupture discs are directed by the Sodium-Water Reaction Rellef System piping to a separation tank where gross separation of IIquid, solid and gaseous products takes place. Gaseous reaction products (primarily hydrogen) then flow through a centrif ugal separator where additional separation takes place. The gaseous reaction products are then vented via a flare stack to the atmosphere. The flare stack burns any hydrogen that may be present in the gas. In order to reduce the amount of water which may be admitted to the IHTS in the event of a large sodium / water reaction in an evaporaive module, blowdown of the evaporator modules through power rellef/ safety valves is accelerated by the Water Dump System. Quick opening water dump valves are located at the inlet to each evaporator module. Water dump piping directs the water / steam to a water dump tank. In addition to the above, in the event of a large sodium / water reaction, sodium dump capability is provided for the Intermediate Heat Transport System (IHTS) and the sodium side of the Steam Generator System. The sodium can be drained rapidly to a sodium dump tank. The Steam Generator Leak Detection System monitors: (1) sodium exiting from l each evaporator and superheater. O 1.2-4 Amend. 64 J an. 1982
e, Monitoring is by hydrogen and oxygen meters which provide a measure of the ( hydrogen and oxygen concentration levels in the sodium, in the event of a water-to-sodium leak, changes in the hydrogen and oxygen concentration levels are detected and off-normal conditions are annunciated. The Steam Generator Auxiliary Heat Removal System (SGAHRS) removes reactor-generated heat following reactor shutdown or trip when the main NSSS heat sink or normal feedwater supply is unavailable. The heat is initially rejected to the environment through a direct steam dump and by condenser tubes heating atmospheric air for long term heat removal. The SGAHRS can also function to remove shutdown heat loads for refueling and other long term outages, s A Direct Heat Removal Service (DHRS) is also provided to remove decay heat. In the remote event that the steam generator decay heat removal paths are not available, the reactor decay heat will be' dumped through the reactor overflow system to an overflow heat exchanger. From the heat exchanger the heat flows through the ex-vessel storage tank (EVST) NaK system to air blast heat exchangers where the heat is transferred to the atmosphere (See Section 5.6.2 for details of the DHRS). 1.2.5 Offsite and Onsite Power The Offsite Power Systems deliver the power to and from the site and include transformers, switchgear, structures, overhead, and underground condult. Included are devices by which the main generator is connected or isolated from the TVA distribution grid. in addition, this system provides the unit station service transformers, reserve transformers, and related primary side y switchgear through which the station auxiliary loads are supplied power from the power grid. The Systems distribute and control the electrical energy for the site. The systems interface with the Offsite Power System at the secondary terminals of the unit station service and reserve transformers. The Onsite Power Systems provide the fofIowing functions:
- a. Receive power from the offsite power supplles and transforms the voltage to the utilization levels of the Nuclear Island, building, lighting and site service loads;
- b. Provide diesel generators for standby power; and batteries and invertors for vital AC and DC power;
- c. Provide cable, conduit, raceway, and shielded penetration systems for interconnecting wiring for electrical power and control, instrumentation, light'ng and communication; O
1.2-5 Amend. 64 Jan.1982
O
- d. Provide control and interlocking operations when these functions are not provided by other systems.
The Offsite and Onsite Power Systems are discussed in detail in Chapter 8. 1.2.6 Instrumentation, Control and Protection The plant control system integrates the various plant instrumentation and control systems into a coordinated, centralized control room and local control stations (as applicable), for the purpose of providing safe, effective and economic plant operation. The system also provides the control and instrumentation circuits and devices required to provide overall plant control which automatically maintains essentially constant steam temperature and pressure at the turbine inlet over the 40 to 100 percent load range of operation. The plant control system is designed to provide load follow capability as required in practical utility operation; that is, maintaining plant parameters within scheduled variations such that near constant temperature and pressure are traintained at the turbine inlet and rates-of-temperature change are minimized in system components during both plant steady-state and load change periods. Refer +.o Chapter 7 for complete description of these systems. g This system interfaces with the reactor and heat transport system through their instrumentation systems. Thus the plant control system utilizes signal inputs and control equipment provided by the reactor system, the heat transport system, the reactor and vessel instrumentation systems, fuel failure monitoring system, flux monitoring system, radiation monitocing system as well as the building electrical power system, power transmission system, auxiliary power system, steam turbine system, and data handling and display system. - The Plant Protection System (PPS) assures that the results of postulated plant fault conditions do not exceed the specified limits for the fuel or release of radioactivity by initiating reactor trip, sodium pump trip, turbine generator trip, containment isolation, or decay heat removal. The PPS includes the shutdown system, containment isolation l system and interfaces with the shutdown heat removal system. The PPS does
- 4) not require the reactor operator or control system to implement a protective action.
The shutdown system consists of two independent diverse systems, either of which is capable of reactor shutdown. The primary shutdown system is configured using a local coincidence logic arrangement while Amend. 64 Jan. 1982
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