ML20050T656

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Amend 51 to License DPR-68,changing Tech Spec to Incorporate Limiting Conditions for Operation During 5th Cycle & Reflecting Changes Resulting from Design Equipment & Procedural Mods Made During Refueling
ML20050T656
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 03/29/1982
From: Vassallo D
Office of Nuclear Reactor Regulation
To:
Tennessee Valley Authority
Shared Package
ML20050T794 List:
References
DPR-68-A-051 NUDOCS 8204150004
Download: ML20050T656 (52)
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{{#Wiki_filter:. i i ff o UNITED STATES g NUCLEAR REGULATORY COMMISSION y ) )7( [jl, s> p WASHINGTON, D. C. 20555 .l, g y c.J TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amencment No. 51 License No. DPR-68 1. The Nuclear Regulatory Commission (the Ccmmission) has found that: A. The application for anendment by Tennessee Valley Authority (the licensee) dated December 9,1981 c'onplies with the standards and requirements of the Atomic Energy Act of 1954, as anended (the Act), and the Comnission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amencment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;~ D. The issuance of this amendment will not be inimical to the cctmen defense and security or to the health and safety cf the public; and E. The issuanc.e of this amendment is in accordance with 10 CFR Part 51 of the Ccamission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License' No. DPR-68 is hereby amended to read as follows: (2) Technical Scecificaticns The Technical Specifica.tions contained in Appendices A and B, as revised through Amendment No. 51, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. o 8204150004 820329 PDR ADOCK 05000296 P PDR

I- ~ 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY C0ffilSSION Y4 A Donenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Attachnent: Changes to the Technical j Specifications Date of Issuance: March 21,1982' 3 1 i e i i l I 9 1 l = 1

e i ATTACHMENT TO LICENSE Af1EllDMENT NO. 51 l FACILITY OPERATING LICENSE fl0. DPR-68 DOCKET N0. 50-296 l Revise Appendix A as follows: Remove the following pages and replace with the identically numbered l. pages: ii 176 274 iii 178 275 vii 181 276 viii 182 277 i 18 182a 278 19 182b 279 24 192 280 26 195 2 81 27 224 282 29 225 285 30 227 286 95 231 123 246 124 262 126 263 133 264 146 264A 149 265 165 270 166 271 ( 167 272 i 167a 273 2. Marginal lines on each page indicate the revised area. 6 l J 1 I ~ l

e Section Page No., C. Scram Insertion Times 128 D. Reactivity A'nomalies 129 E. Reactivity Contral 129 3.4/4.4 Standby Liquid Control System 137 A. Normal System Availability 137 i l B. Operation with Inoperable Components 139 i C. Sodium Pentaborate Solution 139 3.5/4.5 Core and Containment Cooling Systems 146 A. Core Spray System 146 B. Residual Heat Removal System (RRRS) (LPCI and Containment Cooling) 149 C. RHR Service Water System and Emergency Equipment Cooling Water System (EECWS) 155 D. Equipment Area Coolers 158 E. High Pressure Coolant Injection System (HPCIS) 159 F. Reactor Core Isolation Cooling l System (RCICS) 160 l G. Automatic Depressuri:ation System (ADS) 161 H. Maintenance of Filled Discharge Pipe 163 I. Average Planar Linear Heat Generation Rate 165 J. Linear Heat Generation Rate '166 K. Minimum Critical Power Ratio (MC PR) 167 l L. Reporting Requirements 167a 3.6/4.6 Primary System Boundary 184 j A. Thermal and Pressurization Limitations 184 ii A' endment No. 51 m

e ~ 1 ) Pace No. Section B. Coolant Chemistry 187 C. Coolant Leakage 191 D. Safety and Relief Valves 192 193 E. Jet Pumps ( F. Recirculation Pu=c Operation 195 l G. Structural Integrity 196 l 198 H. Shock Suppressors (Snubbers) 231 3.7/4.7 Containment Systems 1 l A. Primary Containment 231 l 247 l B. Standby Gas Treatment System C. Secondary Containment ~ 251 D. Primary Containment Isolation Valves 254 E. Control Room Emergency Ventilation 256 258 F. Primary Containment Purge System I' Containment Atmosphere Dilution System (CAD) 260 i G. H. Containment Atmosphere Monitoring (CAM) System Hg and O Analyzer 261 g 3.8/4.8 Radioactive Materials 299 299 A. Liquid Effluents 302 B. Airborne Effluents 307 C. Mechanical Vacuum Pump D. Miscellaneous Radioactive Materials Sources 308-316 i 3.9/4.9 Auxiliary Electrical System 316 l A. Auxiliary Electrical Equipment 322 B. Operation with Inoperable Equipment I iii " Amendment No. 51 e =

e c ~ 4.2.E ' Minimum Test and Calltration Frequency f or Drywell Leak Detection Instrumentation 10 1

4. 2. F Minimum Test and Calibration Frequency for Surveillance Instrumentation 10 2 4.2.G Surveillance R?quiresents for Control Room Isolation Instrumentation 103 4.2.H Minimum Test and Calibration Frequency f or Flood Protection Instrumentation 104 4.2.J Seismic Monitoring Instrument Surveillance Requirements 10 5 4.6.A Reactor Coolant System Inservice Inspection Senedule 3.5.I MAPI.MGR vs. Averase Planar Exposure Ib,!6c., 1823 3.6.H Shock Suppressors (Snuhbers) 209 3.7.A Primary Containment Isolation Valves 262
3. 7. B Testable Penetrations with Double 0-Ring Seals 268 3.7.C Testable Penetrations with Testabl'e Bellows 269 i

3.7.D Primary Containment Testable Isolation Valves 3.7.E Suppression Chamber Influent Lines Stop-Check Globe Valve Leakage Rates 279 l

3. 7. T Check Valves on Suppression Chamber Influent Lines 280 3.7.G Check Valves on Drywell Influent Lines 281 e

3.7.H Testable Electrical Penetrations 283 c. 0. A Radioactive Liquid Waste Sampling and Analysis 310

4. 8. 5 Radioactive Geseous Waste Sampling and Analysis 311 l

6.3.A Protection Factors for Respirators 373 6.8.A Minimum Shif t Crew Requirements 390 vii Amendment No. 51 i

i e I u-LIST OF ILLUSTRATIONS Ed99.J.S 11133 ISSS s 2.1-1 APRM Flow Ref erence Scram and APRM Rod Block Settings 14 2.1 - 2 APRM Flow Bias Scram delationship to i Normal Operating Conditions 25 l 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Testo 48 a. 2-1 System Unavailability 117 J.u-1 Sodium Pentaborate Solution Volume Concentration Requirements 141 3.4-2 Sodium Pentaborate Solution Temperature R equir ements 142 l 182b 3.5.K-1 MCPR Limits i I 3.5.2 Kp Factor vs. Percent Core Flow 183 3.6-1 Temperature-Pressure Linitations 201 3,6-2 Change in Charpy V Temperature vs. Neutron Exposure 202 6.1-1 TVA Of fice of Power Organization for Operation of Nuclear Power Plants 191 6.1-2 Functional Organization 392 6.2-1 Review and Audit Function 393 6.3-1 In-Plant Fire Program Organization 394 1 e 4 'viii 1 -~ e e Amendment No. 51 N

a s z.1 2 Ay yj u_ _ LJ d1IING SAFCTY SYSTEM SETTINGS RELATED TO FU EL CLAQOING INTEGRITY The a bno r ma l operational t ransient s applicable to operation of the hr own s Fe r r y Nuc l ear Plant have been analyzed throughout the in-ctrum of planned operating conditions up to the design thermal power condition of 3440 MWt. The 4.vayses were based upon plant operation in accordance with the operating map given in Figure 3.7-1 of the FSAR. In addition, 3293 Mwt is the licensed maximum power level of Browns Ferry Nuclear Plant, and this represents tha maximum steady-state power which shall not knowingly be l exceeded. Conservatism is incorporated in the transient analyses in estimating the controlling f actors, such as void reactivity coefficient, control rod scram worth, scram delay tire, pea ki ng factors, and axial power shapes. These f actors are selected conservatively with respect to their effect on the applicalbe transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic perf ormance. Results obtained f rom a General Electr te boiling wat er reactor have been compared with prediettons made by the model. The comparisions and results are l summarized in Reference 1, 2, and 3. The absolute value of the void reactivity coef ficient used in the analysts is conservatively estimated to te about 25% greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent l to approximately 80% of the total scram worth of the control l r od s. The scram delay time and rate of rod insertion allowed by j the analyses are conservatively set equal to the longest delay and slowast insertion rate acceptable by Technical Specifications as further described in Feference 4 The effect of scram worth, scram delav time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time r equirements f or 5% and 20% insertion. By the time the rods are 60% inserted, approximately four dollars of negative reactivity has been fnserted which strongly turns the transient, and accomplishes the desired ef fect. The times for 50% and 90% . insertion are given to assure proper ccmpletion of the expected pe r f ormance in the earlier portion of the transient, and to establish the ultimate f ully shutdown steady-state condition. For analyses of the thermal consequences of the transients a MCPR is conservatively assumed to exist prior to initiation of of the transients. This choice of using conservative values of controlling parameters and initiating' transients at the. design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels. See Section 3.5.K. I 18 l Amendment No. 51

In summary: 1. The licensed maximum power leve. is 3,293 MWt. ~ 2. Analyses of transients employ adequately conservative values of the controlling reactor parameters. 2. The abnormal operational transients were analy:cd to a power level of 3440 Mwt. O. The analytical procedures now used result in a more logical answer than the alternative method of assuning a higher starting power in conjunction with the expected values for the parameters. The bases for individual set points are discussed below: A. Neutron Flux Scram 1. APRM High Flux Scram Trip Setting (. un Mode) R The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in' percent of rated power (3,293 Hwt). Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the tine ccnstant of the fuel. The ref ore, during transients induced by disturbances, l the therral peser of the fuel will be less than that l indicated by the neutron flux at the scram setting. Analyses reported in Section N14 of the Final Safety Analysis deport demonstrated that with a 120 percent scram trip.settina, none of the abnormal operational transients analyzed violate the fuel safety limit and l there is a substantial margin from fuel damage. l Therefore, use of a flew-biased scram provides even additional ma rcin. Figure 2.1. 2 shows the flow biased l scram as a function of core ficw. t ~ l An increase in the APRM scram setting would decrease the l margin present before the f uel cladding integrity saf ety l limi t is reached. The APRM scram setting was determined j by an analysis of margins required to provide a reasonable ranqe f or maneuvering during operation. Reducing this operatino margin vould increase the frequency of snurices scramn, which have an adverse e f f e ct on reactor naf ety because of the resulting thermal stresceu. Thus, the APRM setting was nelected 10 Amendment No. 51

e e position, where protection of the fuel cladding integrity saf aty limit is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolaticn valve closure scram asoures the availability of neutron flux scram protection over the entire range of applicability o; the fuel cladding integrity safety limit. In addition, the isolation -valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With to percent of valve closure, neutron flux the scrans set at does not increase. I. J. & X. Reactor low water level set poin t for initiation of H PC I and RCIC. closino main steam isolation valves, and startino LPCI and core spray pcmps systems maintain adequate coolant inventory and provide These core cooling with the objective of preventing excessive clad The design of these systems to adequately t em pe ratu re s. perf orm the intended f unction is based on the specified low level scram set point and initiation set points. Tr ansient analyses reported in Section N14 of the FSAR demonstrate that these conditions result in adequate safety margins fcr both the fuel and the system pressure. L. References 1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," HEDO-10802, Feb., 1973. Generic Reload Fuel Application, Licensing Topical Report 2. NEDE 24011-P-A and Addenda. "Qualificatien of the One-Dimensional Core Transient Model fer 3. Boiling iater Reactor", SEDO-24154, NEDE-24154-P, October 1978. 4. Letter from R. H. Buchhol: (GE) to P. S. Check (NRC), " Response l to NRC requ'est for information on ODYN computer codel," t September 5, 1980. I l I + 24 ,' Amendment No. 51

a i .e i i LIMITING SAFETY SYSTEM SETTING 5 Ai g,Ty L t te1T i s i

1. t QAJTOR COOLANT S Y ST E.M
2. 2 RE AC"CR COOLANT SYSTEM f

INT EC W ITY Im*EG R ITY I Apel: cab: 11tv A plicability Apphes to limits on reactor Applies tc trip settings of the e x ia.- s y s te.7. pressare, instruments and devices witten are provided to prevent the reactor system safety limts from being exceeded. Oblective Cbiective To establish a limit below To define the level of the wnten the integrity of the process variables at which reactor coolant system is not automatic protective action is threatened due to an initiated to prevent the ove rpressure condition, pressure safety limit from bei~ng exceeded. Soecification S oc ci ! i ca ti on The limiting saf ety system l settings shall be as specified j A. The pressure at the lowest below: point of the reactor vessel shall not exceed 1,375 psig wnenever irradiated fuel is in the reactor vessel. 4. Nuclear system 1105 psig 1 ( t, relief valves 11 psi, l oren--nuclear valves) systern pressure 1115 P815 i L1 poi (4 i esiver) 1125 P818 1 11 poi (5 valves) 8. S c r a ts--nu c l e e r, j 1.05.% pais system high pressure 26 kilendment No. 51 i l 1 l

~ I SATETY :IMIT' LIMITING SAFETY SYSTEM SEITING i i i i i I } 4 i 4 4 1 i i All previous items on this page moved to page 26. l I I t I I 27 i This page intentionally left blank. l Amendment No. 51 i

e

s..

The safety limit of 1,375 psig actually applies to any point { in the reactor vessel;. however, because of the otatic water he4d, the highest pressure point will occur at the bottom of the vessel. Because the pressure is not monitored at this point, it cannot be directly determined if this safety limit has been violated.. Also, because of the potentially varying head level and flow pressure dr.Spo, an equivalent pressure cannot besa pfioris determined for a pressure monitor higher in the vessel. Therefore, following any, transient that is severe enough to cause concern that this safety limit wao violated, a calculation will be perf ormed using all available information to determine if the safety limit was violated. R T.P ER ENC ES 1. Plant Saf ety Analysis (BFNP FSAR Section N14.0) 2. ASME Boiler and Pressure Vessel Code Section III 3. USAS Piping Code, Section B31.1 4 Peactor Vessel and Appurtenances Mechanical Design (STNP TSAR Subsection 4.2) 5. Generic Reload Fuel Application, Licensing Topical Report, NEDE-240ll-P-A and Addenda. l l I l 29 A endment No. 51 l

i I .) 2.2 BASCS 1 REACTOR COOLANT SYSTEM INTEGRITY To meet the safety basis thirteen rel'ief valves have been installed on the unit with a total capacity of 83.77% of nuc1 car boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel ~ pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable everpressure limit of 1375 psig. To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve f ailure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1375 psig. i t ( l 30 l f Amendment No. 51 1

9 TABLE e.2.0 ScavLILLA: ace REC 4JIRLMttrTS ICH I riOTa tritH TAT 1Gr# TilAT INITIATE OR CotrT's)L Tl8E CSCS runction Ep,ct im3a1 Test calibtation Instrument Ciisci f, I ns t rum.en t Ch ann el Suppressimi Chamber High Level (1) once/3 mor.the nonc Instruarnt Channel Reactor Gigh Water Level (1) once/3 ponths once/ day Instrument Channel DCIC Tuttine steam Line siqh ric.w (1) once/3 months none Instruar.nt Channel DCIC Steam Line Spec fligh Tempe r a tu r e (1) once/3 months none i Instrument Channel v. Ht'C I Turbine Steam Line Higte Flow (1) once/3 mont hs none g Instnisent Channel IIPC I Steam Line Spee fligh Tem pe r a tu r e (1) once/3 months none core Spray systra torjic once/6 monthe (6) N/A RCTC System (Initlating) logic once/6 months il/A tt/A e RCIC System (Isolation) togic once/6 ranths (fi) ti/A f IIPCI System (Initiating) Lo}ic once/6 months (6) N/A 7= l UPCI System (Isolation) 149 c oncc/6 monthe ((3) N/A 1 E3 l ADS 1491c oncc/6 months (6) N/A (1. M !#CI (Ini ti a t i ng) Logic once/6 monthe (0) tt/A

s rt Z

6, O. U1 i s 4 e + / 9 s

l 1 i e-l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i i 4.3 REACTIVITY CONTROL 3.3 REACTIVITY CONTROL l 2. The control rod drive 2. The control rod drive housing support housing support ~ system shall be in system shall be place during reactor inspected af ter power operation or reassembly and the when the reactor results of the coolant system is inspection recorded. pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and S pecification 3.3. A.1 is met. 3. a. Whenever the 3.a Prior to the start of l reactor is in control rod the startup or withdrawal at i run modes below

startup, 20% rated power the pod Sequence

~ Control System (RSCS) shall be the operable-capability of the Rod except the RSCS censtraints Sequence Centrol may be suspended by neans of S ys t en (RSCS) and the Rod Worth Mininizer the individual red bypass to properly fulfill switches for their functions shall 1 - special criticality be verified by the tests, or following checks, 2 - control rod. scram timing per 4.3.C.I. b' hen PSCS is bypassed on individual rods for these exceptions RbH must be oper-able per 3.-3.B.3.c and a second licensed operator may not be used in lieu of RkH. ~ 123 Amendment No. 51

e ~ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRDIENTS 4.3 REACTIVITY CONTROL 3.3 REACTIVITY CONTROL b. During the shutdown pro-Sequence portion - cedure, no rod movement is Select a sequence and permitted between the testing attempt to withdraw a rod performed above 20% power in the remaining sequences. and the reinstatement of the Move one rod in a sequence RSCS restraints at or above and select the remaining 20% power. Alignment of rod sequences and attempt to groups shall be acconplished move a rod in each. Repeat prior to performing the tests. for all sequences. c. Whenever the reactor is in Group notch portion - For the startup or run modes each of the six comparator below 20% rated power, the circuits go through test rod worth minimizer shall be initiate: comparator inhibit; operable. A second licensed verify; reset. On seventh operator may verify that the attempt, test is allowed to operator at the reactor con-continue until completion is solo is following the control indicated by illumination of red program in lieu of RWM test complete light. except as specified in 3.3.3.3.a. b. Prior to attaining 20% rated power during rod insertion at shutdown, the tests in 4.3.B.3.a shall be performed to verify RSCS capa-bility. The capability of the rod worth c. minimizer (RWM) shall be verified by the following checks: 2 124 - Amendment No. 51

I e 1.IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i MU ONI' 4.3 REACTIVITY CONTROL -control rod. 5. Prior to ~ obtaining 20% rated power during rod insertion at s hut down, verify the latching of the proper rod group and proper. annunciation atter insert errors. d. When the RM! is not operable, a second licensed operator will verify that-the correct rod program is followed except as specified in 3.3.3.3.a. ~ 126 /dendment No. 51 l e e ga

regardless of the rod pattern. This is true for all normal and abnormal patterns including those which maximize individual control rod worth. At power levels below 20 percent of rated, abnormal control rod patterns could produce rod worths high enough to be of concern relative to the 280 calorie per gram rod drop limit. In this range the RWM and the RSCS constrain the control rod sequences and patterns to those which involve only acceptable rod worths. The Rod Worth Minimirer' and the Rod Sequence Control System provide automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. Ref. Section 7.16.5.3 of the FSAB. They serve as a backup to procedure control of control rod sequences, which limit the maximum reactivity worth of control rods. Exceot during specified exceptions when the Rod Vorth Minimizer is out of service. a second licensed operator can manually fulfill the control rod pattern conformance functions of this In this case, the RSCS is backed up by s yst em. independent procedural controls to assure conformance. The functions of the RWM and RSCS make it unnecessary to specif y a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 20 percent, these devices Above 20 force adherence to acceptable rod patterns. of rated power, no constraint on rod pattern is percent required to assure that rod drop accident consequences are acceptable. Control rod pattern constraints above 20 percent of rated power are imposed by power distribution requirements, as defined in Section 3.5.I, 3.5.J, 4.5.I, and 4.5.J of these technical Power level for automatic bypacu of the specifications. RSCS f unction is sensed by first stage turbine pres'sure. Because the instrument has an instrument error of 110 percent of f ull power the nominal instrument setting is 1 30 percent of rated power. Because it is alicvcble to bypass certain rods in the RSCS during scra time testir.g telev 20'; cf rated power in the startup er run modes, a cecc rd licen cd crerater is net an acceptable substitute fer the EW during thtr. testing. 4 The Source Range Monitor (SRM) system performs no i it has no scram automatic saf ety system functions; i.e., It does provide the operator with a visual function. indication of neutron level. The consequences of reactivity accidents are functions of the initial neutron flux. Tne requirement of at'least 3 counts per second assures that any transient, should it occur, or above the initial value of 10-8 of rated begins at power used in the analyses of transients from cold One operable SRM channel would be adequate conditions. to monitor the approach to criticality using homogeneous A minimum patterns of scattered control rod withdrawal. .[mendment No. 51 133

i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS \\_ %we i 3.5 CORE AND COtTT AINMENT 4.5 CORE AND COffrAINMEttr COOLING COOLING SYSTEMS SYSTEMS Applicability Applicability Applies to the operational Applies to the surveillance status of the core and requirements of the core and containment cooling systems. containment cooling systems when the corresponding limiting condition for operation is in effect. Obiective Obiective To assure the operability of To verify the operability of the core and containment the core and containment cooling systems under all cooling systems under all conditions for which this conditions for which this cooling capability is an cooling capability is an esstential response to plant essential response to plant abnormalities. abnormalities. Specification Speci fica tion A. Core Soray System (CSS) A. Core Spray System (CSS) 1. The CSS shall be 1. Core Spray System operable: Testing. (1) prior to reactor Item Frequencv i startup from a cold condition,

a. Simulated once/

Automatic O peratinc; or Actuation Oycle (2) when there is test irradiated fuel l in the vessel

b. Pump Once/

l and when the Operability month reactor vessel pressure is

c. Motor once/'

greater than Operated - month atmospheric Valve as specified in ~ Operability pressure, except specifications 3.5.A.2. v I 146 Amendment No. 51

e- _ LIMITING CONDITIONS FOR OPEPATION SURVEILLANCE 11tWEMC:TTS i 3.5 CORE AND CONTAINMEtM

4. 5 CORE AND CONTAINMENT COOLING SYSTEMS Co__OMN_G_g]ET_Eg f"sidual Heal Demoval B.

Residual Heat Removal b. System (RHRS) (LPCI and Eystem ( F HR S) (LPCI and Containment Cooling) Containment Cooling) 1.

a. Simulated Cnce/

1. The RHRS shall be Automatic Operating operable: Actuation Cycle Test (1) prior to a reactor startup

b. Pump Opera-Once/

from a Cold bility month Condition; or

c. Motor Opera-Once/

(2) when there is ted valve month irradiated fuel operability in the reactor vessel and when the reactor

d. Pump Flow Once/3 vessel pressure Rate Months is greater than atmospheric,
e. Testable Once/

check valve operatin'q except as specified in cycle specifications 3.5.B.2, through Each LFCI pu=p shall deliver 3.5.B.7 9,000 gp= against an indicated syste= pressure of 125 psig. Two p c.ps in the same lecp shall 2. With the reactor dehver.5,000 sp= against an vessel pressure less indicatea system pressure of than 105 psig, the 200 psig. RHR may be removed from service (except 2. / air tes: en the dry.eell and torus that two RHR pumps. headers and no: les shall be containment cooling cenducted once/5 years. A mode and associated water test cay be performed en heat exchangers must remain operable) for the torus header in lieu of the a period not to air test. exceed 24 hours while being drained of e 149 Aniendment No. 51

I c LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS s 3.5 COPE AND CONTAINMENT 4.5 CORE AND CONTAINMEfC COOLING CCCLING SYSTEM) SYSTEMS I. Avarace Pl a n a r Linear I. Maximum Averace Pla nar Heat Generation Rate Linear Heat Generation Rate (HA PU+G P) During steady state power operation, the Maximum The HAPLHGR for each type Average Planar Heat of fuel as a function of Ganeration Rate (KAPLHGR) average planar exposure for ea ch type of fuel as a shall be determined daily f unction of average planar during reactor operation exposure shall not exceed at 2 251 rated thermal 'the limiting value shown Power. in Tablec 3.$. I-1 through 3.5.I-6. If at any time during operation, it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not. returned to within the prescribed limits within two (2) hours, the reactor shall be trought to the Cold Shutdown condition within 36 hours. Surveillance ar.d corresponding action shall continue until reactor opera' tion is within the prescribed limits. 165 Amendment No. 51

LIMITING CC:iDITIONS FOR OPEPJsTICN SURVEILLANCE REQUIRE.u,gn73

4. 5 CORE AND COtTTAIE.Ety,CMIjG 1' 5 Co nt: Af40 CONTAINMENT E1EIEUE C00f,1 NG SYSTEMS J.

Linear Heat Generation J. Linear Heat Generation Rate (LHGR) , Rate (LHGRI During steady state power The LHGR operation, the linear heat shall be generation rate (LHGR) of checked daily during any rod in any fuel

  • reactor operation at 2 255 location shall not exceed.

rated thermal power. assembly at any axial 13.4 kW/ft. If at any time during operation it is determined by norr.al surveillance l that the limiting value l for LUGR is being l exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR in not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within l 1 36 hours. Surveillance and corresponding action shall ~tinue until 166 i reacto peration is within the prescribed limits. Amendment No. 51

i 1.IMITING CONDITIONS"FOR OPERATION SURVEILLANCE REQUIRDIEffS 3.5 CORE AND CONTAI!O!ENT 4.5 CORE AND CONTAIIO!ENT C00LINC C001.ING SYSTD'S SYSTEMS '3.5.K Minimum Critical Power Ratio 4.5.K Minimum Critical Power Ratio (MCpR) (MCPR) The minimum critical power ratio (MCPR) 1. MCPR shall be deterntned daily as a function of scram time and core during reactor power operation flow, shall be equal to or greater than at >25% rated thermal power and shown in Figure 3.5.K-1 multiplied by following any change in power the K shown in Figure 3.5. 2, where: level or distribution that would g cause operation with a limitin;; T = 0 or ave - B, whichever is control rod pattern as described T A TB greater in the bases for Specification 3.3. T =0.90 sec (Specification 3.3.C.1 2. The MCPR limit shall be deter-A scram time limit to 20% insertion mined for each fuel type 8X8, j from fully withdra.wn) 8X8R, P8X8F, from Figure 3.5.K-1 8[E(0053)Ref5-respectively using: / r I -0.710+1.65 N' ~ B T = 0.0 prior to initial scram , n; a. time measurements for the cycle performed in accordance with I ve = Specification 4.3.C.l. a 1., 6 n b. T as defined in Specification n = number of surveillance rod tests 3.5.K following the conclusion performed to date in cycle (in-of each scram time surveillanci cluding BOC test). test required by Specification < 4.3.C.1 and 4.3.C.2. T 1 = scram time to 207 insertion th from fully withdrawn of the i The determination of the limit l rod must be completed with 72 hour: l l of each scram time surveillanct l N = total number of active rods required by Specification measured in Specification 4.3.C. 4.3.C.1 at BOC l If at any time during steady state operation it is deternined by normal l sueveillance that,the limiting value l for MCPR is being exceeded, action ~ shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the presc'ribed linits within two (2) hours. the reactor shall be brought to the Cold Shutdown condition within 36 hours, surveillance and corresponding action shall continue until reactor l operation is within the prescribed limits. 167 i Amendment No. 51 l .. ~ ~ e.

= ~ * - LIMITJt:0 CONDIT10NS FOR OPERATION SURVEILT.ANCE REOUIRDfENTS 4.5 CORE AND CONTAIMfENT 3.5,C_0RE AND CONTAISMD:T COOLING SYSTDtS COOLING SYSTD!S L. Reporting Requirements If'any of the limiting values identified in Specifications 3.5.I, J, or K are exceeded and the specified remedial action is taken, the event shall be logged and reported in a 30-day written report. i l 167a Amendment No. 51 -_=u_,_ .: r

t l l i !O.49.iC5 testing to ensure that the lines are filled. The visual enecting will avoid starting the core spray or RHR system with a disenarge line not filled. In addition to the visual observation and to ensure a filled discharge line other than j prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line highpoint to supply makeup water for these syst' ems. The conden ca te nead tank located approximately 100 feet above the disena rge hlqh point serves as a backup charainq system when the pressure suppression chancer head tank is not in service. System discharoe pressure indicators are used to determine the water level above the discharge line high point. The indicators will reflect approximately 30 psig f or a water level at the nich point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure snat the discharge lines are filled. Wnen in their normal standby condition, the suction for the H PC I and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that the HPCI and RCIC disena roe pipino remains filled. Further assurance is provided by observing water flow from these systems high points monthly. j t I. gaminum Averace Planar Linear Heat Generation Eate (KAPLHCR) This specification assures that the peak cladding temperature f ollowing the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CTR 50, Appendix K. The peak cladding temperature following a postulated loss-of-coolant accident is primarily a f unction of the average heat ceneration rate of all tne rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to r od power distribution within an assembly. Since expected local variations in power distribution within a fuel asser.bly affect the calculated peak clad temperature by less than 1 20*r. relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sutficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit. The limiting value i for KAPLHGR is sho.n in Inbles 3 5.I-l through 6. The a reference Ealvses supporting these limiting values is presented in J. Linear Heat Generation Rate fLHGR) This specification assures that the linear heat generation rate in any rod is less than the, design linear heat 176 Amendment No. 51 9

g. ,,es, i \\ 4 h 9 As ts reported within 30 days. It must be recognzzed that triere is always an action which would return any of the L paraneters (tMPLHGR, LHOR, o r MC P E) to within prescribed 1intts, nanely ;:cwer reduction. Under most circuntances, t.is Vill not be the only alternative. M. References Ceneric Reload Tuel Application, Licensin apical Report NEDE 24011-P-A and Addenda. ~ 5. Letter from R. H. Buchhol: (GE) to P. S. Check (NRC), " Response to NRC request for information on ODYN computer model," Septem'oer 5, 1980. s 1 t l i ( l [ 1 173 C I . Amendment No. 51 l l l l

1 e I TABLE 3.5.I-1 MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE l Fuel Type: Initial Core - Type 1 Plant: BT-3 t; MAPLHCR Average Planar Exposure (kW/ft) (NWd/c) 11.2 I 200 11.3 1,000 11.8 5,000 12.1 1 10,000 12.3 15,000 l 12.1 20,000 11.3 25,000 10.2 l 30,000 9.6 l 35,000 l TABLE 3.5.I-2 I MAPLEGR VERSUS AVERA0E PLANAR EXPOSURE 4 Fuel Types: 8DRB265L and P8DRB2.65L. Plant: BF-3 l MAPLHCR I Average Planar Exposure (kW/ft) (NWd/t) 11.6 200 11.6 + 1,000 l 12.1 5,000 12.1 t ~10,000 l 12.1 15,000 11.9 20,000 11.3 25,000 10.7 30,000 l 10.2 35,000 9.6 40,000 The values in this table are conservative for'both prepressurized and non-pressurized fuel. I I 181 4 i Amehdment No. 51 1 1

r-T BLE 3.5.I-3 MAPLHCR VERSUS AVERAGE PLANAR EX?05URE Fuel Type: P8DRB299 Plant: BF-3 MAPLHCR Averar,e Planar Exposure (kW/ft) (?Nd/ t) 10.9 200 11.0 l 1,000 11.5 5,000 12.2 10,000 12.3 15,000 12.2 20,000 11.9 25,000 11.3 30,000 10.9 35,000 10.4 40,000 i 10.0 45,000 TABLE 3.5.I-4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE l Fuel Type: P8DRB284Z Plant: BF-3 MAPLHGR Avera;;e Planar Exposure (kW/ft) i (!Td/t) 11.2 200 11.2 1,000 11.7 5,000 12.0 10,000 l 12.0 15,000 11.9 20,000 11.3 25,000 10.8 30,000 10.4 35,000 9.9 40,000 9.5 45,000 182 i Amendment No. 51 l a _.4r=_w g my

j I e ~- TABLE 3.5.I-5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plants BF-3 Fuel Type: P8DRB283 (LTA) Average Planar Exposure MAPLHGR (mwd /t) (kW/ft) 200 11.2 1,000 11.2 5,000 11.7 10,000 12.0 15,000 12.0 20,000 11.9 25,000 11.3 30,000 10.8 35,000 10.4 40,000 10.0 45,000 9.5 i TABLE 3.5.I-6 l l MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE . Plant: BF-3 Fuel Type: P8DR3314 (LTA) Average Planar Exposure MAPLHCR (mwd /t) (kW/ft) 200 10.6 1,000 10.7 5,000 11.3 l 10,000 11.7 15,000 11.5 ' 20,000' 11.2 25,000 10.6 30,000 10.1 t l 35,000 9.7 40,000 9.3 45,000 8.8 l l 182a Amendment No. 51 l l e

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O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 l l Figure 3.5.K-1 i MCPR LIMITS 182b Amendment No. 51 t 1 i v q--wy-.-.4s ,-m.s,.7,. g

s* - LIMITING CCNDITIONS TCR OPERATION SURVEILLANCE REQUIREMENTS 1.6 P,01MAhf SYSTra BCUNQARy 4, 6 pp f MA gy SYSTIM SoyN0npY t D. Relief Valves 1. 1 When more than approrisately one-k$ k'n$) "to half of all relief UU' I' be failed, an orderly valves shall te shutdown shall be bench-checked or initiated and the replaced with a reacter depressurized bench-checked valve to less than 105 peig each operating cycle. within 24 hours. All 13 valves wall have been checked or replaced upon the completion of every second cycle. 2. Once durine, each operating cycle, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve. 3. O t. At least one relief v al v e shall be 4 disassembled and inspected each operating cycle. l l l 192 ' Amendment flo. 51

e-LIMITING CONDITICNS FCR OPEFATICN SURVEII. LANCE REQUIR.E!CNTS

1. 6 PRIKARY SYSTEM BOUT!DARY 4.6 PRIHARY SYSTF.M BOUN CARY F.

Redreulat ion Pumn noeration F. Recirculation Pump Operation 1. Recirculation pump speeds shall be checked and loqqed at least once per day. 1. The reactor shall not be operated with one rceirculation locp out of service for more than 24 hqurs. With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a hot shutdown condition within 24 hours unless the loop ic sooner returned to service. 2. Following one-pump operation, the discharge valve of the low speed pump may not be opened unlesa the speed of the f aster pump in less than 50% of its rated apeed. 195 Amendment No. 51

i i e-3.6/h.4 Ets I limi t sp*eitied f or unidentified leakaqc, the probability is small that imperf ections or cracko associated with such leakage would orow rapidly. However, the establishment of allowabic unidentified leakage greater than that given in 3.6.c on the basis of the data presently available would be pr'emature because of uncerta anties associated with the data. For leakage of the order of 5 e pm, as specified in 3.6.C, the experirental and' analytical data suqqest a reasonable marcin of safety that such Ic4kage magnitude would not result f rom a crack approaching the critical size for rapid propagation. Leakage less than the magnitud. opt.-iiied can be detected reasonably in a matter of few hours utilitinu the available* leakage detection schemes, and if the origin cannot be determined in a reasonably short time the uni t should be shut down to allow further investigation and corrective action. The total leakage rate consists of all leakaoe, identified and unadentified, which flows to the drywell floor drain and equipment drain sumps. The capacity of the drywell floor sump pump is 50 gpn and the capacity of the drywell equipment sump pump is also 50 qpm. Removal of 25 apm f rom either of these surps can be accomplished with considerable margin. a trERINets 1. Nuclear System Leakage Rate Limits (BTNR TSAR Subsection u.10) 224 ~ Amendment No. 51 vv--

e-l 3.6.D/4.6.D Relief Valves To meet the safety basis, thirteen relief valves have been installed on the unit with a total capacity of 83.77% of nuclear boiler rated i steam flow. The analysis of.the worst overpressure transient, - (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vesset pressure which, if a neutron flux scram is assumed considering 12 4 valves operable, results in adequate margin to the code allowable overpressure limit of 1375 psig. j To r.eet cperational design, the analysis of the pInnt isolation trcnsient (generator load reject with bypass valve failure to open) j shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1375 psig. l i i 225 Amendment No. 51 ~ l t

I u-f .1. 6/ 4. 6 D AS D; s A nonle-tister ayutem f ailure could also qcnerate the coincident f ailure of a jo.t pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump dif f ucer body makes failure impossible without an initial noztle-riser system failure. J. 6. F/ 4. 6. F ttecirculation Pu-o Ooeration Steady-state operation without for:ed recirculation will not be permitted for.more than 12 hours. And the st; art of a recirculation pump from the natural circulation condition will not he pemitted unicss the temperature difference between the loop to be started and the core coolant temperature is less than 75 F. This reduces the positive reactivity insertion to an aceittably low value. l the discharge valve of the lower speed loop to remain Requiring cloned until the speed of the faster pump is.bclow 50% of itr. f rom one to'two pu:np rateil speerl provides assurance when goingexcessive vibration of the jet punp o pu r.it io n that not occur. i l

3. 6. G/rs. 6. G Structural Intenrity The requirements for the reactor coolant systc:ns inservice inspection pr ogram have been identified by evaluating the need for a sanglina examination of areas of high stress and highent probability of f ailure in the systes.. an'd the need to racct as closely as possible the requirements of Section XI, of the ESME Doller and Pressure Vessel Code.

The pecet r.im reflecto the built-in limitations of access to the reactor coolant s ys cens. 227 Amendment No. 51

LIMITING CONDITICHS TOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 g.or.r:anL*j;17_s.Yn7r s u.7 cou AIN*rNT systr~s 2 A_g,p ! t e a D t l i t Y ADDlicability Applies to the operating status Applies to the primary and of the primary and secondary secondary containment containment systems. integrity. ~Cbiact i ve o t.1 ee t i v_e To assure th= antegrity of the pr: mary and secondary To verify the integrity of the l centainment systems, primary and secondary t containment. seceif cation 3 ( l Seactftcation A. Primary ConM 12,mant A. Primary Containment 1. At any time that the irradiated fuel is in 1. Pressure sueeressice. the reactor vessel, chame.er and the nuclear systen is pr es suri:ed a. The suopression I above at.mospheric charter water level pressure or work is be checke(' once per being done wh en has day. Whenever heat th-potential t is added to the drain the vessel, the I pressure suppression serm W M h pool water level and testing of the ECCS temperature shall be or relief valves the maintained within the Pool temerature shall following limits be continually monitored except as specified and shall be observed in J.7.A.2. and legged every 5 minutes until the heat a. Minimu:n water level - addition is terminated. l -6.75" (differential pressure control >0 psid) l -7.25" (0 psid differ-ential pressure control) 1 I b. Max 1=u:s water level = -1" l l l 1 231 I i I Amendment No. 51 l

l. I.IMITING CCNDITIONS FOR OpERATICN SURVEILI.ANCE REQUIRE.'ENTS 3.7 COffrAI NMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS I s c. If the f specifications of 3.7.A.S.a through 3.7.A.S.b cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold shutdown condition within 24 hours. 6. Dryvell-Suppression Chamber 6. Drpell-Suppression Chamber Differential Pressure Differential Pressure a. Diff erential pressure a. The pressure diff er-between the dryvell and ential between the suppression chamber shall dryvell and suppression be maintained at equal chamber shall be recorded l to or greater than 1.) at least once each shift, psid except as specified j in (1) and (2) below: j (1) This dif f erential shall be estab-11shed within 24 hours of achieving operating te=perature and pressure. The differential pressure i =ay be reduced to l less than 1.3 paid

24. hours prior to a scheduled shutdown.

(2) This differential may be decreased to l 1ess than 1.1 psid for a maximum of four hours during required operability testing of the HPCI system, RCIC system, and the ,drpell-pressure suppression cha=ber vacuum breakars. \\ Amendment No. 51 N6 e =

TABLE 3.7.A PFittABY COtfrAIIStDtT 13)LATIOff VALVES Dumber of Power Maximum Act.lon on Operated Valves Operating flo n ma t InitiattN Inboard Outboard Time (sec.) Posation Signal Group Valve Iderati f ica t ion 4 4 3<T<5 0 GC 1 Main steamline toolation valves ( FCV 14, 26, 37, & Sl; l-15, 27, 38 & 52) 0 GC 1 1 15 1 Main steamline drain isolation valves (FCV-1-55 & 1-56) 1 Reactor Water sample line isola-1 1 5 C SC tion valves 2 RRRS shutdown cooling supply isolation valves (FCV-74-48 & 47) 1 1 40 C SC 2 30 c sc 2 RHRS - LPCI to reactor (rcV-74-53 6 67) 2 Reactor vessel head spray isola-tion valves (FCV-74-77 & 76) 1 1 30 C SC Sc' 2 RHRS flush and drain vent to 4 20 C M suppression chamber (FCV-74-102, 103, 119, & 120) 2 15 c SC 2 suppression Chamber Drain i (rev-75-57 5 5e) 2 Drywell equipment drain discharge isolation valves (PCV-77-15A & 15n) 2 15 0 GC r. 2 Drywell floor drain disetiarge 2 15 0 GC

M.

1 solation valves (FCV-77-2A & 20) 3 rL El tv 3 rt M O e i 4 9 D

TABLE 3.7. A (Continued) Action os Maxlarure Nucber o* Fower Normal lait 14 ting Operating Signal _ Operated Velves Position Inboard Outboard Time (sec.)_ valve Identification Crous GC 0 Reactor water clesnup system supply 30 1 1 3 1 solation valves FCV-69-1, & 2 1 10 o GC CC FCV 73-81 (Eypass around FCV 73-3) 0 20 1 1 4 HFCIS steamline isolation valves 4 FCV-73-2 & 3 0 CC. 1 1 15 actCS steamline toolation valves 5 FCV-71-2 & 3 SC inlet isola-C ] 6 Dryvell nitrogen purge 1 5 o g tion valves (FCV-76-18) SC C Suppression chamber nitrosen purge 1 5 6 inlet isolation valves (FCv-76-19) SC C isolation E. 6 Dryvell Main Exhaust 2 2.5 valves (l'CV-44-29 and 30) SC C Suppresutoo chamber o.atn exhaust 2 2.5 g ist.lation velvss (FCV-64-32 and 33) 6 r* 3C C Dryvell/ Suppression Chanlier purge 1 2.5 z ? 6 islet (icv-64-17) SC g c Drywell Atmosphere purge intet 1 2.5 g (h64-le) e L

TABLE 3.7.A (Continued) Act ion on Hunber of Power Maximua Operated Valves Operatine Normal

  • Initiating Position Signal __

Group Valve Identification Inboard Outboard Time (Sec.)_ 6 Suppression Chamber purr,c inlet (FCV-64-19) I 2.5 C SC 1 5 C SC 6 Drywell/ Suppression Chamber nitro-gen purge inlet (FCV-76-17) 't 6 Drywell Exhaust Valve Bypass to Standby Cas Treatment System 1 5 C SC (FCV-64-31) 6 Suppression Chamber Exhaust Valve bypass to Standby Cas Treatmpnt 1 5 C SC System (FCV-64-34) 6 System Suction Isolation Valves to Air Compressors "A" and "B" 2 15 0 GC (FCV-32-62, 63) 73 6 Drywell/ Suppression Chamber Nitrogen 1 5 C SC Purge Inlet (FCV-76-24) 6, Torus 11ydrogen Sample Line Valves 2 NA tio t e 1 SC Analyser A (FSV-76-55, 56) 6 Torus Oxygen Sample Line Valves 2 NA !!ote 1 SC li Analyzer A @SV 53, 54) m 6 Drywell liydrogen Sample Line Valves I 1 NA Note 1 SC @y Analyzer A (FSV-76-49, 50) tt y 6 Drywell Oxygen Sample Line Valves 1 1 NA Note 1 SC Analyzer A (FSV-76-51, 52) 6 Sample Return Valves - Analyzer A 2 NA 0 GC (FSV-76-57, 58) 6 Torus llydrogen Sample Line Valves 2 NA flote 1 SC Analyzer B (FSV-76-65, 66) e

TABLE'3.7.A (Continued) Number of Power Maximum Action on Operated Valves Operating Normal Initiating Gro333 Valve Identification Inboard Outboard Time (Sec.) Position Signal 6 Torus Oxygen Sample Line Valvec-Analyzer B (FSV-76-63, 64) 2 NA Note 1 SC r. 6 Drywell llydrogen Sample Line Valves-Analyzer.B (FSV-76-59, 60) I 1 NA Note 1 SC 6 Drywell Oxygen Sample Line Valves-Analyzer B (FSV-76-61, 62) 1 1 NA Note ! SC 6 Sample Return Valves-y Analyzer B (FSV-76-67, 68) 2 NA 0 CC C 7 RCIC Steamline Drain (FSV 6A, 6B) 2 5 C SC 7 RCIC Condensate Pump Drain (FCV-71-7A, 7B) 2 5 C SC 7 IIPCI Ilotwell pump discharge isola-tion valves (FCV-73-17A, 17B) 2 5 C SC a ( '7 IIPCI steamline drain (FCV-73-6A, 6B) 2 5 0 GC El 8 TIP Cuide Tubes (5) I per '+ guide tube NA C GC O NOTE 1: Analyzers are such that one is sampling drywell hydrogen and oxygen (valves f rom drywell open - valves from torus closed) while the other is sampling torus hydrogen and oxygen (valves from torus open - valves from drywell closed) 9 I

TABLE 3.7.A PRIMARY CotrrAll# LENT IDoLATION VALVES Number of Power Maximum Action on operated valves ope ra t_ing Normal Initiating croop valve Idmtification Inboard outtoard Time (sec.) Position signal r. standby liquid costrul systna check valves (CV 6 3-526 & 525) 1 1 NA C Process Peedwater check valves 2 2 ttA 0 Process l (CV-3-550, 572, 554 8 568) control rod hydraulic return chect valves (CV-8 5-576 5 573) 1 1 NA 0 Process RERS - I PCI to reactor check

  • valves (CV-74-54 5 68) 2 NA C

Process Core Spray discharge to reactor check valves (FCV-75-26 and 54) 2 NA C Process 6 Drywell AP air compressor stiction valve (FCV 64-139) I 10 C SC 6 Drywell AP air compressor discharge I 10 C SC valve (FCV 64-140) 6 Drywell CAM discharge valves (FCV 90-257A and 257B) 2 10 0 CC {*. 6 Drywell CAM suction valves (FCV 90-254A and 254n) 2 10 0 GC an.ag 6 Drywell CAM suction valve l r+ (FCV 90-255) I 10 0 GC 2 O o

e TABli 3 7.D AIP. TF.STED ISOLATI0tl VALVES Valv.= Valve Idantifi er.tien s 1-14 Main Steam -15 Main Steam { 1-26 Main Steam 1-27 Main Steam 1-37 Main Steam 9-38 Main Steam 1-51 Main Steam 1-52 Main Steam 1-55 Main Steam Drain 1.56 Main Steam Drain .1-1192 Service Water 2-1383 Service Water 3-554 Feedwater 3-558 Feedwater 3-568 Feedwater 3-572 Feedwater 32-62 Drywell Compressor Suction 32-63 Drywell Comprescor Suction 32-336 Drywell Compressor Suction 32-2163 Drywell Compresser Suetion 33-1070 Service Air 33-785 Service Air 43-13 Reactor Water Sample Lince: 43-14 Reactor Water Sample Linen 63-525 Standby Liquid Control Discharge 63-526 Standby Liquid Control Discharge 64-17 Drywell and Suppression Ch6mber Air-Purgc Inlet 64-18 Drywell Air Purge Inlet 64-19 Suppressien Chamber Air Purgc Inlet 64-20 Suppression Chamber Vacuum Relier 64-c.v. Suppression Chamber Vacuum Relief 64-21 Suppression Chamber Vacuum Relier 64-c.v. Suppressien Chamber Vacuum Relief 64-29 Drywell Main Exhaust 64-30 Drywell Main Exhaust 64-32 Suppression Chamber Main Exhaust 64-33 Suppression Chamber Main Exhaust 64-31 Drywell Exhaust to Standby Cas Treatment 64-34 Suppression Chamber to Standby cas Treatment 64-139 Drywell Pressurizatien, Cc= pressor Suetion' 64-140 Drywell Pressurization, Compressor Discharge 68-508 CRD to RC Pump Scals 68-523 CRD to R'C Pump Seals 68-550 CRD to RC Pump Scals 68-555 CRD to RC Pump Scals, \\ 270 hmendment No. 51 =

~ TABLE 3.7.D AIR TESTED ISOLATION VALVES '.'a l v o Vn tve Identificat_ict; L9-1 RWCU Suppfy t-69-2 EWCU Supply 69,579 nWCU Return 69-624 EWCb Return 71-2 BCIC Steam Supply '!1-3 RCIC Steam Supply ~* 39 ECIC Pump Discharge ' ' ;. 0 ECIC Pump Discharge UTCI Steam Supply

  • .e-73-3 HPCI Steam Supply

' 3 - S '. HPCI Steam Supply Bypass

  • j. 4 HPCI Pump Discharge 73-45 HPCI Pump Discharge 74-47 SHR Shutdown Suction 74 4P EHR Shutdown Suction 74-641 DHR Shutdown Suction 74-662 RHR Shutdown Suction 76-17 Drywell/Suppressien Chamber Nitregen Purge Inlet 76-16 Drywell Nitrogen Puree Inlet 76-19 Suppression Chamber Purge Inlet 76-24 Drywell/ Suppression Chamber Nitrogen Purge Inlet 76-49 Containment Inerting 76-50 Containment Inerting 76-51 Containment Inerting i

76-52 Containment Inerting 76-53 Containment Inerting 76-54 Containment Inerting 7 6-9r, Containment Incrting 76-56 Containment Inerting "6-57 Containment Inerting i 76-58 Containment Inerting l 7 6 - 5,. Containment Incrting 76 60 . Containment Inerting

  • i6 C1 Containment Inerting
6-62 Containment Incrting 7h 63 containment Inceting 76 64 Containment Inerting 76 65 Containment Inerting I6 66 Centainment Inerting 76-67 Containment Iner, ting 76-68 Containment Inerting j

77-2A Drywell Floordrain Sump 77-25 Drywell Floordrain Sump 77-15A Drywell Equipment Drain Sump 77 15D Drywell Equipment Drain Sump Sh 5A Centainment Atecspheric Dilution 44 33 Containment Atmospheric Dilution 54 EC Containment Atmo:pheric Dilution 84 SD containment Atmospheric Dilution l Ch 19 Containment Atmospheric Dilution Main Exhaust to Stnnc'by Cas Treatment 34- 0 Amendment No. 51 271

TABLE 3.7.fi AIR TES,TED ISOLATION val.VES t ia),ve Valve Identifiestien l .14-f00 Main Exhaust to Standby Cas Treatment

  • I.-601 Main Exhaust to Standby Cas Treatment C4-602 Main Exhaust to Standby Cas Treatment

'4-603 Main Exhaust to Standby Cas Treatment 85-576 CRD Hydraulic Return j 10-254A Radiation Monitor Suetien i 'JC-254? Radiation Mcnitor Suetien 30-255 Radiation Monitor Suetien 90-257A Radiation Monitor Discharge 90-257B Radiation Monitor Discharge f i i 272 - Amendment flo. 51 uo., am -

  • ' " ^

d f I e I l l .i ; f 6 (DELETED) i l 1 i i r l l l 1 1 4 t 273-278 Amendment No. 51 - - -. +.., .-._.-l-':'

TABLE 3.7.E PRI*AP.Y C07. TAI;:2:Tr ISCLATIC:f VALVES WIIICf! TIR:!II: ATE DELCtl 71.~: SU,FFRESSIO T POCL WATER I2: VEL - g. i: } Vdve Velve Identificotto?: 12-733 Auxiliary Boiler to RCIC 12-7L1 Auxiliary Boiler to RCIC h3-23A RIE Suppression Cha.'aer Scmple Lines k3-2% Ria Suppression Cheiaer Saa:ple Lines h3-2pA R*3 Suppression Cha:ber Sample Lines L3-273 R:G Suppression Chtcber Somple Lines 2-1143 Deminerr,11::ed Water 71-14 RCIC Turbine Exhaust 71 32 BCIC Vacuum h p Discherce 71-520 RCIC Turbine Exhaust 1-592 '.iCIC Vacuum h :p Discherce 73-23 IIICI Turbine Exhcust '3-2k IIICI Turbine F.xhoust Drain 73-6C3 HPCI Turbine Exhaust 73-609 IIPCI Exhaust Droin [ Th '22 RHR 75-57 Core Spray to Auxiliary Boiler 75-53 Core Sprof to Auxiliary Boiler Core Spray to Auxiliary Boller l 1. Amendment No. 51 279 p ~ em~ e ao e-

~ s-TABLE 3 7.F PRD'.ARY CC TfAD':2:C ISOLATICIT VALVES LOCATED Di UATZ3 S?J6CD SIISMIC CIASS 1 LI"JS Valve Identifiention v31ve m3 LICI Dicchcrge 7h-53 R'S 7h-54 h RHR Suppression Cha-her Spray 7 -5? 74-50 NG Suppression Chamber Sprcy -h-60 R1G Dryvell Spray Th-Gi PJE Drywell Spray 74-67 R!S LICI Di: charge Th 38 MG LICI Discharge 71 R:E Suppression Chamber Spray "L-72 R1G Suppression Chamber Sprey 7k -h R13 Drywell Sprr.y "h 5 RHR Dryvell Spray -h-77 R!m Head Spray 7h,-78 RH3 Head Spray 75-25 Core Spray Discharse 75-2'i Core Spray Dischstge Core Spray Discharge 75-53 75-5h Core sprey Discherce I 280 Amendment No. 51 f e %9m =- +* + ..-e

~- 1 TABLE 3.7.C I i (This table' intentionally left blank) = t i i ? l l 1 l l 1 i 281-282 A$endment No. 51 l Lh=%. _.

.e j 7.A L u.7.A Primary Containment The integrit'y of the primary containment and operation of the core standby cooling system in cornbination, limit the off-site 10 CFR 100 in the doseo to values less than those suggested in <= vent of a break in the primary system piping. 'Ihun, containment integrity is specified whenever the potential for violation of Concern about such primary reactor system integrity exists. the a violation exists whenever the reactor is critical and above An exception is made to this requirement atmospheric pressure. initial core loading and while the low power test program during is being conducted and ready access to the reactor vessel is this time, required. There will be no pressure en the system at ~ thus greatly reducing the chances of a pipe break. The reac ter however, restrictive may be taken critical during this period: operating procedures will be in ef f ect again to minini:e the Procedures and the Rod probability of an accident occurring. worth Minimizer would limit control worth such that a r would not result in any fuel damage. In addition, in the an excursion did occur, the reactor building unlikely event that and standby qas treatment sy s tem, which shall be operational during this time, of f er a suf ficient tarrier to keep offsite doses well below 10 CFR 100 limits. sink f or The pressure suppression ~ pool water provides the heata postulated the reactor primary system energy release following The pressure suppression chamber water the system. rupture of absorb the associated decay and structural sensible volume must released during primary system blowdown from 1,035 psig. heat Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant the pressure resulting f rom isothermal compression plus

accident, exceed 62 psig, the the vapor pressure of the liquid must n'ot The design volume of the suppression chamber maximum pressure.was obtained by considering (water and air) suppression chamberthe total volume of reactor coclant to be condensed is discharged to the suppression chamber and that the drywell volume that l

l is purged to the suppressicn chamber. . Usim; the mir.imum or maximum water levels given in the specification, con-tain ent pressure during the design basis accident is approximately 49 psig, the maximum of 62 psig. The maximum vater level indi-which is below cation of -1 inch corresponds to a downco=er sub=ergence of 3 feet 3 7 inches and a water volume of 127,800 cubic feet with or 128,700ft without the The cini=um drywell-suppression cha.ber differential pressure control. indication of -tr.25 inches with dif ferent ai l pressure con-water level-7.25 inches without dif ferential pressure control curresponds trol and to a downconer submergence of approximately 3 feet and a water volume of approxi s tely 123,000 cubic feet. Maintaining the water level the torus water volutne and do"n-l between these levels vill assure that comer submergence are within the aforementioned limits during nornal I Alarus, adjusted for instrument error, vill notify plant operation. the operator when the limits of the torus water level are approached. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to down-The maximum this specification is adequate. comer submergence, the end of blowdown tested during the Humboldt Bay temperature at anc Bodega Bay tests was 170*F and this is conservatively taken to for complete condensation of the reactor coolant, be the limit although condensation would occur for te=peratures above 170*F. 285 Amendment No. 51 o.. l <e.*. -ee m'.. e me , o-e ee.

Should it te necessary to drain the suppression chamber, thi: should only be done when there is no requirement for ccre st..ndby k cooling systens cperability. Under full power operation conditions, b1cwdown from an initial suppression chamber water temperature of 95*F results in a peak long term water temperature of 1700F which is suf f,icient f or complete condensation. At this temperature and atmospheric pressure, the available NPSH enceeds that required by both the RHa and core spray pumps, thus there is no dependency on containment overpressure. i Ogeri.cntal cata ind cate that excessive stea:n ecndens'i.:;; leads can be avcyed if trs,:e&. te perat' re of tN suppressien pool is maintained te1~s 200 F incal. Specifications have been placed en the envelope of Macter operating conditicns so that the reacter can be depressurited in a timely unner to avoid the regine of potentially high suppression cha.ter leadings. I.imiting suppression pool temperature to 105*F during RCIC,.MFCI, or relief va h6 operation when decay heat and stored energy is removed from the primary system by diccharging reactor steam directly to the suppression chamber assures adequate margin for controlled blowdown anytime,durinc RCIC operation and assures margin f or ccmplete condensatien of steam from the design basis loss-of-coolant accident. I In additien to the li: nits on te.nper.4:ure of the suppression chr.:nber pool i water, cperating precedures define the action to be taken in the event a ~' relief valve inadvertently opens er sticks open. This action would include: i ( (1) use of all available cams to close the valve, (2) initiate superesucn pool water couling heat exchangers, (3) initiate reactor shutdwn, and (4) if other relief valvus are used to depressurize the reacter, their discharge shall be separated frsm that of the stuck-open relief valve to assure mixing cnd uniferraity of energy insertien to the poolu If a locs-of-ccolant accident were to occur when the reacter water temperature is celcw approxinately 330 T, the containnent l pressure will not exceed the 62 peig code pernissible prescure, l even if no ecndensation were to cccur. The naximum allowable pcol tem;e ra ture, whenever the reactor is above 212*F, shall be governed by this specification. Thus, specif ying water volume-temperature requirements applicable for reactor-water temperature above -2120F provides additional margin above that available at 3300F. l l In conjunction with the Mark I Containment Short Term Program, a plant unique l j analysis was perfonned (" Torus Support System and Attached Piping Analysis for l the Browns Ferry Nuclear Plant Units 1, 2, and 3," dated September 9, 1976 and l supplemented October 12, 1976) which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system j and.ittached piping. The maintenance of a drywell-suppression cha=ber differen-tial pressure of 1.1 psid and a suppression cha=ber water level corresponding to a downcomer submergence range of 3.0f. feet to 3.58 feet will assure the i integrity of the suppression cha=ber when subjected to post-LOG suppression pool-hydrodynamic forces. 2.5G Amendment No. 51 l l ,,m,.e. .ma w m-4P + * ' = -*}}

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