ML19312C063
| ML19312C063 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/05/1972 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7911280595 | |
| Download: ML19312C063 (150) | |
Text
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DUKE POWER COMPANY i OCONEE NUCLEAR STATION UNIT 1 i L APPLICATION FOR LICENSES
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Docket 50-269 j 5'k\\ TECHNICAL SPECIFICATIONS . k\ 3 , Y
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TA3LE OF CONTENTS
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Section Pace TECHNICAL SPECIFICATIONS 1 DEFINITIONS 1-1 1.1 RATED POWER l-1 1.2 REACTOR OPERATING CONDITIONS 1-1 1.3 OPERABLE l-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 1-2 1.5 INSTRUMENTATION SURVEILLANCE l-3 1.6 l-3 QUADRANT POWER TILT 1.7 CONTAINMENT INTEGRITY 1-4 1.8 A2 NORMAL OCCURRENCE l-E 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2*l-1 N_. 2.1 SAFETY LIMITS, REACTOR CORE 2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE 2.3-1 INSTRUMENTATION 3 LIMITING CONDITIONS FOR OPERATION 3 1-1 3.1 REACTOR COOLANT SYSTEM 3 1-1 3.1.1 coerational Ccroonents 3.1-1 1 3.1.2 Pressurization, Heatup, and Cooldewn Linitations 3.1-3 3.1.3 Minimum conditions for Criticality 3 1-8 3.1.4 Reactor Coolant System Activity 3.1-10 ) 3.1.5 Chemistry 3.1-12 1 3.1.6 Leakace 3.1-14
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3.1. 7 Moderator Temperature Coe f ficient of Reactivity i
4 o Rg-**nn P2-- 3.1.8 Single too, Rei rictions 3.1-19 3.1.9 Low Power Physics Testing Restrictions 3.1-20 har. il ~ 3.1.10 Cnntrol Rod operarten . 3.1-21 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS 3.2-1 3.3 EMERGENCY CORE CCOLING, REACTOR BUILDING COOLING, 3.3-1 REACTOR BUILDING SPRAY, AND PENETRATION ROOM VENTILATION SYSTEMS 3.4 STEAM AND POWER CONVERSION SYSTEM 3.4-1 3.5 INSTRUMENTATION SYSTEMS 3.5-1 3.5.1 Operational Safety Instrumentation 3.5-1 3.5.2 cpntrol Rod Group and Power Distribution Limits 3.5-7 3.5.3 Engineered Safety Features Protective Systen 3.5-10 Actuation Setpoints 3.5.4 Incore Instru entation 3.5-12 3.6 REACTOR Bt'ILDING 3.6-1
- 1. 7 AUXILIARY ELECTRICAL SYSTEMS 3.7-1 3,3 FUEL LOADING AND REFUILING 3.8-1 3,9 RADI0 ACTIVE EFFLUENTS 3.9-1 3.10 MAXTMUM POWER RESTRICTIONS 3.10-1 8
19.Ip g 3.11 REACTOR 3UILDING POLAR CRANE AND AUXILIARY EDIST 3.11-1 19. 3.12 SECONDARY SYSTEM ACTIVITY 3.12-1 4 SURVEILLANCE REOUIREMENTS 4.1-1 4.1 OPERATIONAL SAFETY REVIEW 4.1-1 4.2 REACTOR COOLANT SYSTEM SURVEILLANCE 4.2-1 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4.3-1 4.4 REACTOR BUILDING 4.4-1 4.4.1 Containment Leakage Tests 4.4-1 4.4.2 Structural Integrity 4.4-7 s 4.4.1 Fedrocen Purze System 4.4-11 4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR BUILDING 4.5-1 CCOLING SYSTEMS PERIODIC TESTING 11 Rev. 18 3/10/72 Rev. 19, 5/5/72 l
e . Pace 4.5.1 Emergency Core Coolina Systems 4.5-1 w la. 4.5.2 Reactor Building Coolina Svstems 4.5-4 4.5.3 Penetration Room Ventilation System 4.5-7 4.5.4 Low Pressure Injection System Leakage 4.5-9 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4.6-1 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7-1 4.7.1 Control Rod Drive System Functional Tests 4.7-1 4.7.2 Control Rod Pro 2 ram Verification (Group vs. Core 4.7-3 Pcsition) 4.8 MAIN STEAM STOP VALVES 4.8-1 4.9 EMERGENCY FEEDWATER PUMP PERIODIC TESTING 4.9-1 4.10 REACTIVITY ANOMALIES 4.10-1 4.11 ENVIRONMENTAL SURVEILLANCE 4.11-1 4.12 CONTROL ROOM FILTERING SYSTEM 4.12-1
4.13 FUEL SURVEILLANCE 4.13-1 4.14 REACTOR SUILDING PURGE FILTERING SYSTEM 4.14-1 5 DESIGN FEATURES 5.1-1 5.1 SITE 5.1-1 5.2 CONTAINMENT 5.2-1 5.3 REACTOR 5.3-1 5.4 FUEL STORAGE 5.4-1 6 ADMINISTRATIVE CONTROLS 6.1-1 6.1 CRCANIZATION, REVIEW, AND AUDIT 6.1-1 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL 6.2-1 OCCURRENCE 6.3 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS 6.3-1 ENCEEDED 6.4 STATION OPERATING PROCEDUEES 6.4-1 v
Rev. 18 3/10/72 111
6 . Page 6.5 STATION OPEFATING RICCRDS 6.5-1 b 6.6 STATION REPORTING' REQUIREMENTS 6.6-1 6.7 RADIOLOGICAL CONTROLS' 6.7-1 v l l l l l 1 l l l l v iv
LIST OF TABI.ES L Table No. Page 2.3-1 Reactor Protective System Trip Setting Limits 2.3-7 3.5.1-1 Instrument Operating conditions ).5-3 4.1-1 Instrument Surveillance Require.ents 4.1-4 4.1-2 Minimum Equip =ent Check Frequency 4.1-8 4.1-3 Minimum Sa:pling Frequency 4.1-7
- 4. ',1-1 Environmental Surveillance Progra: 4.11-2 6.1-1 Minimum Operating Shife Requirements 6.1-7 L
L v
LIST OF FIGURES v riguro Page 2.1-1 Core Protection Safety Limits 2 1-4 2.1-2 Core Protection Safety L1=its 2.1-5 2.1-3 Core Protection Safety Limits 2.1-6 2.3-1 Protective System Maxi =um Allowable Set Points 2.3-5 2.3-2 Protective System Maxi =um Allowable Sc Points 2.3-6 3.1.2-1 Reactor Coolant System Heatup Limitations 3.1-6 3.1.2-2 Ftactor Coolant System Cooldown Limitatiens 3.1-7 3.5.4-1 incere Instrumentation Specification Axial 3.5-14 1:Galance Indication 3.5.4-2 Incore Instrumentation Specification Radial 3.5-15 Flux Indication 3.5.4-3 Incore Instrunantation Specification 3.5-16
\m, 6.1-1 Station Organization Chart 6.1-5 6.1-2 Managenent Organization Chart 6.1-6 L
vi
INTRODUCTICN D These Technica 3pecifications apply to the Oconee Nuclear Station, Unit 1 and are in accordance with the requirements of 10 CFR 50, Secticn 50.36. The bases, which provide technical support or reference the pertinent FSAR section for technical support of the individual specifications, are included for informational purposes and to clarify the intene of the specification. These bases are not part of the Technical Specifications, and they do not constitute limitations or requirements for the licensee. L vii
L TECHNICAL SPECIFICATIONS 1 DEFINITIONS The following terns are defined for uniform interpretation of these speci-fications. 1.1 RATED POWER Rated power is defined as a steady state reactor core output of 2568 MW:. 1.2 REACTOR OPERATING CONDITIONS 1.2.1 Cold Shutdown The reactor is in the cold shutdown condition when it is subcritical by at least 1 percent ak/k and Tavg is no ore thza 200'?. Pressure is defined by Specification 3.1.2. 1.2.2 Hot Shutdown The reactor is in the hot shutdown condition when it is suberitical by at least 1 percent ax/k and T avg is at or greater than 5307. \-- 1.2.3 Reactor Critical The reactor is critical when the neutron chain reaction is self-sustaining and Keft = 1.0. 1.2.4 Hot Standbv The reactor is in the hot standby condition when all of the follcwing conditions exist:
- a. Tavg is greater than 525*?.
- b. The reactor is "just" critical.
- c. Indicated neutron power on the power range channels is less than 2 percent of rated power.
1.2.5 Power Operation The reactor is in a pcwer operating condition when the indicated neutron power is above 2 percent of rated power as indicated en the power range channels. v 1-1
1.2.6 Refueling Shutdown The reactor is in the refueling shutdown condition when, even with all rods removed, the reactor would be suberitical by at least 1 percent ik/k and the coolant temperature a: the low pressure injection pu=p suction is no more than 140*F. Pressure is defined by Specification 3.1.2. A refueling shutdown re-fers to a shutdown to replace or rearrange all or a portion of the fuel asse=- blics and/or control rods. 1.2.7 Refueling Operation An operation involving a change in core geometry by =anipulation of fuel or control rods when the reactor vessel head is recoved. 1.2.8 Refueling Paried Time between normal refuelings of the reacter, not to exceed la conths withcut prior approval of the AEC. 1.2.9 Startup The reactor shall be considered in the startup ade when the shutdown margin is reduced with the intent of going critical.
- 1. 3 GPERA3L2 A cocponent or system is operable when it is capable of performing its intended ,,,/
function within the required range. The component or systen shall be considered to have this capability when: (1) it satisfics the limiting conditions for operation defined in Specification 3, and (2) it has been tested periodically in accordance with Specification 4 and haa set its perfor:ance requirements. 1.4 PROTECTIVE INSTR'JMENTATION LOGIC 1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires, amplifiers and output devices which arc connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/or protection. An instru-nent channel may be either analog or digital in nature. J 1-2
1 f i g ,, 1.4.2 Reactor Pro'. .ctive System ^ The reactor protecti. system is shown in 71gures 7-1 and 7-6 of the FSAR. i It is that co=hination o." protective channels and associated circuitry which forms the auto =atic system that protects the reactor by control rod
; trip. It inc1rd^ the four protective channels, their associated instrument
) channel inputo. manual trip switch, all rod drive control protective trip
- breakers and activating relays or coils.
1.4.3 Protective Channel i A protective channel as shown in Figure 7-1 of the FSAR (one of three or one of four independent channels, complete with sensors, sensor power supply units, a=plifiers and bistable =odules provided for every reactor protective ' safety parameter) is a co=hination of instrument channels forming a single digital output to the protective system's coincidence logic. It includes a l shutdown bypass circuit, a protective channel bypass circuit and reactor trip nodule and provision for insertion of a du==y histable. 1.4.4 Reactor Protective Svstas Lonic i 4 This syste= utilizes reactor trip =odule relays (coils and contacts) in all four of the protective channels as shown in Figure 7-1 of the 7SAR, to pro-
- vide reactor trip signals for de-enargizing the six con
- rol rod drive trip
- breakers. The control rod drive trip breakers are arranged to provide a one i out of two times two logic. Each element of the one out of two ti=es two j .
logic is controlled by a separate set of two out of four logic contacts fron
- the four reactor protective channels.
1.4.5 Engineered Safety Features Systan j This system utilizes relay contact output from individual channels arranged in three analog sab-syste=s and two two-out-of-three logic sub-systems as shown in rigure 7-3 of the 75AR. The logic sub-system is wired to provide
- appropriate signals for the actuation of redundant Engineered Safety Features equipment on a two-of-three basis for any given parameter.
I 1.4.6 Degree of Redundancy The difference between the number of operable channels and the nu=her of
! channels which when tripped will cause an automatic system trip.
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1.5 INSTRUMENTATION SURVEILLANCE J 1.5.1 Trip Test A trip test is a test of logic elements in a protective channel to verify their associated trip action. 1.5.2 Channel Test A channel test is the injection of an internal or external test signal into the channel to verify its proper output response; including alarm and/or trip initiating action where applicable. 1.5.3 Instrument channel Check An instrument channel check is a verification of acceptable instrument performance by observation of its behavior and/or state; this verification includes comparison of output and/or state of independent channela measuring the same variable. 1.5.4 Instrument Channel Calibration An instrument channel calibratien is a test, and adjustment (if necessary),
- o establisa thac :ne channe_ cutput responds with acceptable range and accuracy to known values of the parameter which the channel censures or an accurate simulation of these values. Calibration shall enecmpass the #
entire channel, including equipment actuation, alarm, or trip and shall 'd be deemed to include the channel test. 1.5.5 Heat 3alance Check A heat balance check is a comparison of the indicated neutron pcVer and ccre thermal power. 1.5.6 Heat 3alance Calibration An adjustment of the power range channel a:Plifiers output to agree with the core thermal power as determined by a heat balance on the seccndary side of the steam generator considering all heat losses and additions. J 1-4
1.6 QUADRANT PCWER T!LT Quadrant to average power tilt la expressed in percent as defined by the following equation: ower in any core quadrant 1 100 Average for all quadrants To obtain quadrant power the outputs of the upper and lower detector sections will be averaged. If one of the core detectors is out of service, the three operable detectors or the incore detectors will be used in calculating the average. 1 1.7 CONTAINMINT INTEGRITY l Containment integrity exists when the following conditions are satisfied: 1'
- a. The equipment hatch is closed and sealed and both doors of the per-
' sonnel hatch and emergency hatch are closed and sealed except as in
! b below. j
- b. At least one door on each of the personnel hatch and emergency hatch is closed and scaled during refueling or personnel passage i through these hatches.
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- c. All non-autocatic containment isolatten valves and blind flanges )
J are closed as required.
- d. All auto =atic containment isolation valves are operable or locked l %
closed. l
' e. The containment leakage determined at the last testing interval satisfies Specification 4.4.1.
1.8 ABNORMAL OCCURRENCE An abnormal occurrence scans the occurrence of any plant condition that:
- a. Results in a protective instrumentation setting in excess of a Limiting Safety Systes Setting as established in the Technical Specifications, or
- b. Exceeds a Limiting Condition for Operation as established in the Technical Specifications, or
- c. Causes.a- d-~~" '*c-, any uncontrolled or unplanned release '
l of radioactive raterial.,from the site, or a - l ?. ~ , ' epm by,.** y l
- d. Results in engineered safety systeggcomponent malfunction or system l or component malfunction which cou2d render a safety system incapable or performing its intended safety function, or
- e. Results in abnormal degradation of one of the several boundaries l i which are designed to contain the radioactive =aterials resulting from the fissicn process, or L
- f. Results in uncontrolled or unanticipated changes in reactivity
, greater than 17. ak/k except for trip. i 1-5 1
- g. Results in or if uncorracted can rasult in an unsafe conditicn due to ,/
observed inadequacie; in the inplenentation of adninistrative c procedural centrols. 7g'
- h. Results in conditions arising frcn natural or offsite nan-nade avents that affect or threaten to affect che safe operation of the plant.
V l l l l l 1 i l l J 1-6 Rev. 18 3/10/72 l l
1 . . 1 I l 2 SAFETY LIMITS AND LIMITINC 'AFETY SY3 TIM SETTING 3
\. - 2.1 SAFETY LIMITS, REACTCR CO2I i App licab ili ty ,
Applies to reactor thermal power, reacter power i= balance, reactor coolant j system pressure, coolant temperature, and coolant flow during power operatien i of the p.lant. 1 Oj36c_tive ! To maintain the integrity of the fual cladding. 1 i Specification The combination of the reactor system pressure and crolant temperature shall not exceed tha safety limit as defined by the locus of points established in Figuru 2.1-1. If the actual pressure /te=perature point is belcw and to the right of the line, the safety limit is exceeded. i l The combination of reactor therral power and reactor pcwer imbalance (power i fraction in top half of core minus power fraction in tha bottom half of the
! core) shall not exceed the safety limit as defined by the locus of points j (solid line) for the specified flow set forth in Figure 2.1-2. If the ac t ua i-reacto r-thermal-p owe r/ re actor-pewe r-idb alance p oint is above the line j for the specified flow, the safety li=it is exceeded.
J Bases 1 To maintain the integrity of the fuel cladding and to prevent fission product j release, it is necessary to prevent overheating of the cladding under normal ] operat ing conditions. This is accomplished by operating within the nucleate ] boiling regime of heat transfer, wherein the heat transfer coefficient is 1 large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DN3). At this point there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure. Although DN3 is not an observable parameter during reactor operation, the observable parameters of neutron pcwcr, reactor coolant ficw, te=perature, and pressure can be related to DNS through the use of the W-3 correlation. (1) The W-3 correlation has been developed to predict DN3 and the location of DN3 for axially uniform and non-uniform neat flux distributions. The local DN3 ratio (DNBR), defined as the ratio of the heat flux that would cause DNS at a j particular core locatien to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DN3R, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.3. A DN3R of 1.3 corresponds to a 94.3% probability at a 99% confidence level that DNS l q will not occur; this is censidered a conservative margin to DN3 for all cpera-ting conditions. The dif ference between the actual core outlet pressure and
- the indicated reactor coolant system pressure has been censidered in determining i the core prctection safety limits. The difference in these two pressures is
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5 1 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1 '\-- 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power i= balance, reacter coolant i system pressure, coolant te=perature, and coolant flow during powsr operaticn of the plant. Objective l To maintain the integrity of the fuel cladding. 4 j Specificaticn
'he combination of the reactor syste= pressure and coolant te=perature shall
! not exceed the cafety limit as defined by the locus of points established j in Figure 2.1-1. If the actual pressure /ta=perature point is belcw and to the right of the line, the safety lirdt is exceeded. W l The combination of reactor ther=al pcwer and reactor power i= balance (power i f raction in top half of core minus power fracticn in the bottom half of the a core) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2. If the i actual-reactor-ther=al-pcwer/ reactor-power-1 balance point is above the line for the specified flow, the safety limit is exceeded.
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Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is acco=plished by operating within the nucleate i boiling regime of heat transfer, wherein the heat transfer coefficient is l large enough so that the clad surface temperature is only slightly greater 1 than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (UN3). At this point j there is a sharp reduction of the heat transfer coefficient, which vould result in high cladding te=peratures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, te=perature, and pressure can be related to Dh3 through the use of the W-3 correlation.(1) ' The W-3 correlation has been develeped to predict DN3 and the location of DN3 for axially uniform and non-uniform heat- flux distributions. The local DN3 j ratio (DNBR), defined as the ratio of the heat flux that would cause DN3 at a particular core locatien to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DN3R, during steady-state cperation, normal operational transients, and anticipated transients is limited to 1.3. A DN3R i of 1.3 corresponds to a 94.3% probability at a 99% confidence level that DN3 will not occur; this is considered a conservative margin to DN3 for all cpera-ting conditions. The dif ference between the actual core outlet pressure and the indicated reactor coolant systen pressure has been censidered in determining the core protecticn safety limits. The difference in these two pressures is
- s--
a 4 l 2.1-1 i
For each curve of Figure 2.1-3, a pressure-te=perature point above and to the left of the curve would result in a DN3R greater than 1.3 or a local
*-- quality at the point of minimus DN3R less than 15% for that particular reactor coolant punp situation. The 1.3 LN3R curve for four pump operation more restrictive than any other reactor coolant pump situation becu^ase 4 19 any prensure/terperature point above and to the lef t of the four pu:p curve will be above and to the 1cfc of the other curves.
'E FERENCES (t) FS AR , Se c t ion 3. 2. 3.1.1 (2) FSAR, Section 3.2.3.1.1.c (3) FS AR, Section 3. 2. 3.1.1.k (4) The following papers which were presented at the Winter Annual Maeting, ASME, November 18, 1969, during the "Two-phase Flew and Heat Transfer in Rod Bundles Symposium":
(a) Wilsen, et.al.
" Critical Heat Flux in Non-Uniicrm Heater Rod Sundles."
(b) Ge lle rs tedt , et.al.
" Correlation of a Critical Heat 71ux in a Bundle Cooled by Pressurized Water."
b 2.1-3
J 2600 i 2400 u
; 2200
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a ~ a a.
- 2000 .
3
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l l o a a 1800 /
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1600 560 580 600 620 640 ssa Reactor Outlet Temperature,*f CORE PROTECTION SAFETY LIMIT J 4 ;\ OCONEE NUCLEAR STATION
,cf Figure 2.1 . ;
2.1-4
c; c o:A K ., M i rv._ , ., y d 120
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f' g ONSR
' Limit
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-- 80 (2)
[ i 7 -- 1 60 (3) (4) i a - 40 V t i __ k I i ! t
-60 40 -20 0 +20 +4 0 +60 Reactor Power Imaalance, 3 CURVE REACTCR COOLANT FLC1 (LB/HR) 1 131.3 x 106 2 98.1 x 106 3 64.4 x 106 4 60.1 x 106 CO2E PROTECTION SAFETY LIMITS t< a .5 2:un*3 OCONEE NUC' .EAR STATION
'M Figure 2.1 - 2 2.1-5 9
l 2500 7 4 J 2400
% /
2200 , , E i y 2000 n.
; i 5 I E I o 1803
#iff a
1600 560 550 600 620 640 660 Raactor Outlet Tamperature,*F REACTOR COOLANT FLOW CURVE (LBS/HR) POWER PUMPS OPERATING (TYPE OF LIMIT) 1 131.3 x 106 (1005 ) 114", FOUR PUMPS (ONSR LIMii) ; 2 60.1 x 106 (45.65) B B", TWO PUMPS IN ONE LOOP (QUALITY LIMIT) l 3 98.1 x 106 (74.7%) 6 9 . 5", THREE PUMPS (DNBR LIMli) j 4 64.4 x 106 (49.05) GOS ONE PUMP IN EACH LOOP i (GUALITY LIMIT) CORE PROTECTION SAFETY LlMiiS J
,I ij OCONEE NUCLEAR STATION Figure 6.2.13
- 2. M 1
1 2.2 SAFETY LIMITS - REACIOR COOLANT SYSTEM PRESSURE j \-- Aoplicabilit/ Applies to the limit on reactor coolant system pressure. L Objective ] To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity. l Snecification d 2.2.1 The reactor coolant system pressura shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel. 2.2.2 The satpoint of the pressurizer code safety valves shall be in * ! accordance with ASME, P:ller and Pressurizer Vessel Code, Section III, Article 9, Su==er 1967. { Bases The reactor coolant system (1) serves as a barri2r to prevent radionuclides in the reactor coolant f rca reaching the atmosphere. In the event of a i fuel cladding failure, the r22ctor cc lant system is a barrier against tne release of fission products. Establishing a system pressure limit helps to assure the integtity of the reactor coolant system. The maxi =um transient g,, pre:ssure allcwable in the reactor coolant system pressure vessel under the ASME code, Section III, is 110% of design pressure. (2) The maxi =u= transient pressure allowable in the reactor coolant system piping, valves, and fittings ] ! under USAS Section 331.7 is 110% of design pressure. Thus, the safety limit of 2750 psig (110% of the 2500 psig design pressure) has been established. (2) The settingt, the reactor hi safety valves (2500 psig)(3)gh pressure have trip (2355 psig) been established to assure and never the pressurizer reaching the reactor coolant system pressure safety limit. The initial hydrostatic test was conducted at 3125 psig (125% of design pressure) to verify the integrity { of the reactor coolant system. Additional assurance that the Reactor Coolant pressure does not exceed the safeey limit is provided by setting the pressurizer electromatic relief valve at 2255 psig. RE FERENCES l (1) FSAR, Section 4 (2) FSAR. Section 4.3.10.1 i j (3) FSAR, Section 4.2.4 i 1 1 4
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J ) 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIV2 INSTRLT.:.NTATION
\-- Applicability 1 Applies to instruments monitoring reactor power, reactor power i= balance, l l reactor coolant system pressure, reactor coolant outlet tenperature, flow,
! number of pumps in operation, and high react - building oressure. i
, obj ec tive j To provide automatic protective action to prevent any combination of process j variables from exceeding a safety limit.
Specification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1. and Figure 2.3-2. The pump monitors shall produce a reactor trip for the following conditions:
- a. Loss of two pumps and reactor power level is greater than 55%.
1 b- Loss of two pumps in one reactor coolant loop and reactor power level is greater than 0.0%. (Reactor power icvel trip set point is recer to 55% for single loop operatien.) v c. Loss of one or two pumps during two-pu=p operation. Banen The reactor protective system ccnsist9 of fcur instrueent channels to monitor each of several selected plant conditions which will cause a reactor trip if j any one of these conditions deviates from a pre-selected operating range to 1 the degree that a safety limit may be reached. The trip setting limits for protective system instrumentation are listed in Table 2.3-1. The safety analysis has been based upon these protective system instrumentation trip set points plus calibration and instrumentation 4 errors. Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursiet. too rapid to be detected by pressure and temperature measurements. i During normal plant operation with all reactor coolant pumps operating, reactor l trip is initiated when the reactor power '1evel reaches 107.5% of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the =aximum actual power at which a tri ated could be 114%, which was used in the safety analysis.(4) p would be actu-w 2.3-1
e < Overpower Trip Based en Flev and Imbalance The power level trip set point produced by the reactor coolant system ficw is s_s/ based on a pcver-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high pcwer. The analysis in Secticn 14 demonstrates that the specified pcwer to flew ratio is adequate to prevent a DN33 of less than 1.3 should a low flow condition exist due to any electrical =alfunction. The power Icvel trip set point produced by the power-to-flow ratio provides both high power level and low ficw protection in the event the reactor power level increases or the reactor coolant ficw rate decreases. The pcwer level t rip set point produced by the power to ficw ratio provides overpower DNB pro-tection for all modes of pump operation. For every flow rate there is a maxi-mum permissible pcwer level, and for every pcuer level there is a minimum permissible low ficw rate. Typical pcwer level and lcw ficw rate co=binaticns for the pump situations of Table 2.3-1 are as follows:
- 1. Trip would occur when four reactor coolant pu=ps are operating if power is 110% and reactor flow rate is 100%, or ficw rate is 90.9%
and pcwer lavel is 100%. 2, Trip would occur when three reactor coolant pumps are operating if power is 62.5% and reacter ficw race is 75%, or ficw rate i; 68.2% and pcwer level is 75%.
- 3. Trip would occur when two reactor coolant pumps are operating in a single loop if power is 55.0% and the cperating loop ficw rate ,_ ,f is 50.0% or ficv rate is 45.5% and pcwer level is 50%.
4 Trip would occur when cne reactor coolant pump is operating in eacn loop (total of two pu=pa operating) if the power is 54.0% and reactor ficw rate is 49.0". or flow rate is 45.5% and the power level is 50%. For safety calculations the maximum calibration and instrumentation errors for the power level were used. The power-inbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either pcwer peaking kw/f t limits or DNBR limits. The reactor power imbalance (pcwer in the top half of cora minus power in the bottcm half of core) reduces the pcwer level trip prcduced by the power-to-ficw ratio such that the boundaries of Figure 2.3-2 are produced. The pcwer-to-flow ratio reduces the power level trip and associated reactor power-reactor pcwcr-imbalance boundaries by 1.10% for a 17. flow reduction. Puen Monitors The pump monitors prevent the minimum core DN3R from decreasing below 1.3 by f tripping the reactor due to the loss of reactor coolant pu=p(s). The circuitry l monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal dive. e from that of the power.-to-ficw ratio. The pump monitors also restrict t' ' power level for the number of l pumps in operation. l 2.3-2
. . .~ .- -- . . - - - . - - _ -- - - - - . - -. = . _. _ _ _ _ _ -
Reactor Coolant System Pressure N-- During a startup accident from icw power or a slew rod withdrawal frcm high power, the system high pressure set point is reached before the nuclear overpower trip set point. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design t ransient. (1) i The low pressure (1800 psig) and variable icw pressure (13.26T out - 5989) trip setting limits shown in Figure 2.3-1 have been established to maintain the DNB ratio greater in a pressure reduction. than(2 or*eg al to 1.3 for those design accidents that result : i !
- Due to the calibration and instrumentation errors the safety analysis used a !
j variable icw reactor coolant syste= pressure trip set point of (13.26T out -6029).
; Coolant Outlet Temperature l The high reactor coolant outlet te=perature trip setting limit (619 F) shown i
in FI;;ure 2.3-1 has been established to prevent excessive core coolant , temper.tures in the operating range. Due to calibracien and instru=entatica
- urrors, the safety analysis used a trip set point of 620 F.
i j Reactor Buildinz Pressure The high reactor building pressure trip setting linit (4 psig) prcvides y,, positive assurance that a reactor trip will occur in the unlikely event of less-of-coolant accident, even in the absence of a icw reactor coolant system pressure trip. i Shutdewn Byoass s I i In order to provide for control red drive tests, zero power physics testing, l and startup procedures, there is provision for bypassing certain segments of
- the reactor protection system. The reactor protection system seg=ents which can be bypassed are shewn in Table 2.3-1. Two conditions are i= posed
- when the bypass is used
i 1. By ad=inistrative centrol the nuclear overpewer trip set point j must be reduced to a value 15.0% of rated pcwer during reactor I shutdown.
- 2. A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed, j The purpose of the 1720 psig high pressure trip set point is to prevent nor=al operation with part of the reactor protection system bypassed. This high
, pressure trip set point is lower than the nor=al icw pressure trip set point so that the reactor must be tripped before the bypass is initiated. The over power . trip set point of 15.0% prevents any significant reactor pcwer fros
- being (( duced when performing the physics tests. Sufficient natural circu-lation would be available to re=ove 5.0% of rated pcwer if none of the
'-- reactor coolant pumps were operating.
J j 2.3-3
Single Loop Oneratien
'~
Single loop operation is permitted only af ter the reactor has been tripped. Af*er the pump contact monitor trip has occurred the following actions will permit single loop operation:
- 1. Reset the pump contact monitor power level trip set point to 55.0%.
2 Trip ene of the two protective channels receiving outlet temperature informaticn from sensors in the idle loop. Tripping one of the two protecticn channela receiving cutlet te=perature information from the idle loop assures a protective systa= trip logic of one out of two. REFERENCES (1) FSAR, Section 14.1.2.2 (2) FSAR, Section 14.1.2.7 (3) FSAR, Section 14.1.2.8 (') FSAR, Sectica 14.1.2.3 (5) FSAR, Section 14,1,2.6 J J 2.3-4
2500 y p ee l
- i
- 2300 -
5 - W o g 2l00 - 2 - o 0 m 1900 - 2 U n m i700 < y 1500 , , , , 5% 560 580 600 620 640 Reactor Outlet Temperature. F PROTECTIVE SYSTEM MAXIMUM ALLOTABLE SET POINTS v
/b,\
719aa; OCONEE NUCLEAR STATION
- Figure 2.3 1 2.3-5
I
' PWE LE/L, % .
- 120 J
Four Pump
, Set Points
- 100 Three Pump -- 80 Set Points 60 1
; Two Pump 1 Set Points l
. 40 ,
I (1) ONE PU'AP IN EACH LOOP (2) T'.10 PUMPS IN ONE LOOP
-- 20 s ! g t I !
-60 40 -20 0 +20 +40 +S0 Power imoalance, %
PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SET POINTS
'bi p t.??.'s, OCONEE NUCLEAR STATION Figure 2.3 - 2 2.3-6
~v
r Table 2.3-1 . Fea:ter Protective S tee Trtr _letib;L !! rite Two Reactor One Reactor Four Reactor three Reactor Coolant Pumps Coolant Pump Coalant Pumps Coolatt Pucps Operating in A Operating la Operating Operating Single Loop (6) Each Loop (Operating Power (Operating Power (Operating Power Shutdown (Operating Power Bypass RPS Segnent -100t Pated) -7 5% Rat t d) -46% Rated) -49% Rates) 107.5 107.5 107.5 107.5 5.0(3)
- 1. Euclear Power Max.
(I kated)
- 2. Euclear Pouer thx. Based 1.10 times flow 1.10 times flow 1.10 times flow 1.10 times flow Bypassed ;
19* co Flow (2) and Imbalance, minus reduction minua reduction minus reduction minus reduction due to imbalance due to imbalance due to tabelance m'ne to ichalance (I kated) NA 55Z (5) 55% Bypassed
- 3. Nuclear Power Max. Based NA on Pump Monitors (I, Rated) 2355 2355 2355 2355 1720(4) 4 High Reactor Coolant System Ptessure, psig, ham.
18DO 1800 1800 Bypassed
- 5. Low Reactor Coolant 1800
- e. Systea Pressure, psig.
Hin. (s w (13.26 Tout-5969)(1) (13.26 Tout-598))II) Bypassed
- 6. Variable Low Reactor (13.26 Tout-5589)(1) (11.26 Tout-5983)(1)
Coolant System Pressure, pulg. Mt3. 619 619 619 619
- 7. Reactor Coolant Temp. 619 F., Ham.
4 4 4 4
- 8. I!!gh Reactor Bu11 ding 4 Pressure, psig. Hex.
(1) Tout is in degrees Fahrenheit (*F). (5) Eractor power level trip set point produced Ly puap contact monitor reset to 55.01. (2) keactor Coolant System Flow, I. (6) Specification 3.1.8 applice. Trip one of the (3) Administratively controlled reduction set two protection thannels receiving outlet temper-
,p, a:ute infurwatton f rom sensars in the idle loop.
e only during reactor shutdowu. (4) Automatically set when other segments of the RPS are bypassed. o a Un D N ta .
. . . . emb w e . .e.<e *eg.. o.
_- . -. . . - . . ~ .-. - _ . . . - - - _ - i : 3 LIMITING CONDITIONS FOR OPERATl)N 1
- \--
l 3.1 REACTOR COOLANT SYST M App licab ility j Applies to the operating status of the reactor coolant systen. Objective { j 'o specify those limiting conditions fnr operation of the reactor coolant
- system components which cust be cet to ensure safe reactor operation.
dpecification 3.1.1 Operational Co:ponents
- a. Reactor Coolant Pumos
- 1. Whenever the reactor is critical, single pu=p operation shall be prohibited, single loop operatien shall be restricted to testing, and other pump combinations permissible for given pcwer levels shall be as shown in Table 2-3.1.
- 2. The boron cencontration in the reacter coolant system shall not be reduced unless at least one reactor coolant pu=p or one low pressure injection pump is circulating reactor coolant.
v
- b. Steam Generator
- 1. One steam generator shall be operable whenever the reactor coolant average temperature is above 250*F.
- c. Pressurizer Safety Valves j 1. All pressurizer code safety valves shall be cperable whenever the reactor is critical.
- 2. At least one pressurizer code safety valve shall be operable whenever all reactor coolant system openings are closed, except for hydrostatic tests in accordance with the ASME Section III Eoiler and Pressure Vessel Code.
Bases i. A reactor coolant pump or low pressure injection pump is required to be - l in operatien before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One low pressure injection pump will circulate the equivalent of the reactor coolant system volume in one half hour or less. (1) N I j 3.1-1 i l
i f 1 The low pressure injection system suction piping is designed for 300*F and 370 psig; thus the system with it. redundant components can remove f decay heat when the reactor coolant system is below this te=perature. (2,3) J i i One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pu=p energy, pressurizer heaters, and reactor decay heat. (4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for a rod withdrawal accident at hot shutdown. (5) The pressurizer code safety i valve lift set point shall be set ct 2500 psig i 1% allowance for error and l each valve shall be capable of relieving 300,000 lb/hr of saturated steam at ! a pressure no greater than 3% above the set pressure. l References 1 (1) FSAR Tables 9-11 and 4-3 through 4-7. (2) FSAR Sections 4.2.5.1 and 9.5.2.3. 4 (3) FSAR Section 4.2.5.4. (4) FSAR Sections 4.3.10.4 and 4.2.4. 4 s. 1 (5) FSAR Sections 4.3.7 and 14.1.2.2.3. s_ ,/ i 1 4 I i i
%d 3.1-2 l
3.1.2 Prenourirnrien. Featun. ,ng c3eie_e..n ,,,te3et n, I \-- Specification ! 3.1.2.1 Hydro Tests: For thermal steady state system hydro test the system may be pressurized l to the limits set forth in Specification 2.2 when there are fuel as-semblies in the core and to ASME Code Section III 11=its when no fuel asse=blies are present provided:
- a. Prior to initial criticality the reactor coolant system temperature is 113*7 or greater or
- b. After initial criticality and during the first two years l of operation the reactor coolant system tc perature is 215*F or greater.
3.1.2.2 Leak Tests
- a. Leak tests may be conducted undar the provi-tons of 3.1.2.1 a and b above or
- b. After initial criticality and during the first two years of oparation tha syst:2 =ay be tested to a pressure of 1150 psig provided that the system ta=perature is 175'? or greater, y,, 3.1.2.3 For the first two years of power operation (1.7 x 10 6 thermal megawatt days) the reactor coolant prassure and the systes heatup and cooldown rates (with the exception of the pressuri=er) shall be limited in accorcance with Figure 3.1.2-1 and Figure 3.1.2-2, and are as follows:
Heatup: Allowable combinations of pressure and temperature shall be to the right of and belcw the limit line in Figure 3.1.2-1. The heacup rates shall not exceed those shown on Figure 3.1.2-1. Cooldown: Allowable conbinations of pressure and te=perature for a specific cooldevn shall be to the lef t of and below the limit line in Figure 3.1.2-2. Cooldown rates shall not exceed those shown on Figure 3.1.2-2. 3.1.2.4 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the vessel shell is below 100*F. 3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 100F/hr. The spray shall not be used if the temperature dif ference between the pressurizer and the spray fluid is greater than 410*F. 3.1.2.6 k'ithin two years of power operation, Fi$ures 3.1.2-1 and 3.1.2-2 shall
*-' be updated in accordance with appropriate criteria accepted by the AEC.
3.1-3 gey, 19 May 5, 1972 (Entire Page Revised)
Bases d All reactor coolant system components are designed to withstand the effects of cyclic loads due to system te=perature and pressure changes. (1) These cyclic loads are introduced by unit lead transients, reactor trips, and unit heatup and cooldcwn operations. The nu=ber of ther=al and loading cycles used for design purposes are show in Table 4-8 of the FSAR. The maximum unit heatup and cooldown race of 100*F per hour satisfies stress li=its for cyclic operation. (2) The 200 psig pressure limit for the secondary side of the steam generator at a temperature leas than 100*F satisfies stress levels f or temperatures below the DTT. (3) The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximum NDTT value of 20*? has been deter =ined based on Charpy V-Notch tests. The max 1=um NDTT value obtained for the steam generator shell material and welds was 40*F. Figures 3.1.2-1 and 3.1.2-2 contain the li=iting reactor coolant system pressure-temperature relationship for operation at DTT(4} and below to assure that stress levels are low enough to preclude brittle fracture. These stress levels and their bases are defined in Section 4.3.3 of the FSAR. As a result of fast neutren irradiation in the region of the core, there will be an increase in the NDTT with accumulated nuclear operation. The predicted maximum NDTT increase for the '0-year c:gcaura ia chevn en Figure 4.10 (4) The actual shif t in NDTT will be determined periodically during plant opera-tion by testing # irradiated vessel material eacples located in this The results of the irradiated sa:ple testing will be d reactor vessel. evaluated and compared to the design curve (Figure 4-11 of FSAR) being used to predict the increase in transition temperatura. The design glue for fast neutron (Z > 1 MeV) exposure of the reactor vessel f I is 3.0 xy0 n The calculated maximum values 3.0 x 10 are 2.2 x 10 ngm{cm--sat n/c=2--s and 7.2 2,568MWtratedpcverandanintegrated for 40 years operati x 10'gn. (6 operation at 80 percent load. (4) Figure 3.1.2-1 is based on the design value which is considerably higher than the calculated value. The DTT value
'nr Figure 3.1.2-1 is based en the projectad NDTT at the end of the l
first two years of operation. During these two vears, 6 the energy output has been conservatively estimated to be 1.7 x 10 thermal megawatt days, which is equivalent to 655 days at 2,568 MWt core power. The projected
! l
- 18. l fast neutgen exposure of the reactor vessel for the two years is 1.7 x 10 1d n/cm which is based on the 1.7 x 10 thermal megawatt days and the design value for fast neutron exposure.
The actual shif t in NDTT will be established periodically during plant operation by testing vessel material samples which are irradiated cumula-tively by securing them near the inside wall of the vessel in the core area. To compensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown. J 3.1-4 Rev. 19 5/5/72 (Carry over)
i i The NDTT shif t and the magnitudes of the thermal and pressure stresses are N.,
- 18. sensitive to integrated reactor power and not to instantaneous power level.
I Figure 3.1.2-1 and 3.1.2-2 are applicable to reactor core thermal ratings up to 2,568 MW . l , The pressure limit line on Figure 3.1.3-1 has been aclected such that the reactor vessel stress resulting from internal pressure will not exceed 15 percent yield strength considering the following:
- 1. A 25 psi error in measured pressure.
- 2. System pressure is measured in either loop.
- 3. Maximum differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pump combinations.
i For adequate conservatism in fracture toughness including size (thickness) , effect, a =aximum pressure of 550 psig below 275*F with a =ax1=um heatup and cooldown rate of 50*F/hr has been imposed for the initial two year l period as shown on Figure 3.1.2-1. During this two year period, a fractura toughness criterion applicable to Oconee Unit i beyond this period will be
! developed by the AEC. It will be based on the evaluation of the fracture toughness properties of heavy section (thickness) steels, both irradiated 3
and unirradiated, for the AEC-HSST prc; ram and the PVRC program, and with considerations of test results of the Oconee Unit 1 reactor surveillance program. Y The spray temperature difference restriction is imposed to =aintain the thermal stresses at the pressurizer spray line no::le below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell. REFERENCES (1) FSAR Section 4.1.2.4 i j (2) ASME Boiler and Pressure Code, Section III, N-415 (3) FSAR Section 4.3.10.5 1 [ (4) FSAR Section 4.3.3 (5) FSAR Section 4.4.6 I (6) FSAR Sections 4.1.2.8 and 4.3.3 j u 3.1-5 Rev. 13 3/10/72 Rev. 19 5/5/72 (Carry over)
. o POINT TEMP. PRESS. J A 40 550 8 275 550 C 275 1400 0 380 2275 2400 _
2200 _ 2000 - a
. 1800 _
a fo
~
UPPER PRESSURIZATION LIMIT
~
I400 - ;-C E l E 1200 ._. 2 2 1000 _ v a 800
]= _.
2 3 600 - 5 i S 400 - NAXINUM HEATUP RATE,*F/HR 200 -
= 50 - 100 =
l
' I ' I ' I ' I ' I ll 0 i 0 100 200 300 400 500
'l 275 4 Indicated Reactor Coolant System Temperature.'F burceu, OCONEE NUCLEAR STATION Y# 1 1 I REACTOR COOLANT SYSTEW nEATUP LIMiiATIONS 1 (APPLICABLE UP TO AA INTEGRATED EXPOSURE v 0F 1.7 x 10 18 n'cm2 OR DTT - 144 F) F i gu r e 3.1.2 -1 3.1-6
P0thi TEi? PRESS A 250 2275 3 275 1400 C 275 550 0 250 550 E 250 45C 2400 - F 175 453 A 0 "I H 1:0 200 2200 - RC PUMP CCM2 NATIONS ALLCIABLE; ABOVE 135F ALL y 2000 - SELOW ISSF 1 A,1 8;0 A.2 8;1 - A.0-B;0- A,1 -B (1) EHEN DECAY P.!AT REWOVAL SYSTEM (OH) IS CPERATING flTHOUT ANY RC PutiPS CPERATING, { 1800 O INDICATED CH RETURN TEW?. T3 THE REACTOR [ 1600 - VESSEL SHALL BE LSED. l $ () TMRARRE RAME 2 EOF TO H5F. A
~
Mall;4ua STEP TEMPERATURE CHANGE OF 75F ] } 1400
- IS ALLCTA3LE FOLLCIED 3Y A ONE HOUR 3 1200 -
MINitta HOLO CN TEMPERATURE. IF THE STEP j CHANGE IS TAKEN SELC A 250F RC TEMPERATURE,
- 1000 - THE Mall 2U2 ALLC1ASLE STEP SHALL BE THAT
$ 141Ch YlELCS A FINAL TEMPERATURE OF 175F.
3 300 - THE STE? TEh?ERATURE CHANGE IS DEFINE 0 AS n RC TEMPERATURE (BEFORE STOPPING ALL RC PUMPS) H MM R T TH: REA: TOR 600 n / VESSEL. c f
~
C p b 400 -
/
i 200 - UPPER PRESSURI Z ATION .-. Llili O _ i MAllMUM C00LCO#N RATE, 'F/HR (2)- l
- 100 l- ;
50 t
- f 1 i e i t t
$30 i a 6 260 3 175 la 3 275 400 300 200 100 600 500 incicates Reactor Coolant System Te.maerature.*F 5I)
RE ACTOR COOL ANT SY STEM C00L001N LIMI T ATICNS ( APPLICASL E UP TO OTT = 185'F) v Iha P!fm OCONEE NUCLEAR STATION x Figure 31.2 - 2 Rev.19 5/5/72 X i l L _ ___ ._
3.1.3 Mi imum Conditions for Criticality Soccification J' 3.1.3.1 The reactor coolant temperature shall be above 525 F except for portions of low power physics testing when the requirements of specification 3.1.9 shall apply. 3.1.3.2 Reactor coelant te=perature shall be above DTT + 10 F. l 3.1.3.3 When the reactor coolant temperature is below the minimum ! temperature specified in 3.1.3.1 abo've, except for portions of low power physics testing when the requirements of specification 3.1.9 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization. l 3.1.3.4 The reactor shall be maintained subcritical by at least 1% Ak/k until a steam bubble is for=ed and a water level between
'80 and 396 inches is established in the pressurizer.
1 Bases At the beginning of lif e of the initial fuel cycle, the moderator tempera-ture coefficient is expected to be slightly positive at operating tempera-tures with the operating ccnfiguration cf centrol reds. 4-) Calculations 2 show that above 5250F, the consequences are acceptable. < f s./ l Since the moderator temoerature coefficient at icwer temperatures will be l
! less negative or more positive than at operating temperature, (2) start- l up and operation of the reactor when reactor coolant temperature is less 1 than 525 F is prohibited except where necessary for low power physics )
j
- tests.
The potential reactivity insertion due to the moderator pressure coefficient ( ) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1% ak/k. l 1 In addition, During physics tests, special operating (nrecautions will be taken.1) and the small integrated ak the strong negative Doppler coefficient limit the magnitude of a power excursion resulting f rom a reduction of moderator density.
' The requirement that the reactor is not to be made critical below DTT + 10F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NDTT of the primary coolant system. Heatup to this te=perature will be acce=plished by operating the reactor coolant pumps.
If the shutdown margin required by Specification 3.5.2 is maintained, there is no possibility of an accidental criticality as a result of a decrease of coolant. pressure. J. The requirenent for pressurizer bubble formation and specified water level when the reactor is less than 1% subcritical will assure that the reactor 3.1-8 l
- _ - . ..n,,, - _ y ,,-. - ,-
$ 0 coolant system cannot become solid in the event of a rod withdrawal accident y or a start-up accident.(3) REFERENCES (1) FSAR, Section 3 (2) FSAR, Section 3.2.2.1.4 (3) FSAR, Supplement 3, Answer 14.4.1 Y v 3.1-9 ,
~m
- O 4
3.1.4 Reactor Coolant System Activiry J Specification The total activity of the reactor coolant due to nuclides with half lives longer than 30 minutes shall not exceed 224/E microcuries per m1 whenever the reactor is critical. E is the average (mean) beta and gan=a energies per disintegration, in MeV, weighted in proportion to the measured activity of the radionuclides in reactor coolant sa=ples. Bases The .ihove specification is based on limiting the consequences of a postulated accident involving the double-ended rupture of a steam generator tube. The rupture cf a steam generator tube enables reactor coolant and its associated activity to enter the secondary system where volatila isotopes could be discharged , to the atmosphere through condenser air-ejectors and through steam safety valves (which may lif t momentarily). Since the =ajor portion of the activity entering the secondary system is due to noble gases, the bulk of the activity would be discharged to che atmosphere. The activity release continues until the operator stops the leakage by reducing the reacter coolant system pressure below the set point of the steam safety valves and isolates the faulty steam generator. The operator can identify the faulty steam generator by using the 46N detectors on the steam lines in conjunction with the off-gas =cnitors on the condenser air ejector lines; thus he can isolate the faulty steam generator within 34 minutes gf ter the tube break occurred. During that 34 =inute pericd, a maximum of 2760 ft of 560 F reactor coolant leaked in~o the secondary system. (This is equivalent to a cold makeup volume of 1980 ft ).5 s /' The activity discharged to the atmosphere as the result of a steam generator ( tube rupture will not be increased by the loss of station power since condenser cooling water flow can be maintained by gravity flow from Lake Keovee through t he er.ergency condenser cooling water discharge to the Keowee Hydro tailrace. lhe controlling dose for the steam generator tube rupture accident is the whole-body dose resulting from immersion in the cloud of released activity. To insure that the public is adequately protected, the specific activity of- the reactor coolant will be limited to a value which will insure that the whole-body dose at the site boundary will not exceed 0.5 Rem should a steam generator tube rupture accident occur. Although only volatile isotopes will be released from the secondary system, the following whole-body dose calculation conservatively assu=es that all of the l radioactivity which enters the secondary system with the reactor coolant is released to the atmosphere. Both the beta and gamma radiation from these j isotopes contribute to the whole-body dose. The gamma dose is dependent on the '
! finite size and configuration of the cloud. However, the analysis employs the
+ simple model of a semi-infinite cloud, which gives an upper limit to the ] potential gamma dese. The semi-infinite cloud =odel is applicable to the beta i dose because of the short range of beta radiation in air. It is further assumed that meteorological conditions during the course of the accident correspond j to Pasquill T / l of 1.16 x 10-{pesec/m F and3, 1 which meter per second includes wind speed, a correcticn resulting factar of 2.2 in toa the X Qdilution va ue s ,,/ j 4 3.1-10
l t calculated by the Pasquill method. This correction factor was shcwn appro-
.,,, priate by on alte diffusion measure =ents using S?6 (aulfur hexafluoride) as a gas tracer.
The combined gamma and beta whole body dose from a se=1-infinite cloud is given by: 10 dps/C1) - (1.33 x 10-11 Re=/MeV/m3 )] Dose (Rem) = h(5*A*V X/Q-(3.7 x 10 Dose (Rem) = 0.246 Y A V X/Q Amax(uc/cc) = (Dose) ,3x 0.5 l , 0.246 x I x 7d. 25 x 1.16 x 10-* O.246 I*V X/Q l A,,x(uc/cc) = 224/5 Where A = Reactor coolant activity (uCi/ml = C1/m 3) l V = Reactor coolant volume at 580*F leaked into secondary system (2763 ft3= i 78.25 m3) i X/Q = Atmospheric dispersion coefficient at site boundary for a two hour j period (1.16 x 10-4 sec/=3) E = Average beta and gacma energies per disintegration (MeV)
- s--
Calculations required to determine I will censist of the following: 1 4
- 1. Quantitative measurecent of the specific activity (in units of 'gCi/cc) of radionuclides with half lives longer than 30 minutes, which =ake up at
^ j least 95% of the total activity in reactor coolant samples.
- 2. A de.'ermination of the average beta and ga==a decay energies per disinte-l gration for each nuclide, measured in (1) above, by utilizing kncwn decay j energies and decay sche =es (e.g., Table of Isotopes, Sixth Edition, March 4
1968). 1 __
- 3. A calculation of E by the average beta and ga=ma energy for each radionuclide in proportion to its specific activity, as =easured in (1) above.
1
- REFERENCES i l FSAR, Section 14.1.2.10 I i
i i i
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- - - - , , , -m<.m, .r-, +,.
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! 3.1.5 Chemistry V Specificaticn 3.1.5.1 If the concentration of oxygen in the primary ecolant exceeds 0.1 ppm during power cperation, corrective action shall be initiated within eight hours to return oxygan levels to ;[ 0.1 ppm. 3.1.5.2 If the concentration of chloride in the pri=ary coolant exceeds 0.10 ppm during power operation, corrective acticn shall be 19.l initiated within eight hours to return chloride levels to 19.l ;L O.10 ppm. 3.1.5.3 If the concentratica of fluorides in the pri=ary coolant exceeds 0.10 ppe follcwing modifications er rapair to the pri=ary system involving welding, corrective actica shall be initiated within eight hours to return fluoride levels to ;i 0.10 ppm. 3.1.5.4 If the concentration limits of oxygen , chloride or fluoride in 3.1.5.1, 3.1.5.2 and 3.1.5.3 abova are not restored within 24 hours the reactor shall be placed in a hot shutdown condition within 12 hours thereafter. If the normal operational limits are not rastored within an additional 24-hour period, the reactor shall be placed in a cold shutdcwn condition within 24-hours thereafter. 3.1.5.5 If the oxygen concentration and the chloride or fluoride concentra-tion of the pri=ary coolant system individually exceed 1.0 ppm. the reactor shall be i==ediately brought to the hot shutdown condition using normal shutdevn procedure and action is to be taken d--adiately to return the system to within nor=al operatic specifications. If l normal operating specifications have not been reached in 12 hours, the reactor shall be brought to a cold shutdown condition using nor=al procedure. Bases By maintaining the chloride, fluoride and exysen concentration in the reactor coolant within the specifications, the integrity of the reactor coolant system is protected against potential stress corrosien attack. (1,2) The oxygen concentration in the reactor coolant system is nor: ally expected to be below detectable limits since dissolved hydrogen is used when the reactor is critical and a residual of hydtamine is used when the reactor is subcritical to control the oxygen. The requirement that the oxygen cen-centration not exceed 0.1 ppm during pow 97, operation is added assurance that
\"'
stress corrosion cracks will not occur. If the oxygen, chloride, or fluoride limits are exceeded, measures can be taken to correct the condition (e.g., switch to the spare demineralizer, l replace the ion exchange resin, increase the hydrogen concentraticn in the ' makeup tank, etc.) and further because of the time dependent nature of any adverse effects arising from chlorides or oxygen concentrations in %/ i excess of the limits, it is unnecessary to shutdcen i==ediately, since the condition can be corrected. 3.1-12 Rev. 19. 5/5/72
, _ _ _ - _ . _ _ . ~ - - -
The oxygen and halogen li=its opecified are at least an order of magnitude , is, below concentrations vnich could result in damage to materials found in the reactor coolant system even if =aintained for an extended peried of time. (4) Thus, tne period of eight, hours to initiate corrective action and the period of 24 hours to perform corrective action to restore the concentration within the limits have been established. The eight hour period to initiate corrective action allows time to ascertain that the chemical analyses are correct and to locate the source of contamination. If corrective action , has not been effective at the end of 24 hours, then the reactor coolant system will be brought to the hot shutdown condition within 12 hours and corrective action will continue. If the operational limits are not restored
- 19. within en additional 24 hour period, the reactor shall be placed in a cold shutdown condition within 24 hours thereafter.
The maximum limit of 1 ppm for the oxygen and halogen concentration that will not be exceeded was selected as the hot chutdown limit because these values have been shown to be safe at 500*?. (3) 1 REFERENCES (1) FSAR, Section 4.1.2.7 l (2) FSAR, Section 9.2.2 (3) Stress Corrosion of Metals, Logan N-- (4) Corrosion and Wear Handbook, O. J. OeFaul, Editor 1 l l t l
%W 3.1-13 Rev. 19. 5/5/72 l
- O 3.1.6 Leakage Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be shutdown within 24 hours of detection.
3.1.6.2 If unidentified reactor coolant leakage (excluding normal evaporative losses) exceeds 1 sps or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be shutdown within 24 hours of detection. 3.1.6.3 If any reactor coolant leakage exists through a non-isolable fault l in a RCS strength boundary (such as the reactor vessel, piping, valve body, etc.), the reactor shall be shutdown and cooldown to the cold shutdown condition shall be initiated within 24 hours of detection. 3.1.6.4 If reactor shutdown is required by Specification 3.1.6.1, 3.1.6.2 or 3.1.6.3, the rate of shutdown and the conditions of shutdown shall Se determined by the safety evaluation for each case and justified in writing as soon thereafter as practicable. t 3.1.6.5 Action to evaluate the safety implication of reactor coolant leakage shall be initiated within 4 hours of detection. The nature, as well l j as the magnitude, of the leak shall be considered in this evaluation. The safety evaluation shall assure that the exposure of offsite personnel to radiation is within the guidelinea of 10C7120. i i 1.1.6.6 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2 or s,/ i 3.1.6.3 the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected. 3.1.6.7 When the reactor is critical and above 2% power, two reactor coolant a leak detection systems of dif ferent operating principles shall be in operation, with one of the two systems sensitive to radioactivity. The j systems sensitive to radioactivity may be out-of-service for 48 hours provided two other means are available to detect leakage. 3.1.6.8 Indicated leakage of reactor coolant shall be considered actual leakage unless (1) the indicated leak cannot be substantiated by direct observation or other indication or (2) a safety problem does not exist. Loss of reactor coolant through reactor coolant pump seals and system : valves to connecting systems which vent to the gas vent header and from which coolant can be returned to the reactor coolant system shall
.ot be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1.6.1 - 3.1.6.6 except that i l such losses when added to leakage shall not exceed 30 gpm.
]
l Bases ! i Every reasonable ef fort will be made to reduce reactor coolant leakage including evaporative losses (which may be on the order of .5 gpm) to the lowest possible rate and at least below 1 gpm in order to prevent a large leak from masking the presence of a smaller leak. '4ater inventory balances radiation monitoring equip- s s/
; ment, boric acid crystalline deposits, and physical inspections can disclose 3.1-14 I
w- .- a ,--.y w------ - ,--- - g - - , y 'yet
' reactor coolant leaks. Any leak of radioactive fluid, whether from the reactor coolant system primary boundary or not can be a serious problem with respect g,, to in-plant radioactivity contamination and cleanup or it could develop into
- a still more serious prcblem; and therefore, first indications of such leakage 4
will be followed up as soon as practicable. {i Although some leak rates on the order of GPM may be tolerable from a dose i point of view, especially if they are to closed systems, it must be recog-i nized that leaks in the order of drops per minute through any of the walls of the primary system could be indicative of =aterials failure such as by stress corrosion cracking. If depressurization, isolation and/or other safety measures are not taken promptly, these small breaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, the nature of the leak, as well as the magnitude of the leakage must *oe considered in the safety evaluation. When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will he performed by the Operating Staff and will be documented in writint and approved by the Superintendent. Under these conditions, en allowable reactor coolant system leakage rate of 10 gpm has been established. This j explained leakage rate of 10 gpa is also well within the capacity of one i high pressure injection pump and makeup would be available even under the loss of off-site power condition. i j If leakage is to the reactor building it may be identified by one or = ore of s,, the following methods :
- a. The reactor building air particulate monitor la sensitive to low leak races. The rates of reactor coolant leakage to which the instrument is sensitive are .10 gpm to greater than 30 gpm, assuming corrosion 4
product activity and no fuel cladding leakage. Under these conditions, an increase in coolant leakage of 1 gpm is detectable within 10 i I minutes af ter it occurs.
- b. The iodine monitor, gaseous cogitor and area monitor are not as sensitive I
to corrosion product activity.ll) It is calculated that the iodine l j monitor is sensitive to an 8 gpm leak and the gaseous monitor is sen- { j sitive to a 230 gpm leak based on the presence of tramp uranium (no fission products from tramp uranium are assumed to be present). How-
; ever, any fission products in the coolant will make these monitors l more sensitive to coolant leakage.
1
! c.
In addition to the radiation monitors, leakage is also monitored by a level indicator in the reactor building normal sump. Changes in normal sump level may be indicative of leakage fro: any of the systems located inside the reactor building such as reactor coolant system, low pressure service water system, component cooling systen and steam and feedwater lines or condensation of humidity within the reactor building atmosphere. The sump capacity is 15 gallons per inch of height and each graduation l )
on the level indicates h inch of sump height. This indicator is capable l
} of detecting changes on the order of 7.5 gallons of leakage into the , suma. . A 1 gpm leak would therefore be detectable within less than
-10 minutes.
1 3.1-15 i
a i t j- d. Total reactor coolant system leakage rate is periodically determined by comparing indications of reactor power, coolant temperature, pressurizer ,, / water level and letdown storage tank level over a time interval. All ! of these indications are recorded. Since the pressurizer level is i maintained essentially constant by the pressurizer level controller, any coolant leakage is replaced by coolant from the letdevn storage i tank resulting in a tank level decrease. The letdown storage tank capacity is 31 gallons per inch of height and each graduation on the level recorder represents 1 inch of tank height. This inventory 3 monitoring method is capable of detecting changes on the order of 31 gallons. A 1 gpn leak would therefore be detectable within , approximately one half hour. As described above, in addition to direct observation, the means of detecting reactor coolant leakage are based on 2 different principles, i.e. , activity, sump level and reactor conscant inventory measurements. Two systems of different principles provide, therefore, diversified ways of detecting leakage to the reactor building. The upper limit of 30 gpm is based on the contingency of a complete loss of station power. A 30 gpm loss of water in conjunction with a complete loss of station power and subsequent cooldown of the reactor coolant system by the turbine byp;ss system (set at 1,040 psia) and steam driven emergency i feedwater pump would require more than 60 minutes :: em:ty the pressuricer
' from the cembined effect of system leakage and contraction. This will be ample time to restore electrical power to the station and makeup flow to the reactor coolant system.
REFERENCES 1 FSAR Section 11.1.2.4.1
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3.1.7 Moderator Temperature coefficient of Reactivity
\-- Specification The maximum moderator temperature coefficient at full power shall not 1
e, exceed +0.9 x 10-4 ok/k/0F. Bases The accident analysis described in the FSAR (1) was done for a range of moderator temperature coefficient values including +0.9 x 10-4 Ak/k/0F. The maximum positive value of the moderator coefficient used in the calcu-latio x 10 g of clad ak/k/ F.temperature (2) for the loss-of-coolant accident was +0.9 The threshold value of the positive moderator coefficient that allows azimuthal xenon oscillations is greater than +0.9 x 10-4 ok/k/ F. When the hot zero pcwer value is corrected to obtain the hot full power value, the following corrections will be applied. I A. Uncertainty in isothermal =casurement
- ine measured moderator te=peratura coefficient will contain uncertainty on the account of the following
l 1. ;0.2 F in the AT of the base and perturbed conditions. s ,, 2. Uncertainty in the reactivity measurement of +0.1 x 10-Ak/k. Proper corrections will be added for the above conditions to result in a conservative moderator coefficient. I B. Doppler coefficient at hot zero pcwer j During the isothermal moderator coefficient measurement at hot j zero power, the fuel te=perature will increase by the same amount
- as the moderator. The measured temperature . coefficient must be increased by 0.16 x 10-4(ak/k)/oF to obtain a pure moderator te=perature coefficient.
C-3.1-17 _.w- ,-- w.-_ _ . y, , , , _ , _ . _ , . . , , ,-m %-- _ _ , . p , _ _ - . _ y - _
C. Moderator te=perature change The hot zero power measurement must be reduced by .09 x 10
-4 s.s/
(ak/k)/ F. This corrects for the difference in water temperature at zero power (532 F) and 15% power (580 0 ?) and for the increased fuel temperature effects at 15% pcwer. Above this power, the average moderator temperature re=ains 5 8007. However, the co-efficient, 4 m, must also be adjus ted for the interaction of an average moderator temperature with increased fuel te=peratures. This correction is .001 x 10-4 Ac m/A% pcwer. It adjusts the 15% power an to the moderator coefficient at any power level above 15% i power. For . exam ( .001 x 10-4) (ple, 85%)to, which correct is to.085 100% power, x 10-4L a . c, is adjusted by m 1 D. Dissolved boron concentration This correction is for any difference in boron concentration, if required, between zero and full pcwer. Since the moderator coefficient is = ore positive for greater dissolved boron concen-trations, the sign of the correction depends on,whether boron is added or removed. The correction is 0.lo x 10-0 Aa./APPM. (The magnitude of the correction varies slightly with bo?cn concen-tration; the value presented above, however, la valid for a range in boron concentrations from 1000 to 1400 ppm.) E. Centrol rod insertien This correction is for the difference in control rod worth (% s.,/ j Ak/k) in the core between zero and full power. The correction is 0.17 x 10-4 ac=/%Ak/1, where the sign for rod worth change is ] negative for rod insertion, because the moderator coefficient is q { = ore negative for a larger rod worth in the core. j J F. Isother:al to distributed ca=perature J j The correction for spatially distributed moderator te=perature has been found to be less than or equal to zero. Therefore , zero is j a conservative correction value for distributed effects. i G. Azimuthal xenen stability ] Before commercial operation a test will be performed to verify that divergent azi=uthal xenon oscillations do not occur. i REFERENCES 1' (1) FS AR, Sectica 14 i (2) FS AR, Section '3 ( 3) FS AR, Section 14.2.2.3.4 N./ 3.1-18 4
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4 9 3.1.8 Single Leon Restrictiona
\,,, Specification i The following special limitations are placed ch single loop operation in addition to the limitations set forth in Specification 2.3.
! 3.1.8.1 Single loop operation is authorized for test purposes only. ! 3.1.8.2 At least 23 incore detectors =ceting the requirements of Technical
- 19. Specification 3.5.4.1 and 3.5.4.2 shall be available thro'2ghout this test to check gross core power di tribution.
j 3.1.8.3 The pump monitor trip set point hhall be set at no greater than
- 50% of rated power.
1 3.1.8.4 The outlet reactor coolant te=perature trip set point shall be j set at no greater than 61007. 3.1.8.5 At 15% of rated power and every 10" of rated power above 15%, measurements shall be taken of each operable incore neutron detector and each operable incore thermocouple, reactor coolant loop flow rates and vessel inlet and outlet temperatures, and evaluation of this data determined to be acceptable before proceeding to higher power levels. 3.1.8.6 DRL shall be notified of the scheduled date of single loop testing. Upon completion of test, results shall be reported to DRL. l Subsequent single loop operation shall be contingent upon g,, written approval by LRL. Bases The purpose of single loop testing ia to (1) supplement the 1/6 scale =cdel test information, (2) verify predicted flov through the idle loop, (3) verify changes in power level do not affect flow distribution or core power dis-tribution, and (4) demonstrate that limiting safety system settings (pump monitor trip set point and reactor coolant outlet te=perature trip set
] point) can be conservatively adjusted taking into account instrument errors.
Limiting the pump monitor trip set point to 50% of rated power and the i reactor coolant outlet temperature trip set point to 6100 F to perform this confirmatory testing assures operation well within the core protective safety limits shown in Figure 2.1-3, curve 2. Incore thermocouples will be installed and dat:a will be taken to check outlet core temperature profiles. This data vill be used in evaluating test results. 1 1 1 v 3.1-19 Rev. 19. 5/5/72 J n -
i e 1 1 1 I 3.1.9 Low Power Physics Testinz Restrictions Specification '*d ! The following special limitations are placed on low power physics testing. 3.1.9.1 Reacto. Protective System Requirements
- a. Below 1720 psig Shutdcwn Bypass trip setting limits shall apply i
in accordance with Tcble 2.3-1.
- b. Above 1800 psig nu.-lear overpower trip shall be set at less than 5.0%. Other settings shall be in accordance with Table 2.3-1.
i 3.1.9.2 Startup rate red withdrawal hold shall be in effect at all times. l 3ases ] 19, Technical Specification 3.1.9.2 will apply to both the source and intermediate ranges. The above specification provides additional safety margins during low power physics testing. 4
%/
I i I f 4 l I i .i J Rev. 19, 5/5/72 3.1-20
3.1.10 Centrol Red Oper.2 tion v Specification 3.1.10.1 Allowable co=binations of pressure and temperature for control roi operation shall be to the lef t of and above the limiting pressure versus temperature curve as shown in Figure 3.1.10-1. 3.1 10.2 The dissolved gas concentration shall not exceed 100 standard ec/ liter. 3.1 10.3 If eithcr the limits of 3.1.10.1 or 3.1.10.2 are exceeded, the center control rod drive mechaniss shall be checked for accumulation of un-dissolved gases. Base, The l'.miting pressure versus temperature curve for dissolved gases is determined by th equilibrium pressure versus ce=perature curve for the dissolved gas con-centr,-ion of 100 std. cc/ liter of water. This equilibrium total pressure is the sun of the partial pressure of the dissolved gases plus the partial pressure of water at a given tenperature. 3y maintaining the reactor coolan. tamperatura and pressura aa specified above, any dissolved gases in the reactor coolant system are maintained in solution. Although the dissolved gas concentrat.on is expected to be approximately 20-40 std. cc/ liter of water, the dissolved gas concentratien is conservatively assumed to be 100 std. cc/ liter of water at the reactor vessel outlet temperature. If either the maximum dissolved gas concentration (100 std. cc/ liter of water) is exceeded or the operating pressure falls below or to the right of the limiting pressure versus temperature curve, the center CF.0M should be checked for accunulation of undissolved gases. l l l l 1
\--
' ~
Rev. 13 3/10/72 (New Page) l l 1
2r,00 . _ . _ _ _ . _ . . ..
- i i
1 1800 J 1600 7 l J. t i , E 1400 a
, J ' E i
0 1200 4 E 1 5 O E 1000 I E a o
- 800
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S / M i , E ' t
= i
[
! 3* 600 2
l J E 1 400 , 1 1 i 200
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~ ~M
1 0 0 100 200 300 400 500 600 700
< Incicatea Reactor Coolant System T:.:;erature,*F I LIMITING PRESSURE VS TEMPERATURE CURVE FOR 100 STO CC/LtTER H 2O l l<
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!? CCONEE NUCLEAR STATION i Figure 3.1.10 - 1 Rev.18 3/10/72 i (New Page) 3 Page 3.1 - 22 i
l
3.2 HIGH PRESSURE INJECTION AND CEEMICAL ADDITION SYSTEMS
._ Applicability Applies to the high pressure injection and the che=ical addition systems.
j Obiective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdcwn condition. 4 Specification 1 The reactor shall not be critical unless the following ccnditions are met: 3.2.1 Two high pressure injection pumps are operable except as specified in 3.3. j 3.2.2 A source of concentrated soluF 'e boric acid in addition to the borated water storage tank, f tvailabic and operable. This can be either:
- a. The bo ic acid mix tank containing at least the equivalent of 450 ft of 10,600 pp: boron as boric acid solution at a te=perature I, of at least 10*F above the crystal 11:stion tecperature. System i
! piping and valves necessary to establish a ficw path from the j g,, tank to the high pressure injection system shall also be operable and shall have at least the sa e temperature requirement as the boric acid mix tank. One ass,ociated boric acid pu=p shall be operable.
If the daily average air temperature in the vicinity of this tank and associated flow path piping is less than 85 7,0 at least one channel of heat tracing shall be in operation for this tank and 71 ping.
- b. The concentrated boric acid storage tank centaining at least the equivalent of 550 ft3 of 8700 pp baron as boric acid solution with a temperature of at laast 10*? above the crystallization t e=pe rature. System piping and valves necessary to establish a flow path from the tank to the high pressure injection system shall be operable and shall have the same temperature requirement as the concentrated boric acid storage tank. One associated boric acid pump shall be operable. If the daily averace air temperature in the vicinity of this tank is less than 70 F, at least one channel of heat tracing shall be in operation for this tank and associated piping.
Bases The high pressure injection system and che=ical addition system provide control of the reactor coolant . system boron concentration. (1) This is nor: ally accomplished by using any of the three high pressure injection pu=ps in A.- series witn a boric acid pu=p associated with either the boric acid mix cank or the ' concentrated boric acid storage tank. An alternate method of boration will be use of the high pressure injection pumps taking suction directly f rom the borated water storage tank.(2) 3.2-1
The quantity of boric acid in storage frc= any of the 3 abcve centioned sources is sufficient to borate the reactor coolant system to a 1% suberitical margin in the cold condition at the end of cora life. The =aximum required is the equivalent of 396 ft3 of 10,600 pps boron as boric acid solution. A mininum of 450 ft3 of 10,600 ppm b'oron as boric acid solution ta the boric acid mix tank, a minimum of 550 f t3 of 8,700 ppm boren as boric acid solution in the concentrated boric acid storage tank or a mini =u= of 350,000 gallons of 1800 ppm boron as boric acid solution in the barated water storage tank (3) 3 will satisfy the requirements. The specification assures that at least two 4 of these supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The required amount of boric acid can be added in several ways. ~he quickes: method allows for the necessary boron addition in less than one hour. The slowest method (using one 10 gpm pu=p taking suction from the boric acid storage tank) would require approxi=ately 3 hours to inject enough boron to keep the reactor 1% suberitical with xenon in the core. As xenon deca /s out, more boron would have to be added. Therefore, in order to account for xenon decay, the 10 gpm pump wculd pump for something less than 5 hours.
, At this time, the reactor coolant system wculd be at a temperature of approximately 175'F and the core would be more than 1% suberitical.
The concentration of boron in the boric acid mix tank and concentrated boric acid storage tank may be higher than the cencentration which would crystal 1L:e at ambient conditions. For this reason and to assure a ficw , vi boric acid is available when needed these tanis and their associated ! piping will be kept 10*F above the crystalli:stion temperature for the concentration present. Once in the high pressura injectien system, the i 4 concentrate is sufficiently well mixed and diluted so that nor=al system d temperatures assure horic acid solubility. 1 ] The horic acid mix tank concentration of 10,600 ppm baron corresponds to a precipitation temperature of 80 F, and the concentrated boric acid storage tank concentration of 8700 ppm corresponds to a precipitation temoerature of od"F. It is expected that the surface temperatures of these tanks and 0 associated piping will be 10 F above the precipitation tamperatures. If the air temperature should approach a precipitation temperature, at least one channel of heat tracing in service assures that heat losses to the atmosphere wi11 be made up to maintain this 10 F =argin. REFERENCES l ( L) FSAR, Section 9.1; 9.2 (2) FSAR Figure 6.2 l (3) Technical Specification 3.3 1 e 1 1 1 3.2-2 4 i
/
3.3 EMERGENCY CORE CCOLING, REACTOR BUILDING CCOLING, REACTOR 3UILDING SPRAY AND PENETRATION ROOM VENTILATION SYSTEMS Y Applicability Applies to the e=crgency core cooling, reactor building cooling, reactor t . c,., , building spray and reactor building penetration room ventilation syste=s. Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building cooling, reactor building spray and reacter building penetration room ventilatica systems. Specification 3.3.1 The reactor shall not be nada critical, unless the folicwing conditions are met:
- a. Injection System (1) The borated water storage tank shall contain a minimum of 350,000 gallons of water having a minimum concentration of 1,500 pp barca ar. a :e:perature no less than LC'?. The manual valve, LP-28, on the discnarge line from the borated water storage tank shall be locked open.
b (2) Two out of three high pressure injection pumps shall be operable. (3) Two Engineered Safety feature icw pressure injection pumps shall be eparable. (4) Two low pressure injection coolers shall be operable. (5) Two 3'4ST lavel instrument channels shall be operable. (6) The two reactor building emergency sump isolatica valves shall be either manually or remote-manually operable.
- b. Core Flooding System (1) The two core flooding tanks shall each contain a minimun
- 18. of1040130 f t 3 of borated water at 600 1 25 psig. l (2) Core flooding tank boren concentration shall not be less ~
than 1,800 ppa baron. (3) The electrically operated discharge valves from the core flood tank shall be open and breakers locked open and tagged. v (4) Two core flood tank pressure instrument channels shall be operable. 3.3-1 Rev. 16 3/10/72 L. _
- c. Rdactor Building Spray System and Reacter Euilding Cooling System The following subsystems shall be operable:
J (1) Two reactor building spray pumps and their associated spray nozzle headers. (2) Three reactor building cooling fans and associated cooling units.
- d. Low Pressure Service Water System (1) Two low pressure service water pu=ps shall be operable.
(2) The valve in the LPSW discharge from the reactor building cooler (LPSW 108) shall be locked open.
- e. Reactor Building Penetration Room Ventilation System Both penetration room fans and their associated filters shall be '
operable . Manual operated system valves (PR-12, PR-14, PR-16, and h PR-18) shall be locked open.
- f. Engineered Safety feature valves and interlocks associated with i each of the above systems shall be operable. ..
3.3.2 Except as noted in 3.3.3 below, maintenance shall be alicwed during -! power operation on any component (s) in the high pressure injection,
! core flooding, low pressure injection, low pressure service water, reactor d 18, building spray.or penetration roc = ventilation systems which will not remove more than one train of each system from servica. Components shall not be removed from service so that the affected system train is inoperable h for more than 24 consecutive hours. If the system is not restored to N meet the requirements of Specification 3.3.1 within 24 hours , the reactor shall be placed in a hot shutdown condition within 12 hours. (
lf the requirements of Specification 3.3.1 are not met within an additional 48 hours, the reactor shall be placed in a coAd shutdown j i-condition within 24 hours. 3 3.3.3 Exceptions to 3.3.2 shall be as follcws:
- a. Both core flooding tanks shall be operational at all times.
k[ Both motor operated valves associated with the core flooding tanks- J b. shall be fully open at all times. j / 1
- c. One reactor building cooling fan and associated cooling pnft shal. Sm 0 y ?
be permitcod to be out-of-service for seven days provided7only one i reactor building spray pump and associated spray nozzle header is_orit of service at the same time per Specificatien 3.3.2. j
- 18. 3.3.4 Prior to initiating maintenance on any of the components, the duplicate (redundant) component shall he tested to assure operability.
d Rev.~ 18 3/10/72 3.3-2 j
i Bases (,, The requirements of Specification 3.3.1 assure that, before the reactor can be made critical, adequate engineered safety features are operable. Two high pressure injection pumps and two low pressure injection pumps are specified. Ecwever, only one of each is necessary to supply e=ergency coolant to the reactor in , the event of a loss-of-coolant accident. Both core flooding tanks are required j as a single core flood tank has insufficient inventory to reflood the core. (1)
, The borated water storage tanks are used for two purposes:
i A. As a supply of borated water for accident conditiens. B. As a supply of borated water for flooding the fuel transfer canal during refueling operation.(2) 350,000 gallons of borated water are required to supply e=ergency core cooling ' and reactor building spray in the event of a loss-of-core cooling accident. This amount fulfills requirements for emergency core cooling. The borated water storage tank capacity of 388,000 gallons is based on refueling volume requirements. Heaters maintain the borated water supply at a temperature to preven-t freezing. The boron concentration is set at the amount of boron required to maintain the core 1 percent suberitical at 70*F without any 5 control rods in the core. This concentration is 1,338 ppm boron while the minimum value specified in the tanka is 1,3C0 ppm boron. The post accident reactor building cooling may be accomplished by three cool-i. Inz units, by two spray units or by a combination of two cooling units and one spray unit. The specified requirements assure that the required post accident components are available.(3) The spray system utilizes common suction lines with the low pressure injection system. If a single train of equipment is removed frca either system, the other train must be assur . to be operable in each system. When the reactor is critical, =aintenance is allcwed per Specification 3.3.2 1 and 3. 3. 3 provided require =ents in Specification 3.3.4 are met which assure operability of the duplicate components. Operability o the specified com-ponents shall be based on the results of testing as required by Technical Specification 4.5. The maintenance period of up to 24 hours is acceptable if the operability of equipment redundan t to that removed from service is demon-j s trated i==ediately prior to removal. The basis of acceptability is a lov likelihood of failure within 24 hours following such demonstration. J In the event that the need for emergency core cooling shculd occur, functioning
] of cne train (one high pressure injection pu=p, one low pressure injection pump, and both core flooding tanks) will protect the core and in the event of a main coolant loop serverence, limit ene peak clad te=perature to less than 2,300*F and the metal-water reaction to that representing less than 1 percent of the clad.
t s-l 3.3-3 1
m l Three low pressure service water pumps serve Oconee Units 1 and 2 and two low pressure service water pu=ps serve Oconee Unit 3. There is a =anual cross- y connection on the supply headers for Units 1, 2 and 3. One low pressure service water punp per unit is required for normal operation. The normal operating requirenents are greater than the escrgency requirements following a loss-of-coolant accident. A single train of reactor building penetration roos ventilation equipment retains full capacity to control and minimize the release of radioactive materials from the reactor building to the environment in post-accident < conditions. 4 REFERENCES 1 l (1) FSAR, Section 14.2.2.3 (2) FSAR, Section 9.5.2 (3) FSAR, Section 14.2.2.3.5 (4) FSAR, Section 6.4 i s.o/ i 4 4 i i } i ' 3.3-4 i
3.4 STEAM & FGWER CONVERSION SYSTDi L Anplicability Applies to the turbine cycle componants for receval of reactor decay heat. Obiective To specify minimum conditions of the turbine cycle equipment necessary to assure the capability to remove decay heat from the reactor core. Specification The reactor shall not be heated above 250*? unless the following conditions are met: 3.4.1 Capability to re=ove a decay heat load of 5 percent full reactor power from at least one of the following means:
- a. A hotwell pumo, condensate booster pump, and a main feedwater pump.
- b. The emergency facdwatar pump.
- c. A hotwell punp and a condensate booster pump.
3.4.2 The sixteen steam system safety valves are operable.
\-%
3.4.3 The turbine bypass system shall have four valves operable. 3.4.4 A minimum of 72,000 gallons of water per operating unit shall be available in the upper surge tank, condensate storage tank, and I hotwell. 3.4.5 The emergency condenser circulating water system shall be operable as per Specification 4.1. Bases The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 250*F. Feedwater makeup is supplied by operation of a hotwell pump, condensate booster pump and a =ain feedwater pump. The feedwater flow required to remove decay heat corresponding to 5 percent full power with sa. erated steam in the pressure range from 30 paia (saturation pressure at 250F) to 1065 psia (lowest setting of steam safety valve) as a function of feedwater temperature is: L 3.4-1 Rev. 18 (Entire Bage Revised) iN
*? Flev, C7M 60 750 g 90 770 120 790 180 840 One hotwell pump plus one condensate booster pump will supply at least 3000 GPM at 550 psia, and one hotwell pump plus one booster pump plus one main feed pump will supply at least 3000 gym at 1065 psia. The emergency feed pump will supply 10S0 gpm at 1065 psia.
t In tha e'ent of ccmplete loss of electrical power, feedwater is supplied by a turbine driven c=ergency feedwater pump which takes suction from the upper surge tanks and hotwell. Decay heat is removed frem steam generator by steam relief through the turbine bypass system to the condenser. Condenser cooling water flow is provided by a siphon effect from Lake Keovee through the condenser for final heat rejection to the Keovee Hydro ?lant tailrace. The minimum amount of water in the upper surge tank and condensate storage tank is the a=ount needed for 11 hours of operaring per unit. This is based on the conservative estimata of normal makeup being 0.57. of throttia flow. Throttle flow at full load, 11,200,000 lbs/hr., was used to calculate the operation time. For decay heat removal the operation time with the volume of water specified would be considerably increased due to the reduced throttle
; flow.
The relief capacity of the sixteen steam system safety valves is 13,105,000 l
- 19. l %/
lbs/hr. The capacity of the four turbine bypass valves is 2,817,000 lbs/hr. REFERENCE FSAR, Section 10 J 3.4-2 Rev. 18 3/10/72 (Entire Pace Retiead) ' Rev.19 5'/5 / 72'
3.5 INSTRUMINTATION SYSTDiS L 3.5.1 Ocerational safety Instrumentatien Applicability _ Applies to unit instrumentation and control systems. Obiective To delineate the conditions of the unit instrumentatica and safety circuits necessary to assure reactor safety. Soecifications 19* 3.5.1.1 The reactor shall not be in a startup : ode or in a critical state unless the require =ents of Table 3.5.1-1, Colu=ns A and 3 are met. 19* 3.5.1.2 In the event that the number of protective channels operable f alls below the limit given under Table 3.5.1-1, Columns A and 3; operation shall be limited as specified in Colu=n C. 3.5.1.3 For on-line testing or in the event of a protective instrument or channel failure, a key-operated channel bypass switch associated with 19, each reacect protactive channel say be used to 1:ck t'..e channel trip relay in the untripped state. Status of the untripped state shall be indicated by a light. Only one channel bypass key shall be
\-- accessible for use in the control room. Only one channel shall be locked in this untripped state or contain a du=sy bistable at any one time.
3.5.1.4 The key-operated shutdown bypass switch associated with each reactor protective channel shall not be used during reactor power operation. 3.5.1.5 During startup when the intermediate range instruments cc=e on 1 scale, the overlap between the inter =ediate range and the source l range instrumentation shall not be less than one decade. If the j gg, overlap is less than one decade, the flux level shall not be greater than that readable on the source range instruments until the one decade overlap is achieved. 3.5.1.6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the failed trip device shall be manually removed within 30 minutes. The condition will be corrected and the re=aining trip devices shall be tested within eight hours. If the condition is 19. not corrected and the re=aining trip devices tested within the eight hour period, the reactor shall be placed in the hot shutdown condition within an additional four hcurs. v 3.5-1 Rev. 19. 5/5/72
e . 34RPS Every reasonable effort will be =ade to =aintain all safety instrumentation in operation. A startup is not permitted unless three power range neutron instru-ment channels and two channels each of the following are operable: four reactor coolant temperature instrument channels, four reactor coolant flow instrument channels, four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and high reactor building pressure instrument channels. The engineered safety features actuation system must have two analog chaanels functioning correctly prior to startup. Operation at rated pcVer is permitted as long as the systems have at least the cedundancy requirements of Column 3 (Table 3.5.1-1). This is in agreement with redundancy and single failure criteria of Izzz 279 as described in FSAR Section 7. There are four reactor protective channels. A fifth channel that is isolated from the reactor protective system is provided as a part of the reactor control system. Normal trip logic is two out of four, Required trip logic for the power range instrumentation channals is two out of three. Minimum trip logic on other channels is one out of two. The four reactor protective channela were provided with ey operated bypass switches to allev on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alarm and lights to q indicate when that channel is bypassed. There will be ona reactor protective system bypass switch key permitted in the control room. That key will be under the administrative control of the Shif t Supervisor. Spare keys will be maintained in a locked storage accessible only to the Superintendent. Each reactor protective channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used. There are fcar shutdown bypass keys in the control room under the administrative control of the Shift Supervisor. These keys will not be used during reactor power operation. The source range and intermediate range nuclear instrumentation overlap by one g* decade of neutron flux. This decade overlap will be achieved at 10-10 amps on the intermediate range instrument. Power is normally supplied to the control rod drive mechanisms from two separate parallel 600 volt sources. Redundant trip devices are employed in each of these sources. If any one of these tri, devices fails in the untripped state on-line repairs to the failed device, when practical, will be made, and the remaining trip devices will be tested. Four hours is ample time to test the remaining trip devices and in many cases make on-line repairs. REFERENCE FSAR, Section 7.1 Y 3.5-2 Rev, 19. ' 5M 2
. e TABLE 3.5.1-1 L
INSTRL'MENTS CPERATING CONDITIONS 8 (A) (S) (C) Minimus Mini =um Operator Action If Conditions Operable Degree Of Of Colu=n A and 3 Functional Unit Channels Redundancv Cannot Be Met
- 1. Nuclear Instru=entation 1 0 3 ring to hot shutdown within Intermediate Range 12 hours (b) i Channels
- 2. Nuclear Instrumentation 1 0 3 ring to hot shutdown within Source Range Channels 12 hours (b)(c) i
- 3. RPS Manual Pushbutton 1 0 Bring to het shutdown within 12 hours
- 4. RPS Power Range 3(a) 1(a) Bring to hot shutdown within Instrument Channels 12 hours
- 5. RPS Reactor Coolant 2(d) 1 Bring to hot shutdown within Temperature Instrument 12 houra Channels
- 6. RPS Pressure-Temperature 2(d) 1 Bring to hot shutdown within Instruments Channels 12 hours
- 7. RPS Flux Imbalance 2 1 Bring to hot shutdown within Flow Instrument Channels 12 hours
- 8. RPS Reactor Coolant Pressure 1
- a. High Reactor Coolant 2 1 Bring to hot shutdown within Pressure Instrument 12 hours i Channels !
I
- b. Low Reactor Coolant 2 1 Bring to hot shutdown within I Pressure Channels 12 hours
- 9. RPS Power-Number of Pumps 2 1 3 ring to hot shutdown within Instru=ent Channels 12 hours 1
- 10. RPS High Reactor Building 2 1 Bring to hot shutdown within I Pressure Channels 12 hours L
Rev. 19 May 5, 1972 (Entire Paga Revised) i 3.5-3 a
{ TA3L3 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (Cont'd) a (A) , Minimum (3) (C) Operable Minimum Operator Action if Conditions Analog Degree Of Of Column A and 3 Functional Unit Channals Redundanev Canno Be Met
- 11. ESP High Pressure Injection System
- a. Reactor Coolant 2 1 Bring to hot shutdown within Pressure Instru- 12 hours (a) ment Channels
- b. Reactor Building 2 1 3 ring to hot shutdown within 4 PSIG Instrument 12 hours (e)
Channels
- c. Manual Pushbutton 2 1 3 ring to hot shutdown within 12 hours (e)
- 12. ESF Lcw Pressure In-jection System
- a. Reactor Coolant 2 1 3 ring to hot shutdown within ~s/
Pressure Instrc=cnt 12 hours (e) Channels
- b. Reactor Building 2 1 Bring to hot shutdown within 4 PSIG Instrument 12 hours (e)
Channels
- c. Manual Pushbutton 2 1 Bring to het shutdown within l 12 hcurs (e)
- 13. ESF Reactor Building i Isolation & Reactor )
Building Cooling Systes ! I
- a. Reactor Building 2 1 3 ring to hot shutdown within i 4 PSIG Instrument 12 hours (e)
Channel t'
- b. Manual Pushbutton 2 3 ring to hot shutdown within 12 hours (e) 14 ESF Reactor Building Spray System
- a. Reactor 3uilding 2 1 Bring to hot shutdcun within s_/
10 PSIG Instrument 12 hours (e) Channel l Rev. 19 May 5, 1972 l 3.5-4 (Entire Page Revised) 1
TABLE 3.5.1-1 i INSTRUMZNTS OPERATING CONDITIONS (Cont'd) I (A) Minimum (3) (C) Operable Mini =um Operate:' Action If Conditions Analog Degree Of Of Column A and B
- Functional Unit Channels. Redundancv Cannot Be Met i b. Manual Pushbutton 2 1 3 ring to hot shutdown within 12 hours (e)
- 15. Turbine Stop Valves 2 1 Bring to hot shutdown within Closure 12 hours (f)(e)
(a) For channel testing, calibration, or maintenance, the minitu= nu=her of cperable channels may be two and a degree of redundancy of one for a =ax1=um of four hours. (b) When 2 of 4 power range instrument channels are greater than 10% full povar, hot shutdown is not required. (c) When 1 of 2 intermediate range instru=ent channels is graater than 10-10 amps, hot shutdown is not required. (d) Single loop operation at power (af ter testing and approval by the AEC/DRL) is not per=1tted unless the operating channels are the two receiving Reactor Coolant Temperature from operating loop. (e) If minimum conditions are not =et within 48 hours af ter hot shutdown, the unit shall be in the cold shutdewn, condition v1 thin 24 hours. (f) One operable channel with zero =inimum degree of redundancy is allowed for 24 hours before going to the hot shutdown condition. l ) w Rev. 19 May 5, 1972 (Entire Page Revised) 3.5-5
V V i J 3.5-6 Rev. 19 5/5/72 (Page Deleted)
_ . _ _ . _ - . . - - . - __. . = . . -_ __ - -._- - _ . . _ . _ _ _ _ _ _ - _ _ _ - i , , I l I 3.5.2 Control Rod Groue and Power Distributien T.inits
- Applicability t
This specification applies to power distribution and operation of control rods during power operation. j I objective j t l To assure an acceptable core power distribution during power operation, to !
! set a limit on potential reactivity insertion from a hypothetical control j rod ejection, and to assure cora subcriticality after a reactor trip.
Specification l 3.5.2.1 The available shutdown margin shall be not less than 1% ak/k with the highest worth control red fully withdrawn. I j 3.5.2.2 Operation with inoperable rods: i n. Operation with more than one inoperable rod as defined in
. Specification 4.7.1 and 4.7.2.1, in the safety or regulating red banks shall not be permitted.
- b. If a control rod in the regulating and/or safety rod banks ia
; declared inoperable in the withdrawn position as defined in ! N, Specification 4.7.1.1 and 4.7.1.3, an evaluation shall be 1 initiated i=cediately to verify the existance of 1% Ak/k hot i shutdown margin. Boration may be initiated either to the worth of the inoperable rod or until the regularing and transient rod banks are fully withdrawn, whichever occurs first. Simultaneously a program of exercising the remaining regulating and safety rods shall be initiated to verify operability. , c. If within one (1) hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determined that a 1% ak/k het shutdown cargin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be ; brou3ht to the hot standby condition until this =argin is ,' established.
1
- d. Following the fetermination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised within 24 18.l hours and exercised weekly until the red problem is solved. l 3.5.2.3 The worth of a single inserted control red shall not exceed 0.5%
l ak/k at rated power or 1.0% ak/k at hot zero power except for physics testing when the requirements of Specification 3.1.9 shall apply.
- 3.5.2.4 Power distribution shall be within. design ther=al limits.
I-
\~-
3.5-7 Rev. la' 3/10/72 _ h
f f I
- a. If the quadrant power tilt exceeds 10% axcept for physics tests, power shall be limited to 80% of the ther=al power allowabic for '-"#
the reactor coolant pu=p combination. 1
- b. If the quadrant pcwer tilt exceeds 20% except for physics tests, j pcwer shall be limited to 60% of the ther=al power allowable for the reactor coolant pump co=bination.
l j c. If a control rod in the regulating and/or safety rod banks is declared inoperable per 4.7.1.2. power shall be reduced to !l 60% of the thermal power allowable for the reactor coolant pump I combination,
- d. If a transient rod or APSR is declared inoperabic the jl retaining rods in the group shall be tri =ed to the sa=e
- position and nor=al operatien may continue with the following restriction for the inoperable bank:
l i (1) No =aneuvering or dilutten cycle is allowed that j 18.l requires movecent of the transient rod bank. I i (2) The APSR bank shall not be moved but the inhalance li=1ts as defined in Specification 2.1 apply. , 3.5.2.5 Operating regulating red group overlap shall not exceed 30% i between two sequential groups. l 3.5.2.6 The control rod drive patch panels shall be locked at all times with s, j limited access to be authorized by the superintendent. Bases ] The 30% overlap between successive control rod groups is allcwed since the worth of a rod is Icwer at the upper and lower part of the stroke. Control I rods are arranged in groups or banks defined as follows: j Group Function .i 1 Safety 2 Safety 3 Safety j 4 Safety 5 Regulating 6 Regulating 7 Xenon transient override s ! S APSR (Axial Pcwer Shaping Bank)
- s. ,/
l Control rod groups are withdrawn in secuence beginning with group 1. Groups 5, 6, and 7 are overlapped 25%. The normal position at pcwer is for groups 6 and 7 to be partially inserted. 3.5-S . Rev. is 3/10/72 S n- ,,,e7 -, -r -~-e r - m y . _.-
, r- -
y
6 t-1 The minimum available rod worth provides for achieving hot shutdown by i y reactor trip at any time assuming the highest wor:h control rod remains in the full out position.
- Inserted rod groups during power operation will not contain single rod worths
! greater than 0.5% Ak/k. This value has been shown to be safe by the safety I analysis of the hypothetical rod ejection accident.(2) A single inserted ! control rod worth of 1.0% ak/k at beginning of lifa, hot, zero power would result in the same transient peak thermal power and therefore the same i environ = ental consequences as a 0.5% ak/k ejected rod worth at rated power.
+
The quadrant power tilt limita set forth in Specification 3.5.2.3 have been , established within the ther=al analysis design base using the definition of
- quadrant power tilt given in Secticn 1.6.
! During the physics testing program, the high flux trip setpoints arc
; administratively set as follows to assure an additional safety margin is provided:
Test Power Trio Setpoint i l 0 <5% l 15 50% t 40 50% l 50 60% 75 85% D RFFERENCES
- (1) FSAR, Secticn 3.2.2.1.2 (2) 7SAR, Section 14.2.2.2 i
V l l i J 1 1
. D i
i 3.5-9 _. __ _-- _, ,_ ,.. _._..- ,._ a . . _ . - _ _ . . . . , . _ _ . _ -- ,,, .,-
1 t 3.5.3 Engineered safetv Pestures Protective Svate: Actuation Setcoints Applicability This specification applies to the engineered safety features protective systen actuation setpoints. Objective To provide for automatic initiation of the engineered safety features protective system in the event of a breach of RCS integrity. Specification i The engineered safety features protective actuation setpoints and per=issible . bypasses shall be as follows: Functional Unit Action Setooint Reactor Building Spray 10 'f High Reactor Building gp psig Pressure High-Pressure Injection g4 psig Low-Pressure Injection g4 psig Scart Reactor 3uilding
- 19. Cooling & Reactor Building Isolation g4 psig g Penetration Room Ventilation g4 psig Low Reactor Coolant High Pressure Injection (l) gl500 psig System Pressure Lcw Pressure Injection (2) g500 psig ,
(1) May be bypassed below 1750 psig and is automatically reinstated above 1750 psig. (2) May be bypassed below 900 psig and is automatically reinstated above 900 psig. Bases High Reactor Building Pressure
/# )(
The basis for the J6 psig and 4 psig setpoints for the high pressure signal is to establish a setting which would be reached ic=ediately in the event of a DBA, cover the entire spectrum of break sizes and yet be far enough above nor=al operation maximum internal pressure to prevent spurious initiation. Low Reactor Coolant System Pressure The basis for the 1500 psig low reactor coolant pressure setpoint for high pressure injection initiation and 500 psig for low pressure injection is to 3.5-10 Rev. 19. 5/5/72
catablish a value which is high enough auch that prctection is provided for the entire spectrum of break si cs and is far enough belcw nor:s1 operating pressure to prevent spurious initiation.(1) References (1) FSAR, Section 14.2.2.3. v 1 l l I i l b 3.5-11
3.5.4 Incore Insert =entation Applicability d Applies to the operability of the incore instru=entation system. Objective To specify the functional and operational requirements of the incore instru-nentation system. Specification Above 807 of operating power determined by the reactor coolant pu=p ccmbina-tion, reference table 2.3.1, at least 23 individual incere detectors shall be operable to check gross core power distribution and to assist in the periodic calibration of the out-of-core detectors in regard to the core imbalance trip limits. The detectors shall be arranged as follows and cay be a part of both basic arrangements. 3.5.4.1 Axial Imbalance
- a. Three detectors in each of 3 strings shall lie in the same axial plane with 1 plane in each axial core half.
- b. The axial planes in each core half shall be sy= metrical abcut the core mid-plane, J
- c. The detector strings shall not have radial sy==etry.
3.5.4.2 Radial Tilt
- a. Two sets of 4 detectors shall lie in each core half. Each set of 4 shall lie in the same axial plane . The two sets in the sama core I half may lie in the same axial plane. ,
l
- b. Detectors in the same plane shall have quarter core radial.sy==etry. !
Bases I l A system of 52 incore flux detector asse=blies with 7 detectors per assembly ! has been provided. The system includes data display and record functions and is used primarily fcr out-of-core nuclear instrumentation calibration and for core power distribution verification. A. The out-of-core nuclear instrumentation calibration includes:
- 1. Calibration of the split detectors at initial reactor startup, during the power escalatien program, and periodically thereafter.
- 2. A comparison check with the incere instrumentation in the event one of the four out-of-core detector asse blies gives abnormal readings during operation.
V 3.5-12
O . t
- 3. Confirmation that the out-of-core axial pcwer splits are as expected.
B. Core power distribution verification includes:
- 1. Measurement at low power initial reactor startup to check that power distribution is consistent with calculations.
- 2. Subsequent checks during operation to insure that pcwcr distribution is consistent with calculations.
- 3. Indication of power distribution in the event that abnormal I situations occur during reactor operation.
C. The safety of unit operation at or below 80% of operating power (l) for the reactor coolant pump co=binations without the core imbalance trip system has been determined by extensive 3-D calculations. This will be verified during the physics startup testing program. l D. The minimum requirement for 23 individual incore detectors is based 4 on the following: I
- 1. An adequate axial imbalance indication can be obtained with 9 1
individual detectors. Figure 3.5.4-1 shows three detector
! strings with 3 detectors per string that will indicate an axial imbalance that is within 8% (calculatad) of the real core i=-
i balance. The three i etector strings are the center one, one
\-- frem the inner ring of symmetrical strings and one from the outer ring of symmetrical strings. Both steady state and design transient data from the Ocenee 1 maneuvering analysis were used l
for * *1 camparison.
- 2. Figure 3.5.4-2 shows a detection scheme which will indicate the radial power distributx: with 16 individual detectors.
! The readings from 2 detectors in a radial quadrant at either plane can be compared with readings from the other quadrants t
to measure a radial flux tilt.
- 3. Figure 3.5.4-3 combines Figures 3.5.4-1 and 3.5.4-2 to illustrate a set of 23 individual detectors that can be specified as a minimum for axial imbalance determination and radial tilt indication, as well .as for the determination of gross core power distributions. Startup testing will verify the adequacy of this set of detectors for the above functions.
E. At icast 23 specified incore detectors will be operable to check power distribution above 80% power determined by reactor coolant pump combini-tion. These incore detectors will be read out either on the computer or on a recorder. If 23 detectors in the specified locations are not operable, power will be decreased to or below 80% for the operating reactor coolant pump combination. REFERENCE (1) FSAR, Section 4.1.1.3 3.5-13 i
J
.o
- --- Lack radial sy::::::etry
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4 7 -% g [ / # 1V h :\
; k ,\ "
p / Axial Plane
! x - -
o.
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r/f_ _%\ N
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=
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\ \
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g N- ----- - : Bot:cs Axial Core Half aM
- r- ~
J l% M i [+ +
% N>g h
\ \ /
' N /
\ -
'~/,/
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/
- 'N \
INCORE INSTRUMENT 4Tl0N SPECIFICATl0N AllAL l# BALANCE INDICATION
'8'C*'i
{"M> OCONEE Figure NUCLEAR 3.5.4 -1 STATION 3.5-14
1 I _ g Radial
---.I-I I
I
+ m..
-l% N 7 %
. __ [ % %_
Radial Sy=etry
, ( -4
/ in this plane i N -l- /
e
, MA o
O [ ~ N 5M (
/
\ /
/
" % /
* ~"
\ /
w' ~ v s . , - ~
/ % Radial Sy= metry
( \. q j
/
in this plane
,- s
/ N
/ N INCORE INSTRUMENTATION SPECIFICAT10h RADIAL FLUI TILT INDICATION v
pr eom OCONEE NUCLEAR STATION Figure '3.5.4 - 2 3.5-15
4 d M I
/ [/r2 [ k Ab N 5
5 E k x - dy" -
= 7- -~T y N
$ ([/ 3I[7#_ 4> h b 48 /
k( m-i
)
E
=
a '..'s x\ ' -
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S % d 2,/- - s M [ G114 O d8 /
$ N / 1' N 40 M ,
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s' %
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INCORE INSTRUMENTATION SPECIFICATICN V tb bailp's\ OCONEE NUCLEAF. STATION
- . Figure 3.5.4 - 3 3.5-16
4 3.6 REACTOR BUILDING
, .,_ Appli cability Applies to the containment when the reacttr is suberitical by less than 1% ak/k.
Objective To assure containment integrity during startup and operation. Specification 3.6.1 Containment integrity shall be maintained whenever all three (3) of the following conditions exist:
.t . Reactor coolant pressure is 300 psig or greater.
- b. Reactor coolant te=perature is 200 F or greater.
- c. Nuclear fuel is in the core, t
3.6.2 Containment integrity shall be maintained when the reactor coolant system is open to the containment atmosphere and the requirements for a re fuelin3 shutdown are not =ct. v fh9 3.6.3 Positive reactivity insertions which would result in the reactor eh,h/, s_ being suberitical by less than 1% ak/k shall not be made by control A rod motion or boron dilution whenever the containment integrity is not intact. 3.6.4 A reactor shall not oe critical if the reactor building internal pressure exceeds 1.5 psig or 5 inches Hg vacuum. 3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all nanual containment isolation valves which should be closed are closed and tagged. Bases The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence no pressure buildup in the containment if the Reactor Coolant System ruptures. The selected shutdown conditions are based en the type of activities that are being carried out and will preclude criticality in any occurrence. The reactor building is designed for an internal pressure of 59 psig and an external pressure 3.0 psi greater than the internal pressure. The design external pressure of 3.0 psi corresponds to a margin of 0.5 psi above the differential pressure that could be_ developed if the building is sealed with an internal temperature of 120*F with a barometric pressure of 29.0 inches of Hg and the building is subsequently caoled to an internal temperature of 80*F - with a concurrent rise in barometric >ressure to 31.0 inches of Hg. The I weather conditions assumed here are conservative since an evaluation o' ' l 1 3.6-1 l l
National Weather Service records for this area indicates that from 1918 :o 1970 the lowest barometric pressure recorded is 29.05 inches of Hg. and the highest is 30.85 inches of Hg. d When containment integrity is established, the limits of 10 CFR 100 will not be exceeded should the max 1=um hypothetical accident occur. REFERENCES FSAR, Section 5 1 I l l 1 l 1 l l l 3.6-2
2 3.7 AUXILIARY ELICTRICAL SYS!IMS V Applicability Applies to the availability of off-site and on-site electrical power for station operation and for operation of station auxiliaries. Obiective To define those conditions of electrical power availability necessary to provide for safe reactor operation and to provide for continuing avail-ability of engineered safety features syste=s in an unrestricted =anner and to prescribe safety evaluation and reporting requirenents to be followed in the event that the auxiliary electric pcwer systems become degraded. J. Specification 3.7.1 Under normal conditions the reactor shall not be brought critical from a cold shutdown condition unless the following conditions are met:
- a. At least two 230 kV transmission lines shall be in service. One of these two lines may be out for test or =aint2nanca for periods up to 24 hours once the reactor has been brought critical.
b. Start-up Transfor=er No. CTl or CT2 shall be connected to the Unit 1 4160 V Main Feeder Bus No. 1 and No. 2.
c. An operable Keowee Hydro unit shall be available to supply power through the Underground Feeder Sus, Transformer No. CT4, and the Station 4160 V Standby Susas Nos.1 or 2 feeding the associated 4160 Main Feeder Bus. The other Keowee Hydro unit shall be available to supply pcwer througn Start-up Transformer CTl or CT2 to Unit 14160V Main Feeder Buses Nos. 1 and 2, except that, one Keowe; Hydro Unit =ay be inoperable for periods not exceeding 14 hours for test or maintenance. d. Reactor Coolant Pump Buses 1TA or AT3 shall be energized. e. The two 4160 volt main feeder buses shall be energized except one may be de-energized for test or =aintenance for periods not exceeding 24 hours af ter the reactor has been brought critical. f. The three 4160 V Engineered Safety Features Switchgear 3uses ITC,1TD, and ITE shall be energized except one may be de-energized for test or maintenance for periods not exceeding 24 hours after the reactor has been brought critical. v 3.7-1
- g. The three 600 V load centers IX8, 1X9, and 1X10 plus the three 600V-208Y Engineered Safety Features MCC Buses 1XS1, 1XS2, and IXS3 shall be energized, except that one string may be de-energized for test or maintenance for d i periods of not exceeding 24 hours after the reactor has baen brought critical.
- h. The Unit 1 & 2 125 VDC instrumentation and control batteries with their respective chargers, buses, diode monitors, and
! diodes supplying vital instrumentation and control static '
i inverters shown in Figure 8-5 of the ?SAR, consistent with ) Notes 1 through 3 thereon, shall be operable except that one j complete single string or any portion thereof of redundant 4 chargers, batteries, buses and all associated isolating j transfer diodes may be de-energized for test or maintenance for periods of not exceeding 24 hours.
- 1. The 125V DC switching station batteries with their respective
] 19. chargers, buses, and isolating diodes shall be operable except
; single strings of redundant buses, chargers, batteries, and all of the related diode assemblies =ay be de-energized for j test or maintenance for periods of not exceeding 24 hours, i
- j. The Keowee batteries with their respective chargers, buses, and laolating diodes shall be operable except that redundant components may be de-energized for test or maintenance for periods not exceading 24 hcurs.
j k. The level of the Keowee Reservoir shall be at least 775 feet above sea level, j i
- 3.7.2 In the event that all conditions in Specification 3.7.1 are me:
1 except that one of the two Keowee Hydro units becomes unavailable ! for longer than the test or maintenance period of 24 hours 'or !
! the underground feeder circuit to the standby buses is lost, i the reactor shall be permitted to remain critical or restarted i provided the following restrictions are observed:
- a. A Lee Steam Station combustion turbine shall be operating and energizing the 4160 V standby power bus through the 100 kV transmission circuit which is completely separate from the system grid and non-safety-related loads. This shall be in ef-feet prior to restart or within 30 minutes if a't. power when a Hydro unit is lost.
I
- b. The reactor coolant Tavg shall be above 500*F. l
- c. Within 24 hours after loss of a hydro unit, or the underground feeder, this fact shall be reported to the Director, Division of Compliance, Region II, to be followed by a written. report of the circumstances of the outage if it is expected to exceed 24 hours and the estimated time to return the Hydro unit o. i underground feeder circuit to operating condition. !
3.7.3 In the event that all conditions of Specification 3.7.1 are met I except that all.230 kV transmission lines are lost, the reactor shall be permitted to remain critical or restarted provided
- the following restrictions are observed
- ,
3.7-2
- .ev. 19. 5/5/72
. i
j l } %, a. A Lee Station combustion turbine shall be operating and energizing j the 4160 V standby power bus through the 100 kV transmission 1 I circuit which is completely separata from the system grid and non-safety-related loads. It shall be available prior to a restart or within 30 minutes if this degradation occurs dLring operation. ! b. Within 24 hours of loss of all 230 kV transmission lines, this , degradation shall be reported to the Director, Division of Com- , pliance, Region II, in accordance with Spicification 6.6. If )- the outage is expected to continue beyond 24 hours, the circum-l stances of the cutage and the esti=ated time to return at least two 230 kV circuits to service shall be reportid. 3.7.4 Any degradation beyond specification 3.7.2 or 3.7.3 above shall be reported to the Director, Division of Compliance, Region II, within 24 hours. A safety evaluation shall be performed by Duke Power Company for the specific situation involved which justifies the safest course of action to be taken. The results of this evaluation i together with plans for expediting the return to the unrestricted cperation conditions of specification 3.7.1 above shall be sub-mitted in a written report to the Director, Division of Reactor Licensing, with a copy to the Director of Region II Compliance Office j within 5 days. i s,, Bases 1 I; The auxiliary electric.1 power systems are so arranged that no single con-tingency can inactivate enough engineered safety features to jeopardize plant safety. These systems were designed to meet the folicwing criteria:
" Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity and testability to permit the functions required of the engineered safety features. As a minimum, the onsite power system and the offsite power system shall each, independently, provide this capacity assuming a failure of a single active component in each power system."
The aux 1. iary power system meets the above criteria and the intent of proposed AEC Criterion 17 (published February 20, 1971) with one exception. That exception is the Unit 1 startup transformer (CTI) whose loss would cause loss of all offsite power. This is acceptable because of the high reliability of such transformers and the relatively short period of time that will clapse before a redundant transformer, C12, the Unit 2 startup transformer, is permanently-installed as a redundant means of obtaining Unit 14160 ESF bus power from the 230 kV switchyard. The adequacies of the ac and de systems are discussed below as are the bases for permitting degraded conditions for ac power. , i 4 MW 3.7-3
1 ] Capacity of AC Systems
'~
The auxiliaries activated by ES7 signals (4.8 MVa) plus other safety
- related auxiliaries activated by the operator follewing a loss-of-coolant 4
accident (4.9 MVa) require a total ac power capacity of 9.7 MVa. The minimum ac power espacity available from the onsite power systems (Keowee liydro Units) is 12 MVa (limited by transformer CT4) if furnished by the underground circuit or 45 MVa (limiced by CTl or CT2) if furnished through t the 230 KV switchyard. The minimum capacity available from the multiple I 230 KV of fsite transmission lines is 45 MVa (limited by CT1 or CT2) . j Capacity available from the backup 100 KV offsite transmission line (Lee j. Station Cas Turbine Generator) is 12 MVa (limited by CTS). I Thus, the minimum available capacity from any one of the multiple sources of ac power,12 MVa, is adequate. Capacity of DC Systems , i Normally, ac pcwer is rectified and supplies the de system buses as well as keeping the storage batteries on these buses in a charged state. Upon loss of this normal ac source of power the de auxiliary systems important to reactor safety have adequate stored capacity (ampere-hours) to inde-1 pendently supply their required emergency loads for at least I hour. One hour is considered to be conservative since there are redundant sources of ac power previding energy to thesa de auxiliary systems. The loss of all , ac power to any de system is expected to occur very infrequently, and for very short periods of time. The following tabulation demonstrates the nargin of installed battery charger rating and battery capacity when com- *=d pared to 1 hour of operation (a) with ac power (in amps)and (b) without ac power (in ampere hours) for each of the three safety-related de syste=s installed at Oconee: A. 125 Vdc Instrumentation and Control Power System Charger ICA, IC3 or 1CS Rating a. 600 amps ea. 1 Battery 1CA and ICS Co=bined Capacity b. 698 ampere-hours ! Actual active loads on both 125 Vdc a. First =in. - 1371 amps I & C buses 1DCA and 1DCB next 59 min. - 568.5 a=ps f during ist hour of LOCA b. 581.9 ampere-hours B. 125 Vdc Switching Station Power System Charger SY-1, SY-2 or SY-S Racing a. 50 a=ps ea. Battery SY-1 or SY-2 Capacity b. 14.4 ampere-hours Actual active load per battery a. First min. - 130 amps during ist hour or LOCA next 59 min. - 10 amps
- b. 12 ampere-hours
.18. l C. 125 Vdc Keowce Station Power System l
1 l Charger No. 1, No. 2 or Standby Racing a. 200 amps ca. s./ , Battery No. 1 or No. 2 Capacity b. 233 ampere-hours 3.7-4 Rev. 18 3/10/72 4-
Actual active load per battery a. First min. - 103*. amps during ist hour of LOCA next 59 min. - 179.4 amps
- b. 193.6 ampere-hours Redundancy of A.C. Syste=a
! There are three 4160 engineered safety feature switchgear buses. Each bus l
can receive power from either of two 4160 main feeder *ouses. Each feeder bus in turn can receive power from the 230 KV switchyard through the co= mon Unit 1 startup transformer (CTI) or from the 4160 V standby bus. Unit 2 startup transformer (CT2) can be placed in service in one hour. The standby bus can receive power from the Hydro Station through the underground feeder circuit or from a combustion turbine generator at the Lee Steam Station over an isolated 100 KV transmission line. The 230 KV switchyard can receive power f rom the onsite Keowee Hydro Station or from several offsire sources
! via transmission lines which connect the Oconee Station with the Duke Power System power distribution network.
] Redundancy of DC Systems i A. 125Vdc Instrument and Control Power System 1
; All reactor protection and engineered safety features loads on this i syatec can be pcwcred from either the Unit 1 or Unit 2125Vdc Instrument j and Control Power Buses. The Unit 1 125Vdc Instrument and Control Power Suses can be powcred from evo battery banks and three battery chargers.
As shown above, one battery (e.g. , ICA) can supply all loads for one g hour. Also, one battery charger can supply all connected ES? and re-l actor protection loads. I
- 3. 125Vdc Switching Station Power System There are two essentially independent subsyste=s each complete with an ac/dc power supply (battery charger), a battery bank, a battery charger bus, motor control center (distribution panel). All safety-related equipment and the relay house in which it is located are Class I (Seismic) design. Each subsystem provides the necessary de power to
- 1
! i
- a. continuously monitor operations of the protective relaying
- b. isolate Oconee (including Keovee) from all external 230 KV grid l faults
- c. connect on-site power to Oconee from a Keowee hydro unit or
- d. restore of f-site power to Oconee from non-faulted portions of the external 230 KV grid.
Provisions are included to manually connect a standby battery charger to either battery / charger bus. C. 125Vdc Keowee Station Power System - i 4
'.. There are essentially two independent physically separated Class I 3.7-5 1
1
(Seismic) subsyste=s, each complete with an ac/dc power supply (charger) a battery bank, a battery / charger bus and a de distribution center. , Each subsystem provides the necessary power to auto =atically or manually N=*/ start, control and protect one of the hydro units. An open or short in any one battery, charger or de distribution center cannot cause loss of both h /dro units. The 230 KV sources, while expected to have excellent availability, are not under the direct control of the Oconce Station and, based on past experience can not be assumed to be available at all times. The operation of the on-site hydro-station is under the direct control of the Oconee Station and requires no off-site power to startup. Therefore an on-site backup source of auxiliary power not subject to failure from the same cause as offsite power is provided in the form of twin hydro-electric turbine generators powered through a common penstock by water taken from Lake Keowee. The use of a common penstock is justified on the basis of past hydro-plant experience of the Duke Power Company (since 1919) which indicates that the cumulative need to devater the penstock can be expected to be limited to about one day a year, principally for inspection, plus perhaps 4 days every tenth year. In all other cases it is expected that wnen one hydro unit is out for maintenance, the other unit will be available I for service. In the event that only one hydro unit is available to backup the off-site power sources, and it is considered i=portant for the Oconee Unit i reactor to remain critical or return to criticality frc= a hot shutdown condition, ,,/ the Lee Station combustion turbine is again available to assure a continued supply of shutdown power in the event that an external event should cause loss of all off-site power. In a similar manner, in the event that none of the sources of off-site power is available and it is considered i=portant to continue to maintain the Oconee Unit I reactor critical or return it to criticality from a hot shut-down condition a Lee Station gas turbine can be made available as an additional backup source of power, thus assuring continued availability of auxiliary power to perform an orderly shutdown of Ocenee Unit 1 should a problem develop requiring shutdown of both hydro units. There may be a rare occasion where both Hydro units are unexpectedly lost and there are compelling reasons to maintain the Oconee Unit I reactor critical or return it to criticality frem hot shutdown condirions for a 1 specific period of time rather than require it to remain soberitical or be shutdown. A scheduled shutdown for inspection or a shut 3cwn to perform minor maintenance would not constitute a compelling reason. Factors to consider in justification of such a rare. limited period of criticality without the hydro station available wouid include number of offsite 230KV power sources available, availability of Unit 2 startup tr ansformer, availability of the Lee Gas Turbine, weather conditions and all other factors , which could bear on potential for loss of these power sources. Also, the l evaluation should show that reactor safety will not be compromised if during ' operation under. such further degradation, an additional loss of ac power _j should be suffered. 3.7-6
l I 3.8 FUEL LOADING AND REFUELING
\-' Applicability i
- Applies to fuel loading and refueling operations. !
! Objective j To assure that fuel loading and refueling operations are performed in a
- responsible manner.
Specification 3.8.1 Radiation levels in the reactor building refueling area shall be onitored by RIA-48 and RIA-49. Radiation levels in the spent fuel storage area shall be =enitored by RIA-41. If any of these instruments becc=es inoperable, portable survey instru=entation, having the appropriate ranges and sensitivity to fully protect individuals involved in refueling operatier., shall be used until the permanent instrumentation is returned to service. 3.8.2 Core suberitical neutron flux shall be continuously monitored by at least two neutron flux monitors, each with ccatinuous indication available, whenever core gec=etry is being changed. When core gec=ctry is not being changed, :: least ona neutren flu:: =onitor shall be in service. 4 N-- 3.8.3 At least one lcw pressure injection pu=p and cooler shall be ope rable . 3.8.4 During reactor vessel head re= oval and while loading and unloading fuel f ro= the reactor, the boren concentration shall be maintained at not less than that required to shutdown the core to a keff 1 0.99 if all control rods were removed. 3.8.5 Direct co==unicatiens between the control roo= and the refueling personnel in the reactor building shall exist whenever changes in core geometry are taking place. 3.8.6 During the handling of irradiated fuel in the reactor building at least one door on the personnel and emergency hatches shall be closed. The equip =ent hatch cover shall be in place with a =inimum of four bolts securing the cover to the sealing surfaces. 3.8.7 Soth isolation valves in lines containing automatic contain=ent iso-t lation valves shall be operable, or at least one shall be closed. 3.8.8 When two irradiated fuel asse=blies are being handled simultaneously within the fuel transfer canal, a minimum of 10 feet separation shall be maintained between the assemblies at all times. Irradiated fuel asse=blies may be handled with the Auxiliary Hoist provided no other irradiated fuel asse=bly is being handled in the L fuel transfer canal. 3.8-1 4
I L 3.8.9 If any of the above specified limiting conditions for fuel loading and refueling are not cet, movement cf fuel into the reactor core %./ shall ccase; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may in-crease the reactivity of the core shall be made. 3.8.10 The reactor building purge system, including the radiation =onitor, 18.l RIA 45, which initiates purge isolation, shall be tested and veri-l < fied to be operable i==ediately prior to refueling operations. 3.8.11 Irradiated fuel shall not be re=oved from the reactor until the unit
- has been subcritical for at least 72 hours.
i , Bases Detailed written procedures will be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.7 of the FSAR incorporating built-in inter-locks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and 4 safety. If no change is being made in core geo=etry, one flux monitor is , sufficient. This permits caintenance'on the instru=entation. Continuous moni-j toring of radiation levels and neutron flux provides i= mediate indication of an unsafe condition. The low pressure injection pump is used to maintain a uniform , boron concentration. (1) The shutdown margin indicatec in Specification 3.3.4 will keep the core suberitical, even with all control rods withdrawn from the j core. (2) The boren concentration will be maintained above 1,300 ppm. Al- ,,,/ though this concentration is sufficient to maintain the core keff < 0.99 if all
< the control rods were removed from the core, only a few control rods will be
- removed at any one time during fuel shuffling and replacement. The kegg with all rods in the core and with refueling boron concentration is apprcximately 0.9. Specification 3.3.5 allows the control room operator to inform the reactor j building personnel of any impending unsafe condition detected from the main i control board indicators during fuel movement.
4 The specification requiring testing of the reactor building purge isolation is , to verify that these components will function as required should a fuel hand-ling accident occur which resulted in the release of significant fission products. Specification 3.8.10 is required as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for j 72 hours.(3) REFERENCES l l (1) FSAR, Section 9.7 ; (2) FSAR, Section 14.2.2.1 (3) FSAR, Section 14.2.2.1.2 s_si Rev. 18 3/10/72 I 3.8-2
-m - -
m , ~ - e, -- ,,m
3.9 RADI0 ACTIVE EFFLUENTS
-- Applicability j Applies to disposal of radioactive liquid, and gaseous wastes from the station.
Obj ec t ive To define the conditions for release of radioactive liquid waste to the Keowce flydro dam tailrace and radioactive gases to the unit vent to assure that any radioactive material released is kept as low as practicable and, in any event, is within the limits of 10 CFR 20. I Specification ; 3.9.1 General Equipment installed in the radioactive waste systems shall be main-tained and used to assure that except under unusual conditions, releases of radioactive materials in effluents will be kept at small fractions of the limits specified in 20.106 of 10 CFR 20. The Licensee is per-mitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than such small fractions, but atill within the limits specified in 20.106 of 10 CFR 20. It is expected that using this operational flexibility under unusual operating conditions the s,, Licensee will exert his best efforts to keep levels of radioactive material in effluents as low as practicable. 3.9.2 Liquid Wastes 3.9.2.1 Operating procedures shall be developed and used, and equipment which has been installed to maintain control over radioactive materials in liquid effluents produced during normal reactor operations, including expected occurrences, shall be maintained and used to keep levels of radioactive material in effluents released to unrestricted areas as icw as practicable. Inputs to the low activity waste tank may be released directly to the Keowee tailrace without further processing than the filtration system so long as the eff,1uents contain only trace quantities of radioactive =aterial. 3.9.2.2 Prior to release to the Keowee Hydro dam ta11 race, two independent samples of liquid wastes shall be analyzed to determine gross beta-- ga=ma and/or isotopic activity concentrations. 3.9.2.3 Radioactive liquid vaste shall be continuously monitored for flow and gross radioactivity concentration during release. 3.9.2.4 Wherein practicable, radioactive waste releases shall be coordi-nated with normal operation of the Keowee Hydro Station. L 3.9-1 L
3.9.2.5. The release of radioactive liquid effluents shall be such that the concentration of radionuclides during discharge to tne Keowee Hydro tailrace (based on a minimum dilution flow of s,/ 30 CFS or the available flow from the Keovee Hydro 3tation if ooerating) does not exceed the limits of 20.106 of 10 CFR 20, Appendix B. for unrestricted areas. 3.9.2.6. Under unusdal conditicas when waste treat =ent equipment is inoperable , liquid waste shall be held up for decay for the maximum period practicable prior to release and every reasonable ef fort shall be made to return any such disabled equipment to its operable conditions before releases are made. 3.9.2.7. If the limits of Specification 3.9.2.5 cannot be =et, radioactive liquid ef fluents shall not be released. 3.9.2.8. The continuous water sa=pler in the Keowee River near the site boundary shall be operable during discharge of liquid waste to the Keowee Hydro dam tailrace. If it is inoperable and the waste cannot be held up any longer, grab sa=ples shall be taken 1n the Keowee River where the continuous sa:pler is ]^cated in order to deternine the concentrations of radioactivity in the river during the batch discharge cperation. Refer to Table 4.11-1 for sample frequancy and analyais requiranants. 3.9.3 Gaseous Wastes 3.9.3.1. The annual average release rates of gaseous and airborne particulate s s/ wastes shal: Sm 11=ited in accordance with the following equation: [4.61 x 10-6 sec/m3 x1 Qi/(MPC)i] I 1.0 where Qi is the annual release rate in C1/sec of any radio-nuclide in the gaseous wastes mixtures; (MPC)1 is the permissible concentration for unrestricted areas in units of Ci/m3 (uci/cc= C1/m 3 ) for any radionuclide taken from Appendix B, Column 1, Table II of 10 CFR 20; and 4.61 x 10-6 sec/m3 is the long tena atmospheric dispersion factor (X/Q), at the exclusion area boundary of 1609 =eters for a ground level release. 3.9.3.2. For purpose of calculating permissible releases by the formula in 3.9.3.1, MPC for halogens and particulates with half-lives longer than eight days shall be reduced by a factor of 700 from their listed values in 10 CFR 20, Appendix 3, Colu=n 1, Table II. l l i
~/ \
i 3.9-2 1
I 3.9.3.3. During release of radioactive gaseous wastes from the gaseous vaste disposal syste= to the unit vent, the following conditions g,, shall be met:
- a. The gaseous radioactivity monitor, iodine conitor, and the particulate monitor in the unit vent shall be operable.
- b. The waste gases and particulates shall be passed through the high efficiency particulate filters and charcoal filters provided (except as noted in 3.9.3.9) 3.9.3.4 Purging of the reactor building shal.~. be governed by the following conditions:
- a. Reactor building purge shall be through the high efficiency particulate filters and charcoal filters until the activity concentration is below the occupational limit inside the re-actor building, at which ti=e bypass may be initiated,
- b. Reactor building purge shall be through the high efficiency particulate filters and charcoal filters whenever irradiated fuel is being hcndled or any objects are being handled over irradiated fuel in the Reactor Building.
- c. The limits of 3.9.3.4a and 3.9.3.4.b shall be met or the containment shall not be purged.
3.9.3.5 During power operation, whenever the air ejector of f-gas monitor
- s. , is inoperable, grab samples shall be taken from the air ejector discharge and analyzed for gross radioactivity daily.
3.9.3.6 Potentially highly radioactive gaseous waste from the gaseous waste disposal syste= and vent headers of unit co=ponents shall be provided a mini =um holdup of thirty days (except as noted in 3.9.3.7) when the release of the gaseous waste would exceed 1% of 10CFR20 limits as determined by the following equation: [4.61 x 10-6 sec/m3 x Qi/(MPC)1] 1 0.01 3.9.3.7. Under unusual conditions, gaseous waste may be discharged from the waste gas header directly to the unit vent for a period not to exceed seven days if the holdup system equipment is not available and the releases meet Specification 3.9.3.1 and 3.9.3.3. Every reason-able effort shall be =ade to re-establish the availability of the holdup system equip =ent. 3.9.3.8. Under unusual conditions, gaseous radioactive waste from the gaseous waste disposal system can be released - to the unit vent without passing through the HEPA and charcoal filters if the filter system is inoperable. This = ode of release may not exceed a seven-day period, and the release rate must be within the li=its of Specification 3.9.3.1 and 3.9.3.3a. v 3.9-3 Rev. 19. 5/5/7.*
3.9.3.9. Under unusual conditions, when the filter system is inoperable, ,_ f gaseous wastes shall be held up for the max 1=um period practiccble prior te release. Every reasonable effort shall be made to return inoperable filters to the operable condition before releases to the environment are made. 3.9.3.10 The maximum activity to be centained in one gas holdup tank shall be limited to 17,200 Ci/E. E will be assumed to be the same as the E of the noble gases in the reactor coolant system as determined in accordance with Table 4.,,1-3,o,f Specification 4.1.2. 3.9.3.11 Gaseous waste releases shall be restricted so as to yield con-centrations in the area of the temporary constructicn workers' quarters in the east-southeast section of the exclusion area that are no greater than that which could exist at the normal 1609 meter exclusion area boundary. Bases Waste processing equipment shall be maintained and used in accordance with 10CFR50.36a and administrative procedures developed in accordance with 10CFR50.34a to assure that releases of radioactivity will be maintained as icw as practicable with the intent to be less than 1% of the limits of 10C7220 Appendix 3 Table 2 Column II af ter dilutien ylth the total hydro , flow occurring during discharge of liquid radioactive wast . Provision is made for flexibility of operation compatible with health and safety to be sure that the public is provided a dependable source of power. Even under %.s/ unusual operating conditions, which may temporarily result .in releases higher than usual, the limits of 10CFR20 will still be maintained. It is not intended that this waste be continually reprocessed. Such reprocessing would be the prerogative of the licensee. Unusual operations, as used in these specifications, a're those conditions existing when not all processing equipment is operable. A. Liquid Wastes Radioactive liquid wastes will be collected in waste storage tanks. Treatment of liquid wastes for recovery of the water by evaporation and/or ion exchange and disposal of concentrated evaporator bottoms and spent resin as solid wastes will be perfor=ed to maintain quantities of radioactive materials released as low as practicable. Contents of the low activity waste tank and the condensate test tank will be mixed and sampled for analysis to deter =ine the resulting concentration upon dilution. The minimum dilution flow without Keowee Hydro Station operation is 30 cfs and will be periodically verified. It is intended where practicable, to coordinate radioactive liquid waste releases with the operation of the Keowee Hydro Station to provide dilution flows greater than 30 cis. However, calculations to determine batch discharge rates will be made on the basis of a 30 cfs dilution flow s ,/ or the available flow from the Keovee Hydro Station if operating. 3.9-4 Rev. 19 5/5/72 (Carry over)
\'
Inputs to the low activity waste tank are expected to contain only trace quantities of radioactivity and consequently may be released directly to the Keowee tailrace without further processing than the filters associated with the tank. In the event that significant activity, on the order of 2 x 10-8 pCi/ml, is found in the low activity waste tank, the contents will be processed either by evaporators and/or demineralizers. High activity wastes drain to either the miscellaneous waste holdup tank or the high activity waste tank. The specification regarding returning inoperable equip =ent to the operabic condition is intended to preclude unnecessary delays in recovering from unusual operating conditions in conformance with i 10 CFR 50.36a. B. Gaseous Wastes Radioactive gases will be those resulting from the fission process and neutron activation. These gases will be collected in the waste gas tanks from the various liquid storage tanks associated with the reactor, Gaseous wastes in the reactor building atmosphere are released by purging to the unit vent. Any gaseous vastes in the Auxiliary Building atmosphere will be released through the ventilation system to the unit vent which is =onitored. L Temporary construction quarters are located inside the exclusion area. During the period of time that these quarters are in 2se, administrative procedures will limit releases of gaseous waste in this sector of the exclusion area in accordance with appropriate meteorological restrictions. In addition to this, these quarters will be monitored to assure that the administrative procedures are effective and that the dose limits prescribed by regulations are not exceeded. > The long ters atmospheric dilution factor (X/Q) at the exclusion area boundary of 1609 meters for a ground release as used in this specification incorporates results from SF6 (sulfur hexafluoride) gas tracer experiments (see Oconee FSAR Appendix 2A) and includes a 0.53 dilution factor for inversion occurrences at low wind speeds. The (y/Q) value to be used under Gaseous Wastes in Specification 3.9, as 4.61 x 10-6. High concentrations of airborne radioactivity are not expected in the containment unless the reactor has significant failed fuel and/or there is significant uncollected primary coolant leakage in the containment. In order to reduce the amount of radioactivity'which would be purged to the atmosphere, the containment air will be exhausted v 3.9-5 Rev. 19 5/5/72 (Carry over)
J I I
~
through h'igh cfficiency particulate and charcoal filters until con-tinuous occupancy occupational exposure limits are attained in the containment. After these limits are attained, the maximum concentration of radioactivity at the site boundary would be approximately 0.12% of 10 CFR 20 Limits for unrestricted areas assuming the restrictive at-mosphere diffusion factor applied to the accident analysis (1.16 x 10-4
- sec/m). However, contain=ent purging would normally be timed to coin-cide with better atmospheric dispersion conditions. It is considered reasonable to allow by-passing of the purge filters und.er these con-ditions since with small concentrations of radioactivity little, if any, discernable health benefit would be achieved by using the filters while the useful life of the filters would be consumed. In addition, it is anticipated that the containment will be purged only on a periodic basis when personnel are required in the containment.
i J J 3.9-6 Rev. 19 5/5/72 (Carry over)
e . I 3.10 MAXIML,M POWER RESTRICTION
\-- Applicability a
Applies to the nuclear steam supply system of Unit I reactor. Objective To maintain a power margin in reserve until system performance has performed under operating conditions and design objectives have been verified. Specification Unit I power level may not be increased above 2452 MW; until operated in the
- range of 2352 to 2452 MWT for 30 days, except that 50 percent of the time the
- power can be as low as 2,000 MW T , and subsequent approval is granted by the AEC/DRL Regulatory Staff.
Bases I The Preliminary Safety Analysis Report section of the application for a con-struction permit was based on a maximum power level of 2452 MW7. Subsequent
- af ety evaluations done as part of the Final Safety Analysis Report were done for power levels of 2568 MWT. However, since this is the first nuclear steam 4 apply of this design to go into service, a power margin of 116 MW is tem-porarily being held in reserve until the system has performed at significant power levels for a reasonable period of time. Following evaluation of the i
s_, summary report of plant startup and power escalation test programs and , evaluations, (required by these Technical Specifications), ar.d in.the absence 3 of any significant deviation in plant performance from that predicted by design and required for safety, it is expected that this comporary restriction will be lifted. I 4 i a I
. s_. ;
l 3.10-1
)
I i
, e -
- 19. 3.11 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST b
Applicability Applies to the use of the reactor building polar crane over the stea= generator compartments and the fuel transfer canal and the Auxiliary Hoist over the fuel transfer canal. Objective To identify those conditions for which the operation of the reactor building polar crane and auxiliary hoist are restricted. 4 Specification
- 19. 3.11.1 The reactor building polar crane shall not be operated over the fuel transfer canal when any fuel assembly is being moved.
(
, 19. 3.11.2 The auxiliary hoist shall not be operated over the fuel transfer ! canal when any fuel assembly is being moved unless the hoist is 3
being used to move that assembly. l
- 19. 3.11.3 During the period when the reactor vessel head is re=oved and irradiated fuel is in the reactor building and fuel is not being j moved, the reactor building polar crane and auxiliary hoist
; shall Se operated over the fuel transfer canal' only where j s,, necessary cad in accordance with approved operating procedures 4
stating the purpose of such use.
- 19. 3.11.4 When the reactor vessel head is removed and the polar crane is being operated in areas away from the fuel transfer canal, the flagman shall be located on top of the secondary shield wall when the polar crane hook is above the alevation of the fuel transfer canal.
3.11.5 19, During the period when the reactor coolant system is pressurized above 300 psig, and is above 200 F, and fuel is in the core, the reactor building polar crane shall not be operated over the steam generator compartments. J Bases Restriction of use of the reactor building polar crane and auxiliary hoist over the fuel transfer canal when the reactor vessel head is removed to those operations necessary for the fuel handling and core internals opera-tions is to preclude the dropping of materials or equipment into the reactor vessel and possibly damaging the fuel to the extent that an escape of fission products would result. The fuel transfer canal will be delineated by readily visible markers at an elevation above which the reactor building polar crane would not normally handle loads. g,, Restriction of use of the reactor building polar crane over the steam gene-rator compartments during the time when steam could be formed from dropping a load on the steam generator or reactor coolant piping resulting in rupture of the system is required to protect against a less of coolant accident. 19' 3.,1-1 - Rev. 19 5/5/72
9
- 3.12 SECONDARY SYSTEM ACTIVITY
\-- Applicability Applies to the limiting conditions of secondary system activity for operation of the reactor.
Objective To limit the =axi=um secondary system activity. Specification The iodine-131 activity in the secondary side of a steam generator shall not exceed 1.4 pCi/cc. Bases For the purpose of derermining a maxicum allowable secondary coolant activity, the activity contained in the mass released following a loss of load accident , is considered. As stated in FSAR Section 14.1.2.8.2, 148,000 pounds of water is released to the atmosphere via the relief valves. A site boundary doae limit of 1.5 re= is used. The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the at=osphere by the condenser air ejector, thus, in the event of a less of load incident there are only s=all quantities of theta gases which would be released, i i 1-1311s the significant isotope because of its low M?C in air and because the other iodine isotopes have shorter half-lives, and therefore, cannot build up to significant concentrations in the secondary coolant, given the limitations on primary system leak rate and technical specification limiting activity. One-tenth of the contained iodine is assumed to reach the site boundary, making allowance for plateout and retention in water droplets. I-131 is assumed to contribute 70% of the total thyroid dose based on the ratio of I-131 to the total iodine isotopes given in Table 11-3 of the FSAR. The maximum inhalation dose at the site boundary is then as follows: Dose (rem) = C1' V B DCF -(0.1) X/Q C = Secondary coolant activity (2.0 uCi/cc I-131 equiva' lent) V = Secondary water volume released to at=osphere (90 m3) B = Breathing rate (3.47 x 10-4 m /3 set) X/Q = Cround level release dispersion factor (1.16 x 10-4 sec/m3) DCF = 1.48 x 106 rem /Ci 0.1 = Fraction of activity released The resultant dose is 1.15 rem compared to the Radiation Protection Guide of 1.5 rem for an annucl individual exposure in an unrestricted area. V 3.12-1
._ _ - ._ ~.
s o 4 SURVEIM.A5CE' STANDARDS v Specified intervals may be adjusted plus or minus 25% to accommodate normal nest schedules. 4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation. Objective To specify the minimum f requency and type of surveillance to be applied to unit equipment and conditions. Spe c i fi ca tion 4.1.1 The minimum frequency and type of surveillance required for reactor protective system and engineered safety feature protective system in-strumentation when the reactor is critical shall be as stated in Table 4.1-1, 4.1.2 Eq uipmen t and sampling test shall be perforr.ed as detailed in Tables 4.1-2 and 4.1-3. s. itases Ch e d Fa i l u res such as blown instrument fuses , defective indicators , faulted ampli fiers which result in " upscale" or "downscale" indication can be
.asily recognized by simple observation of the functioning of an instrument or system. Furthe rmore , su.h failures are, in many cases , revealed by alarm or annuniciator actiun. Comparison of output and/or state of independent channels measuring the same variable supplements this type of buil t-in surveillance. Based on experience in operation of both conventional and nuclear systecs, when the unit is in operation, the minimum checking f requency stated is deemed adequate for reactor sys tem ins trumentation.
Cal i b ra t i on Calibration shall be performed to assure the presentation and acquisition of accurate in fo rma tion. The nuclear flux (power range) channels amplifiers shall be calibrated (during steady state operating conditions)when indicated neutron power
- 18. and core ther=al power differ by more than 2 percent. During non-steady state l operation, the nuclear flux channels amplifiers shall be calibrated daily to j 4.1-1 Rev. 15 3/10/72s.
, _ - . __ - ~
t . compensate for instrumentation drift and changing rod patterns and core physics i parameters. ,,j Channels subject only to " drift" errors induced within the instru-mentation i tself can tolerate longer intervals between calthrations. Process system instrumentation errors induced by /*tft can be
- expected to remain within acceptable tolerances if recalibration is performed at the intervals of each refueling period.
Subatantial calibration shif ts within a channel (essentially a channel f ailure) will be revealed during routine checking and testing procedures. 4 Thus, minimum calibration frequencies set forth are considered acceptable. Testing On-line testing of reactor protective channels is required once every four l weeks on a rotational or perfectly staggered basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within j the system ano to minimize the likelihood of the sa=e systematic test errors being introduced into each redundant channel. l The rot.4 tion schedule for the reactor protective channels is as follows: Channels A, B, C, &D Before Startup Channel A One Week After Startup '"# Channel B Two Weeks After Startup Channel C Three Weeks After Startup 1 j Channel D Four Weeks After Startup
, The reactor protective system instrumentation test cycle is continued with one 1
- channel's instrumentation tested each week. Upon detection of a failure that 1 prevents trip action, all instrumentation associated with the protective channels l will be tested after which the rotational test cycle is started again. If
] actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting. < I 4
; The protective channels coincidence logic and control rod drive trip breakers are trip tested every four weeks. The trip test checks all logic combinations and is to be performed on a rotational basis. The logic and breakers of the four protective channels shall be trip tested prior to startup and their l Individual channels trip tested on a cyclic basis. Discovery of an unsafe i tailure requires the testing of all channel logic and breakers, after which the trip test cycle is started again.
Tiie equipment testing and system sampling frequencies specified in Table 4.1-2 and Tabic 4.1-3 are considered adequate to maintain the equipment and ! systems in a safe operational status. '"# 4 { 4.1-2
t
- REFERENCE b FSAR Section 7.1.2.3.4 f
l l t I i t 4 4.1-3 I.
-s-4,
~
f O TABLE 4.1-1 INSTROENT SURVEILLANCE REQUIRESENIS _C}{ANNEL DESCRI PTION Ci!ECK IEST CAtl BluTE REMARKS
- 1. Pretective Channel NA M NA Coincidence Logic
- 2. Cont ro1 Rod Drive NA M NA Trip Breaker
- 3. Power Range Amplifier D(1) NA (1) (1) !! eat Balance Check daily. Ileat balance calibration whenever indicated neutron power & core thermal power dif fer by more than 2%.
- 4. Power Range Channel S M M(1)(2) (1) Using incore instrumentation.
(2) Axial offsct upper & lower chambers af ter o each startup if not done previous week.
; 5. Intermediate Range Channel S(l) P NA (1) When in service E~
- 6. Source Range Channel S(l) P NA (1) When in service
- 7. Reactor Coolant Temperature S M R Lhannel
- 8. Ill Eh Reactor Cool. nt S M R Pressure Channel
- 9. Low Reactor Coolant S M R Pressure Channel 10 Flux-Reactor Coolant Flow S M R Comparator
- 11. Reactor Coolant Pressure S M R Ten;pe ra t u re Compa rat o r
- 12. Pump Flux Comparator S M R
_. __......._m-_._..___m.._ _ _ _ . - _ . - . - . . . _ . . . _ . _ _ _ . . _ . . _ - - _ ( ( TABLE 4.1-1 Cont. OIANNEL DESCRIPTION GIECK TEST CEIBRATE REMARKS
- 13. !!igh Reacte r Building Pressure Channel .D M R
- 14. liigh Pressure Injection Logic Channel NA M NA
- 15. liigh Pressure Injection Analog-ChanneIs
- a. Reactor Coolant Pressure Channel S M R
- b. Reactor Building 4 psig Channel S M R s~
l 7 16. l.ow Pressure Injection p
- v. Logic Channel NA B/M NA lh. Low Pressure Injection Analog Channels
- a. Reactor Coolant S M R Pressure Channel
- b. Reactor Building S M R 4 psig Channel
- 18. Reactor Building Emergency Cooling and Isolation -
gn System Logic Channel NA B/M NA l [ 19. Reactor Building Emergency Cooling and Isolation l System Analog Channels l l a. Reactor Building l 4 psig Channels S M R l l 1
. - , . - - _ - - - - . - _ . . ~ . _ - - - - _ - . - - _ - _ _ - . . . . _ _ . .- .- ~ _ . ~ _ _ . _ . . - _._ . . .
4 IABLE '. 1-1 Cont. ' CilANNEI DES CRIP TION CllECK TEST _CAlfBRATE REFL\RKS
- 20. Reactor Building Spray
, Systen Logic Channel NA B />f NA
- 21. Reactor Building Spray System Analog Channels
- a. Reactor Building 10 psig Osannels ~ NA M R
- 22. Pressurizer Temperature Channels S NA R
- 23. Control Rod Absolute Position S(l) NA R (2) (1) Oieck with Relative Position Indicator 24.
(2) Calibrate rod misalignment channel 7 Control Rod Relative o' Position S(1) NA R (2) (1) Qieck with Absolute Position Indicator 25. (2) Calibrate rod misalignment channel Core Flooding Tanks o l a. Pressure Channels S NA R
- b. level Channels S NA R
- 26. Pressurizer Level Channels S NA R i
- 27. I.etdown Storage Tank
,4 1.evel Channels D NA R
- 28. Radiat ion Monitoring Systems W(1) M Q (1) Check functioning of self-checking feature on each detector.
- 29. Iligh and Low Pressure Injection-Systems: Flow Channels NA NA R 1
i
( ( ( TABLE 4.1-1 Cont. CilANNEI. DESCRIPTION OIECK TEST CALIBRATE REMARKS
- 30. Borated Water Storage Tank level Indicator W NA R
- 31. Boric Acid Mix Tank
- a. Level Channel NA NA R
- b. Temperature Channel M NA R
- 32. Concentrated Boric Acid Storage Tank
- a. Imvel Otannel NA NA R
- b. Temperature Channel M NA R
*~
yw
- 33. Containment Temperature NA NA R
- 34. Incore Neutron Detectors M(1) NA NA (1) Check functioning ; including functionic of computer readout or recorder readout
- 35. Emergency Plant Radiation Instrumenta M(1) NA R (1) Battery Check
- 36. Environmental Monitors M(1) NA R (1) Check functioning
- 37. Strong Motion Accelerometer Q(1) NA Q (1) Battery Check
- 38. Reactor Building Emerg. NA NA I4 Sump Level
- 39. Steam Generator Water
, Level W NA R e
- 40. . Turbine overspeed Trip R
*H 19
- Engineered Safeguards NA R NA 41.
w Channel 1 !!P Injection 3 Manual Trip D~
TARLE 4.1-1 Cont. CitANNEL DESCRIPTION CHECK TEST CAI.IBRATE REMARKS
- 42. Engineered Safeguards NA R NA Channel 2 HP Injection Manual Trip
- 43. Engineered Safeguards NA R NA Channel 3 LP Injection Manual Trip
- 44. Engineered Safeguards NA R NA Channel 4 LP Injection Manual Trip
- 45. Engineered Safeguards NA R NA Channel 5 RB Isolation
& Cooling Manual Trip
. 46. Engineered Safeguards NA R NA T' Channel 6 RB Isolation 7 & Cooling Manual Trip
- 47. Engineered Safeguards NA R NA Channel 7 Spray Manual Trip
- 48. Engineered Safeguards NA R NA Channel 8 Spray Manual Trip
- 49. Reactor Manual Trip NA R I NA to S - Each Shift T/W - Twice per week R - Each Refueling Period u -
n4
- p. D - Daily B/M - Every 2 months NA- Not Applicable
< r, g* W - Weekly Q - Quarterly B/W- Every two weeks h M - Monthly P - Prior to each startup if not donc previous week DV3
. n 1
Table 4.1-2 Minimum Eculement Test Frequenev Item Test Frecuency
- 1. Control Rods Rod Drop Times of all Each Refueling shutdown full length rods
- 2. Control Rod Movement Movement of each rod Every two weeks
- 3. Pressurizer Safety Valves Setpoint 50% each refueling period
- 4. Main Steam Safety Valves Setpoint 25% each refueling period
- 5. Refueling System Functional Each refueling period In te rlocks
- 6. Turbine Steam Step Valves Moverent of each stop Monthly valve
- 7. Reactor Coolant System Evaluate Daily Leakage
- 3. Ciarcoal and High DOP Test on IiE?A Each refueling period Efficienty Filters for filters. Freon Test and at any time work Penetration Room, Control on Charcoal Filter n filters could alter s Room, and R3 Purge Filters Units their integrity.
- 9. Emergency Condenser Cooling Functional Each refueling period i Water System I
- 10. High Pressure Service Functional Monthly Water Pumps and Power Supplies
- 11. Spent Fuel Cooling System Functional Each refueling period prior to fuel handling I
4.1-8
4 TABLE 4.1-3 Minimus Sampling Frecuency item Check Frecuency
- 1. Reactor Coolant a. Radio-Chemical Anal Monthly E determinatien (2)ysis(1) Semiannually
- b. Gross Beta-Ga=ma Activity (l) 5 times / week 3 c. Tritium Radioactivity Monthly i
- d. Chemistry (Cl, F and 02) 5 times / week
- e. Boron Concentration 2 ti=es/ week
]
- 2. Borated Water Storage Boron Concentration Weekly and after
; Tank Water Sample each makeup
) 3. Core Flooding Tank Boron Concentration Monthly and after Vater Sample each makeup
, 4. Spent Fuel Pool Water Beren Concentration Monthly and after . S ample each makeup
- 5. Secondary Coolant a. Gross Beta-Gamma Activity Weekly
- b. Iodine Analysisl3)
- 6. Concentrated Boric Boron Concentration Twice weekly Acid Tank I
- 7. I.ow Activity Waste a. Gross Beta-Ga=ma Activity Prior to release
- Tank & Condensate of each batch
! Tes t Tank b. Isotopic Analysis Quarterly I c. Gamma Scan Monthly
- d. Tritium quarterly H. Waste Gas Decay Tank a. Isotopic Analysis Quarterly l
- b. Cross Beta-Gamma Activity Prior to release of each I batch i 1
- 9. Unit Vent Sampling a. Iodine Spectrum (4) Weekly System b. Particulates (4) Weekly i 10. Keowee Hydro Dam a. Measure Leakage Flow Rate Annually j
- Dilution Flow '
- 11. Condenser Air a. Measure Iodine Partition One time if and when i
Partition Factor Factor in Condenser primary to secondary leak develops 1 l 4.1-9
- - , --. . - . . - . - _ - . _ , . . . . . . _ _ , . . - - _ - . . . . , - - m . ,
o . (1) When radioactivity level is greater than 10 percent of the limits of Specification 3.1.4, the sampling frequency shall be increased to a _,f minimum of once cach day. (7) K determination will be started when gross beta-gamma activity analysis Indicates greater than 10 uC1/ml and will be redetermined each 10 aci/ml increase in gross beta-gamma activity analysis. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95% of radionuclides in reactor coolant with half lives of >30 minutes. (3) When gross activity increases by a factor of two above background, an todine analysis will be made and performed thereafter when the gross beta-gamma activity increases by 10 percent. (') When activity level exceeds 10 percent of the limits of Specification 1.9. the sampling frequency shall be increased to a minimum of once each day. This can be done by 1RIA-44 (Unit Vent Iodina) monitor. When the gross activity release rate exceeds one percent of maximum release rate and the average gross activity release rate increases by
'20 percent over the previous day, an analysis shall be performed for lodines and particulates. This can be done by 1RIA-44 (Unit Vent fodine Monitor) and RIA-43 (Unit Vent Particulate Monitor).
V V 4.1-10
N-4.2 REACTOR COOLANT SYSTE'i SURVEILLANCE Appitcability Applies to the surveillance of the reactor coolant system pressure boundary. O,biective To assure the continued integrity of the reactor coolant system pressure boundary. Specification 3 4.2.1 Prior to initial unit operation, an ultrasonic test survey shall be made of reactor coolant system pressure boundary welds as required to establish preoperational integrity and base line data for future inspections. 4.2.2 Post operational inspections of components shall be made in accordance with the methods and intervals indicated in IS-242 and 15-261 of Section XI of the ASME Boiler & Pressure Vessel Code, 1970, including 1970 Winter addenda, except as follows: IS-261 Item Component Exception v I 1.4 Primary Nozzle to Vessel 1 RC outlet nozzle to be Welds inspected after approx.
< 31/3 years cperation.
2nd RC outlet nozzle to i be inspected after approx. j 6 2/3 yrs operation. 4 RC inlet nozzles and 2 core
! flooding nozzles to be in-spected at or near end ! of interval 3.3 Primary Nozzle to Safe End Not Applicable Welds 4.3 Valve Pressure Retaining Not Applicable Bolting Larger than 2" f
f j Nwe 3 2 4.2-1
- , . - . _ .. , . , . _ . ~ , , . - , - , . , ,
i 1 6.1 Valve Body Welds Not Applicable v 6.3 Valve to Safe End Welds Not Applicable 6.6 Integrally Welded Valve Supports Not Applicable 6.7 Valve Supports & hangers Not Applicable 4.2.3 The structural integrity of the reactor coolant system boundary shall be maintained at the level required by the original acceptance standards throughout the life of the station.Any evidence, as a result of the tests outlined in Table IS-261 of Section XI of the code, that defects have developed or grown, shall be investigated, in- - cluding evaluation of comparable areas of the ra2ctor ecolant system. 4.2.4 To assure the structural integrity of the reactor internals throughout the life of the unit, the two sets of =ain internals bolts (connecting the core barrel to the core support shield and to the lower grid cylinder) 3 hall remain in place and under tension. This will be verified by visual inspection to determine that the welded bolt : locking caps remain in place. All locking caps will be inspected after hot functional testing and whenever the internals are removed from the vessel during a refueling or maintenance shutdewn. The core barrel to ccre support shiald caps will be inspected each refueling shutdown. 4.2.5 Sufficient records of each inspection shall be kept to allow comparison and evaluation of future inspections. 4.2.6 The inservice inspection program shall be reviewed at the end of f five years to consider incorporation of new inspection techniques and equipment which have been proven practical and the conclusions of this review and evalt.ation shall be discussed with the AEC/DRL. 4.2.7 The inspection of each reactor coolant pump flywheel shall include: a volumetric examination, in place, of the areas of higher stress concentration at the bore and key way at approximately three year intervals. A surface examination of exposed surfaces, and a complete , 18. volumetric examination at approximately 10 year intervals when dis-assembly and/or flywheel removal is required for maintenance or 1 repair. Disassembly or flywheel removal is not required to perform j these examinations.
- 4.2.8 Vessel specimen capsules shall be withdrawn at one, seven, twelve, and eighteen years. Withdrawal schedules may be modified to coincide with those refueling outages or unit shutdowns most closely approaching the withdrawal schedule. Specimens thus a withdrawn shall be tested in accordance with ASTM-E-185-70. A l copy of the test report shall be forwarded to DRL within 9 0 days.
4.2.9 During the first two refueling periods, two reactor coolant system piping cibows shall be ultrasonically inspected along their longitudinal welds (4" beyond each side) for clad bonding and for '/ cracks in both the clad and base metal . The elbows to be inspected Rev. 18 3/10/72
are numbers 7 and 9 in assemblies A57 and 357 respectively as identified in B&W Report 1364 dated December 1970. s-Bases The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor (:oolant Systems, 1970, including 1970 Winter addenda, edition. The program places major emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties. The vessel specimen surveillance program is based on equivalent exposure years of I.76, 11.18, 20.41, and 30.0. The 1.76 year number is based on a 50 F shif t of NOTT; the 30-year figure is based on 3/4 vessel service life; and the 11.18 ami the 20.41 year figures are equally spaced between. Early inspection of reactor coolant system piping elbows is considered desirable lo order to reconfirm the integrity of the carbon steel base metal when ex-plonively clad with sensitized stainless steel. If no degradation is observed during the two annual inspections, surveillance requirements will revert to Sect inn XI of the ASME Boiler and Pressure Vessel Code. v i i q a 1 3
;i 4.2-3 i
n i 1 s 4.3 TESTING FOLLOWING OPENING OF SYSTEM e Applicabilitv Applies to test requirements for Reactor Coolant System integrity. Obiective To assure Reactor Coolant System integrity prior to return to criticality following normal opening, codification, or repair. Soecification j 4.3.1 When Reactor Coolant System repairs or codificatiens have been 4 cade, these repairs or nodifications shall be inspected and tested to meet all applicable code requirements prior to the reactor being made critical. 4.3.2 Following any opening of the Reactor Coolant System, it shall be leak tested at not less than 2285 psig prior to the reactor being made critical. 4.3.3 The limitations of Specification 3.1.2 shall apply. g,, 33908 Repairs or modifications made to the Raactor Coolant Syste: are inspectable and testable under applicable codes, such as 3 31.7, and ASME Boiler and 18 lPressureVesselCode,SectionXIIS-400. ja 19.f l REFERENCES FSAR, Section 4 I s%, { 4.3-1 Rev. 18. 3/10/72 Rev. 19. 5/5/72 f 6
- - . - , , - w . . .~ - + -4, -, r , , , , vw - e --- , =-
e 4.4 REACTOR BUILDING 4.4.1 Containment Leakage Tests Aonlicability Applies to containment leakage. Obj ect ive To verify that leakage from the reactor building is maintained within allowable limits. Specification
; 4.4.1.1 Integrated Leakage Rate Tests 4.4.1.1.1 Design Pressure Leakage Rate The maximum allowable integrated leakage rate, La, from the reactor building at the 59 psig design pressure, Pp, shall not exceed 0.25 weight percent of the building atmosphere at that pressure per 24 hours.
4 4.4.1.1.2 Testing at Reduced Pressure 1 s_ The periodic integrated leak rate test may be performed at a test pressure, Pt, of not less than 29.5 psig provided the resultant leakage rate, Lt, does not exceed a pre-established fraction of La determined as follows:
- a. Prior to reactor operation the initial value of the integrated leakage rate of the reactor building shall be measured at de-sign pressure and at the reduced pressure to be used in the periodic integrated leakage rate tests. The leakage rates thus measured shall be identified as Lps-and L 2 respectively.
- b. Le shall not exceed La L tm below 0.7 T for values of Lem L pm Lm p c.
Le shall not exceed La Fjg for values of L tm above 0.7.
} Pp hm
- d. If Lem/ Lpm is less than 0.3, the initial integrated test results shall be subject to review by the AEC to establish an acceptable value of Lt.
v 4 4.4-1
4.4.1.1.3 Conduct of Tests (a) The test duration shall be at least 24 hours, except that if s./ both the following conditions are met, the test duration shall be at least 10 hours. T 18. (1) All test conditions, including the test procedure, shall be similar to the initial integrated leakage rate tests. I ! (2) When the test is terminated, building pressure shall have stabilized and shall not be increasing. (b) Test accuracy shall be verified by supplementary means, such as ! measuring the quantity of air required to return to the starting point or by imposing a known leak rate to demonstrate the validity of measurements. (c) Closure of containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal
- operation of the valves without preliminary exercises or j adjustment.
I 4.4.1.1.4 Frequency of Test 1 Af ter the initial preoperational leakage race test, two integrated t Icakage rate tests shall be performed at approximately equal inter-vals between each major shutdown for inservice inspection to be per-i formed at 10 year intervals. In addition, an integrated test shall s d' be performed at each 10 year interval, coinciding with the inservice inspection shutdown. The test shall coincide with a shutdown for i major fuel reloading. 4.4.1.1.5 Conditions for Return to Criticality
- a. If Lt is less than 50% of the value permitted in 4. 4.1.1. 2 , 1c cal leakage rate testing need not be completed prior to return to j criticality following a periodic integrated leakage rate test.
- b. If Lt is between 50 and 100% of the value permitted in 4.4.1.1.2 a
the return to criticality will be permitted conditioned upon demonstrating that local leakage rate measured at full design pressure, accounts for all leakage above 50% of L . If this t cannot be demonstrated within 30 days of returning to criti-cality, the reactor shall be shutdown. b I l 4.4.1.1.6 Corrective Action and Retest If repairs are necessary to meet the criteria of 4.4.1.1.1 or ! 4.4.1.1.2, the integrated leak rate test need not be repeated pro-vided local leakage rate measurements are made before and after repair to demonstrace that the leakage rate reduction achieved by %/ repairs reduces the overall measured integrated leak rate to an acceptable value. 4.4-2 Rev. 18 3/10/72
.= .-.
4.4.1.1.7 Report of Test Results Each integrated leak rate test will be the subject of a summary technical report which will include a description of test methods i used and a su= mary of local leak detection tests. Sufficient data and analysis shall be included to shew that a stabilized leak rate was attained and to identify all significant required correction factors such as those associated with humidity and barometric pressure, and all significant errors such as those associated with instrumentation sensitivities and data scatter. 4.4.1.2 f.ocal Leakage Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for each of the following components: (a) Personnel hatch (b) Emergency hatch (c) Equipment hatch seals (d) Fuel transfer tube seals (c) Reactor building normal sump drain line w (f) Reactor coolant pump seal outlet line (g) Reactor coolant pump seal inlet line (L) Quench Tank drain line (1) Quench Tank return line (j) Quench Tank vent line 4 (k) Normal makeup to reactor coolant system 1 (1) High pressure injection line. (m) Electrical penetrations (n) Reactor building purge inlet line (o) Reactor building purge outlet line (p) Reactor building sample lines (q) Reactor coolant letdown line 1 4.4-3
l 4.4.1.2.2 Conduct of Tests (a) Local leak rate tests shall be performed at a pressure of s/ not less than 59 psig. (b) Acceptabic methods of testing are halogen gas detection, soap bubbles, pressure decay, hydrostatic flow or equivalent. 4.4.1.2.3 Acceptance Criteria The total leakage from all penetrations and isolation valves shall not exceed 0.125% of the reactor building atmosphere per 24 hours. 4.4.1.2.4 Corrective Action and Retest (a) If at any time it is determined that the criterion of 4.4.1.2.3 above is exceeded, repairs shall be initiated immediately. (b) If conformance to the criterion of 4.4.1.2.3 is not demon-strated within 48 hours following detection of excessive local leakage, the reactor shall be shutdown and depressu-rized until repairs are effceted and the local leakage =cets the acceptance criterion as demonstrated by retest. 4.4.1.2.5 Tes t Frequency s,/ Local leak detection tests shall be performed at a frequency of at least each refueling period, except that: (a) The equipment hatch and fuel transfer tube seals shall
-be additionally tested after each opening.
(b) The personnel hatch and emergency hatch outer door seals shall be tested at six month intervals, except when the hatches are not opened during that interval. In no case shall the test interval be longer than 12 months.
- 4. 4.1. ) Isolation Valve Functional Tests Every three months, remotely operated reactor building isolation valves snall be stroked to the position required to fulfill their safety function unless such operation is not practical during plant operation. The latter valves shall be tested during each refueling period.
4.4.1.4' Annual Inspection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be performed annually and prior to any integrated leak test, to uncover any evidence of.deterioratien which may af fect either the s/ rontainment's structural integrity or leak-tightness. The dis-4.4 a _ ______-___:_______._-
~.
covery of any significant deterioration shall be accompanied by
- corrective actions in accord with acceptable procedures, non-destructive tests, and inspections, and local testing where prac-LIcal, prict to the conduct of any integrated leak test. Such repairs shall be reported as part of the test reaults.
4.4.1.5 Reactor Building Modifications Any major modification or replacement of components affecting the reactor building integrity shall be followed by either an inte-
- grated Icak rate test or a local leak test, as appropriate, and shall meet the acceptance criteria of 4.4.1.1.4 and 4.4.1.2.3 respectively.
. !!ases (l) The reactor building is designed for an internal pressure of 59 psig and a h steam-air mixture temperature of 286*F. Prior to initial operation, the con-i tainment will be strength tested at 115% of design pressure and leak rate
! tested at the design pressure. The containment will also be leak tested prior ! to initial operation at approx 1=ately 50% of the design pressure. These tests j will verify that the leakage rate from reactor building pressurization satifies the relationships given in the specification.
I rhe perturmance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the containment in case s,, of an accident that would pressurize the interior of the containment. In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic test is to be perfor=ed without preliminary leak detection surveys or leak repairs, and containment isolation valves are to be closed in the normal manner. The test pressure of 29.5 psig for the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate.and it duplicates the pre-operational j lenkage rate test at 29.5 psig. The specification provides a relationship for j relating the measured leakage of air at 29.5 psig to the pntential leakage at
; 59 psig. The minimum of 24 hours was specified for the integrated leakage ] rate test to help stabill:e conditions and thus improve accuracy and to better evaluate data scatter. The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests } can best be performed during refueling shutdowns.
J' { The specified frequency of periodic integrated leakage rate tests is based on three major considerations. First is the low probability of leaks in the liner, because of conformance of the complete containment to a 0.25% leakage 1 rate at 59 psig during pre-operational testing and the ebsence of any signifi-cant stresses in the liner during reactor operation. Second is the more f re-quent testing, at design pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.125%) of leakage that is specified as acceptable-from penetrations and isolation ~ valves. Third is the tendon stress surveillance program which provides assurance-that an important part of the. structural integrity of the containment is maintained. v 4.4-5
. e 1cakage to the penetration room, which 12 permitted to be up to 50% of the total allow.-ble containment leakage, !
, .'acharged through high efficiency particulate air (liEPA) and charcoal ccra to the unit vent. The filters are %/
conaervatively said to be 90% effic -t for iodine removal, a More frequent testing of vario>- ;enetrations is apecified as these locations a re mo re susceptible to leat1;e than the reactor building liner due to the mechanical closurn involver.. Particular attention is given to testing those penetrations with resi'.Aent sealing materials, penetrations that vent directly to the reactor bullding accosphere, and penetrations that connect to the re-actor coolant system pressure boundary. The basis for specifying a total leakage rate of 0.125% f rom penetrations and isolation valves is that at least one-half of the allowable integrated leakage rate will be from those sources. Valve oper..bility tests are specified to assure proper closure or opening of the reactor building isolation valves to provide for isolation of functioning of Engineered Safety Features systems. Valves will be stroked to the position required to fulfill their safety function unless it is established that such t e:.t Ing is not practical during operation. Valves that cannot be full-stroke tested will be part-stroke tested during operation and full-stroke tested during each normal refueling shutdown. RI;Fl.:4ENCI:S (1) FSAR, Sections 5 and 13. d v 4.4-o l
4 i 4.4.2 Structural Integrity Applicability i
- Applics to the structural integrity of the reactor building.
l Objective i t To define the inservice surveillance program for the reactor building.
, Specification j 4.4.2.1 Tendon Surveillance i
! For the initial surveillance program, covering the first five years i of operation, nine tendons shall be selected for periodic inspection l for symptoms of material deterioration or force reduction. The
- surveillance tendons shall consist of three horizontal tendons, one in each of three 120' sectors of the containment; three vertical tendons located at approximately 120* apart; and three dome tendons located approximately 120* apart. The following nine tendons have been selected as the surveillance tendons:
Dome 1D23 2D28 3D2s l s,, Horizsntcl 13H9 51H9 53H10 Vertical 23V14 45V16 61V16 I 4.4.2.1.1 Lift-Off. 1 i i i Lift-off readings shall be taken for all 9 surveillance tendons. 4.4.2.1.2 Wire Inspection and Testing . One surveillance tendon of each directional group shall be relaxed and one wire frc. cach relaxed tendon shall be removed as a sample and visually int .cted for corrosion or pitting. Tensile tests shall also be pecfor=ed on a minimum of three specimens taken from ' the ends and middle of each of the three wires. The specimens 4 shall be the maximum length acceptable for the test apparatus to be used and shall include areas representative of significant l corrosion or pitting. I' After *he wire removal, the tendons shall be retensioned to the stress level measured at the lift-off reading and then checked by a final lif t of f reading. s_ l 4.4-7 I t
Shoula the inapection of one of the wires reveal any significant corrosion (pitting or loss of area), further inspection of the ,j other two sets in that directional group will be made to determine the extent of the corrosion and i:s significance to the load-carrying capability of the structure. The sheathing filler will be sampled and inspected for changes in physical appearance. Wire samples shall be selected in such a manner that with the third inspection (end of fifth year) wires from all 9 surveillance tendons shall have been inspected and testad. 4 . 4 . . 2 Inspection Intervals and Reports The inspection intervals, measured rom the date of the initial structural test, shall be one year, 'ree years and every five years there.ifter or as modified based on ex,'rience. Tendon surveillance may he conducted during reactor operation provided design conditions renarding loss of adjacent tendons are satisfied at all times. A quantitative analytical report covering results of each inspection shall be submitted (required by Technical Specification 6.6.4.7.c) and shall especially address the follcwing conditions , should they develop: (1) Broken wires. (2) The force-time trend line for any tendon, when extrapolated, that extends beyond either the upper or lower bounds of the predicted design band. %/ ( 1) Unexpected changes in corrosien conditions or sheathing filler properties. 4 . i . . 3 I M Anchorage Concrete Surveillanca
- a. The end anchorages of the surveillance tendons and adjacent concrete surface will be inspu.ced. In addition, other locations tor surveillance will be determined by information obtained frem design calculations , pres t.cssing records , obse rvations , and de-formation measurements made during prestressing.
- b. I'h e inspection interval will be one-half year and one year after the operation of Unit 1 and will occur during the warmest and cold 1st part of the year.
- c. The laspections made shall include:
(1) Visual inspection of the end anchorage concrete exterior surfaces. (2) A determination of the temperatures of the liner plate area or containment interior surface in locations near the end anchorage concrete under surveillance. s/ 4.4-8
(3) Measurement of concrete temperatures at specific end anchorage concrete surf aces being inspected. (4) The mapping of the predominant visible concrete crack patterns. a (5) The measurement of the crack widths, by use of optical ccm-parators or wire feeler gauges. (6) The measurement of movements, if any, by use of de=ountable mechanical extensometers,
- d. The measurements and observations shall be compared with those to which prestressed structures have been subjected in normal and ab-normal load conditions and with those of preceding measurements and observations at the same location on the reactor containment.
- e. The acceptance criteria shall be as follows:
If the inspections determine that the conditions are favorable in comparison with experience and predictions, the close inspections will be terminated by the last of the inspections stated in the schedule and a report will be prepared which documents the find-ings and recommends the schedule for future inspections, if any. If the inspections detect sympt:ms of greater than normal cracking or movements, an immediate investigation will be made to determine the cause. s_ 4.4.2.4 Liner Plate Surveillance 4.4.2.4.1 The liner plate will be examined prior to the initial pressure test in accessible areas to determine the following:
- a. Location of areas which have inward deformations. The magni-tude of the invard deformations shall be measured and recorded.
These areas shall be permanently marked for future reference and the inward deformations shall be measured between the angle stilfeners which are on 15-inch centers. The measurements shall be accurate to + .01 inch. Temperature readings shall be obtained on both the liner plate and cutside containment wall at the locations where inward deformations occur.
- b. Locations of areas having strain concentrations by visual exami-nation with emphasis on the condition of the liner surface.
The location of these areas shall be recorded. 4.4.2.4.2 Shortly after the initial pressure test and approximately one year after initial start-up, a reexamination of the areas located in Section 4.4.2.4.1 shall be made. Measure =ents of the inward de-formations and observations of any strain concentrations shall be made. i 1 4.4-9 1 1
4.4.2.4.3 If the difference in the measured inward deformations exceeds 0.25 inch (for a particular location) and/or changes in strain concen-tration exist, an investigation shall be made. The investigation %/ will determine any necessary correctiva action. 4.4.2.4.4 The surveillance program shall be discontinued after the one year after initial start-up inspection if no corrective action was needed. I f corrective action is required, the frequency of inspection for a continued surveillance program shall be determined.
%'"?2 l' rov i s ions have been made for an in-service surveillance program, covering the Itrst live years of the life of the unit, intended to provide sufficient evidence to m iintain confidence that the integrity of the reactor building is being pre-
- crved. This program consists of tendon, tendon anchorage and liner plate sur-veillance, fo accomplish these programs, the following representative tenden groups have been selected for surveillance:
!!orizontal - Three 120* tendons comprising one complete hoop system below grade.
Vertical - Three tendons spaced approximately 120* apart. Dome - Three tendons spaced approximately 120' apart. lho inspectfon Juring this initial five year period of at Icast one wire from c.ich of the nine surveillance tendons (three wires per group per inspection)
- 12. con,idered sut'ficient representation to detect the presence of any wide
.p re.nl teniton corrosion or pitting conditions in the structure. This program si!! Le sub ice t to review and revision as wi ranted based on studies and on ,
r esu l t.s ob tai ned fo r t: As and other prestressed concrete reactor buildings I during this period of time. j 1 Ri:FliRI.SCE j l EsAR Section 5.6.' ) ' 1 l l l l l i l 4.4-10 t- __ _ _ _ _ _ _ _ _ _ . - _ _ _ __ ___ _ _ -. __ __ __
1 , . d 4.4.3 Hvdrogen Purge System Applicability Applies to testing Ret -tor Building Hydrogen Purge System.
- Objective l
To verify that this system and components are operable. Snecification 4.4.3.1 Operating Tests An in-place system test shall be performed during each refueling period using the written emergency procedures. These tests shall consist of visual inspection, hook-up of systa=, a flow measurement using flow instruments in the portable purging station and pressure drop measurements across the filter bank. Flow shall be design flow or higher, and pressure drops across the filter bank shall not ex-ceed two times the pressure drop when new. Fan motors shall be operated continuously for at least one hour, and valves shall be proven operable. This test shall demonstrate that under simulated emergency conditions the system can be taken from storage and placed into operation within 48 heurs. 4.4.3.2 Filter Tests Daring each refueling period, leakage tests using DOP on HEPA units and Freon-ll2 (or equivalent) on charcoal units shall be performed at design flow on the filter. Removal of 99.5% DOP by each entire HEPA filter unit and removal of 99.0% Freon-112 (or equivalent) by each entire charcoal absorber unit shall constitute acceptable performance. These tests must also be performed af ter any main-tenance which may affect the structural integrity of either the filtration system units or of the housing. 4.4.3.3 H2 Detector Test l' Hydrogen concentration instruments shall be calibrated each refueling period with proper consideration to moisture effect. I i Bases l j The purge system is composed of a portable pur,ing station and a portion of the l penetration room ventilation system. The purge system is operated as necessary to maintain the hydrogen concentration below the control limit. The purge dis- ) charge from the Reactor Building is taken from one of the penetration room ventilation system penetrations and discharged to the unit vent. A suction may be taken on the Reactor Building via isolation valve PR-7 (Figure 6-5 of the 18- FSAR) using the existing vent and pressurization connections. s_ The purge rate is controlled through the use of a portable purging station (Insert, Figure 14A-5.1 of the FS AR) . The station consists of a purge blower, uehumidifier, filter train, purge flowr ter, sample connection and flowmeter and associated piping and valvesp j/ l. 4.4-11 ' Rev. 18 3/10/72 i 1
-,o - , . . - , - - - . . ~ -- 4 , , _ # , -- . ,-
The blower is a rot- positive type rated 60 scfm. The dehumidifier con-sists of two redundant heating elements inserted in a section of ventilation duct. The function of the dehumidifier is to sufficiently increase the te=pera-ture of the entering air to assure 70 percent relative humidity entering the filter train with 100 percent saturated air entering the dehumidifier. The
;mrpose of the dehumidifier is to assure optimum charcoal f'lter ef ficiency.
lleating element control is provided by a therecswitch. Humidity indication is provided downstream of the heating ele =ents by a humidity readout gage. The filter train provides prefiltration, high efficiency particulate filtra-tion and charcoal filtration. The filter train assembly is identical in desian to the waste gas filter train assembly which is rated at 200 scfe, thus conservatively capable of performing the assigned function. Face velocity to the charcoal filter is very Icw. The charcoal filter is composed of a module consisting of two inch deep double tray carbon cells. The purge flow to the unit vent is metered using a 0-60 scfm rotometer. The purge sample flow 1; metered using a 0-12 scfm rotometer. Both of these rotometers have an accuracy of 2 two percent of full scale, and each has remote readout capability. The purge discharge rate is controlled by a blower discharge throttling valve. The purge sample activities can be collected, counted and analyzed in the radto-chemistry lahoratory. Makeup air to the Reactor Building is supplied by a compressed air system connection to one of the afore=entioned existing
..n t an! pressurization connections.
That portion of the penetration room ventilation system piping and valves which ii u s e ... is a part of the purge ay: stem io permanen:1y installed and R designed for seismic loading through the existing vent and pressurization connections. The remainder of the purge system is the portable purging station which is stored in an area where an earthquake will not damage it. Followi g a LOCA, V there is adequate time before purging is required to permit checkc,ut of the portable pur,;ing station and to optimize the system operation to minimize the total dose to the public. r-e r J References ,.. , , . , i.
\ f p$i h
%g /gi rSAR Section 14A
? Jv YV
4.5 EMERGENCY CORE COOLING SYSTEM AND REACIOR SUILDING COOLING SYSTEM PERIODIC TESTING s-4.5.1 Emergency Core Cooling Systems Applicability Applies to periodic testing requirement for emergency core cooling systems. Objective To verify that the emergency core cooling syste=s are operable. Specification 4.5.J 1 System Tests 4.5.1.1.1 High Pressure Injection System (a) During each refueling period, a system test shall be conducted to demonstrate that the system is operable. A test signal will be applied to demonstrate actuation of the high pressure injection system for emergency core cooling operation. (b) The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly; all appropriate pu=p breakers shall have opened or closed and all valves shall have completed their travel. 4.5.1.1.2 Low Pressure Injection System (a) During each refueling period, a system test shall be conducted to demonstrate that the system is operable. The test shall be perfor=ed in accordance with the procedure s ummarized below: (1) A test signal will be applied to demcastrate actuation of the low pressure injection system , for emergency core cooling operation. l (2) Verification of the engineered safety features function of the low pressure service water sys tem which supplies , cooling water to the low pressure coolers shall be made j to demonstrate operability of the coolers. (b) The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly; all appropriate pump i breakers shall have opened or closed, and all valves l shall have completed their travel. s_ 4.5-1
4.5.1.1.3 Core Flooding System d (a) During each refueling period, a system test shall be con- %/ ducted to demonstrate proper operation of the system. During pressurization of the Reactor Coolant System, verification shall be made that the check and isolation valves in the core flooding tank discharge lines operate properly. (b) The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened. t 4.5.!.2 Copponen t_ __Tes t s 4.5.1.2.L Pumps e i At intervals not to exceed 3 conths, the high pressure and low i oressure injection pu=ps shall be started and operated to verify proper operation. Acceptable performance will be indicated ; if the pump starts, operates for fif teen minutes, and the , cischarge pressure and flov are within 210% of a point
- on the punp head curve.
4.5.1.2.2 Valves - Power Operated (a) At intervals not to exceed three month = each engineered safety features valve in the emergency core cooling systems and each engineered safety features valve associated with _j emergency core cooling in the low pressure service water system shall be tested to verify operability. i (b) The acceptable performance of each power operated valve will he that motion is indicated upon actuation by 7 appropriate signals. j ".TTd 1he eme rgency core cooling systems are the principle reactor safety features in the event of a loss of coolant accident. The removal of heat from the ' core provided by these systems is designed to limit core damage. - 1he high pressure injection system under normal operating conditions has one pump operating. At least once per month, operation will be rotated to another high pressure injection pump. This will help verify that the high pressure injection pumps are operable. 1ho requi rerents of the low pressure service water system for cooling water I are more severe during normal operation than under accident conditions. Ro - 4 t at ion of the pump in operation on a monthly basis will verify that two pumps a re ape rab l e . 1ho low p resnure inicction pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves
,/
4 4.5-2 i
-rm- .- ..,%
y . --- , , w . ,, ,m.,,
- ,. . - -n .-. ,
In the borated water storage tank fill line. This allows water to be pumped f rom the borated water storage tank through each of the injection lines and back to the tank. With the reactor shutdown, the valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood catka verify that the check and isolation valves have opened. REFERENCE FSAR Section 6 v l l i l V
- .3-3 !
I
a.-*. e . - + --w -mm -o> .J4.+ __wa-A -a__ --
.s _m ,
p., _ 1 a __ ,, 4.5.2 Reactor Building Coolinz Systems APP.1_1 cpbi1,13 d } Applies to testing of the reactor building cooling systems. Obj ec t ive To verify that the reactor building cooling systems are operable. _$pecification I
- 4.5.2.1 System Tests i
4.5.2.1.1 Reactor Building Spray System (a) During each refueling period a system test shall be conducted to demonstrate proper operation of the system. A test signal will be applied to demon-strate actuation of the reactor building spray system (except for reactor building inlet valves to prevent water entering nozzles). Water will be circulated from the borated water storage tank through the l reactor building spray pumps and returned through the test line to the borated water storage tank. (b) Station compressed air will be introduced into the j spray headers to verify the availability of the ,,, j headers and spray nozzles at least every five years. (c) The test will be considered satisfactory if visual observation and control board indication verifies that all components have responded to the actuation signal properly; the appropriate pump breakers shall have closed, all valves shall. have completed their travel. 4.5.2.1.2 Reactor Building Cooling System i (a) During each refueling period, a system test shall be conducted to demonstrate proper operation of the system. The test shall be performed in accordance with the procedure summarized below: ; l (1) A test signal will be applied to actuate the l reactor building cooling system for reactor l building cooling operation. i 4
%g 1
1 4.5-4 i
- . - - . -. - . - . _ . = -.- . - . . - -_ . ..-. - -- --- -
(2) Verification of the engineered safety features function of the low pressure service water system which supplies coolant to the reactor building coolers shall be made to demonstrate operability of the coolers. (b) The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly; the appropriate pump breakers shall have completed their travel. Fans are running at half speed and haat removal capacity is no lower than 75% of the initial tested value. 4.5.2.2 Component Tests 4.5.2.2.1 Pumps At intervals not to exceed 3 months the reactor building spray pumps shall be started and operated to verify proper operation. Acceptable performance will be indicated if the pump starts, operates for fifteen minutes, and the discharge pressure and flow are within
*10% of a point on the pump head curve.
4.5.2.2.2 Valves At intervals not to exceed three months each engineered safety features valve in the reactor building spray and reactor building cooling system and each engineered safety fea:ures valve associated with reactor building cooling in the low pressure service water A-- system shall be tested to verify that it is operable. MNF" The reactor building cooling systems and reactor building spray system are designed to remove the heat in the containment atmosphere to prevent the building pressure from exceeding the design pressure. The delivery capability of one reactor building spray pump at a time can be tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corres-ponding pump. Pump discharge irssure and flow indication demonstrate performance. With the pumps shut down and the borated water storage tank outlet closed, the reactor building spray injection valves can each be opened and closed.by operator action. With the reactor building spray inlet valves closed, low pressure air or fog can be blown through the test connections of the reactor building spray nozzles to demonstrate that the flow paths are open. The equipment, piping, valves, and instrumentation of the reactor building cooling system are arranged so that they can be visually inspected. The cooling unit s and associated piping are located outside the secondary concrete shield. Personnel can enter the reactor building during power operations to inspect and maintain this equipment. The service water piping and valves outside the reactor building are inspectable at all times. Operational tests and s ,, inspections will be performed prior to initial startup. 4.5-5
One low pressure service water pump will normally be operating. At least once per month, the operation will be rotated to another low pressure service water , piimp . Testing will be, therefore, unnecessary. The reactor building fans are normally operated periodically, constituting the test that these fans are operable. REFERENCE FSAR, Section 6 J J 4.3-6
4.5.3 Penetration Room Ventilation System s_, Aoplicability Applies to testing of the reactor building penetration roon ventilation system. OlQcctive To verify that the penetration room ventilation syste= is operable. Specification 4.5.3.1 bystem Tests 4.5.3.1.1 At intervals not to exceed 3 months, a system test shall be conducted to demonstrate proper operation of the system. This test shall consist of visual inspection, a flow measurement using the flow instrument installed at the outlet of each unit and pressure drop measurements across each filter unit. In addition, a test signal will be applied to demonstrate proper actuation of the penetration room ventilation system. Fan motors shall be operated continuously for at least one? ar, and the louvers and other mechanical system shall be proven operable and adjustable f rom their remote location.
- 4. 5. 3.1. 2 The test will be considered satisfactory if control board indicatica verifies that all components have responded properly to the actuation signal, if flow rate -through the system is design flow
or higher, aad if pressure drops across any filter bank do not exceed two times the pressure drop which existed when the filters were new.
- 4. 5.1. 2 Filter Tests No less frequently than each normal refueling period, "in-place" Icakage tests using DOP on HEPA units and Freon-112 (or equivalent) on charcoal units shall be performed at design flow on each filter train. Removal of 99.5% DOP by each entire HEPA filter unit and removal of 99.0% Freon-ll2 (or equivalent) by each entire charcoal adsorber unit shall constitute acceptable performance. These tests must also be performed after any maintenance which may affect the structural integrity of the filtration system units. .
Bases The penet ration room ventilation system is designed to collect and process pot ent ia l reactor building penetration room leakage to minimize environmental activity levels resulting from post accident reactor building leaks. The system consists of a sealed penetration room, two redundant filter trains, and two redundant fans discnarging to the unit vent. The entire system is activated by a 4 psig reactor building pressure engineered safety features signal and initially required no operator action. v 4.5-7
o , Each filter train is constructed with a prefilter, an absolute filter, and
- a charcoal filter in series. The design fir *' rate through each of these filters y 1 is 1000 sefa, which is significantly higher than the approximately 15 scfm maxtmum icskage rate frem the reactor building at a leak rate of 0.25% per day.
Except for periodic ventilation of the penetration room, the penetration room ventilation system is not normally used. Quarterly testing of this system will show that the system is available for its engineered safety features function. During this test, the system will be inspected for such things as water, oil, or other foreign material, gasket deterioration, and unusual or excessive noise or
- 18. vibration when the fan =otor is running.
l l Less frequent testing will verify the efficiency of the absolute and charcoal i filters, i l l l*
%,/ '
i i i i I i
%,/
4.5-6 Rev. 18 3/10/72
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i 1 4.5.4 Low Pressure Injection System Leakage Applicability Applies to Low Pressure Injection System leakage. Objective To maintain a preventative leakage rate for the Low Pressure Injection System which will prevent significant offsite exposures. Specification 4.5.4.1 Acceptance Limit The maximum allowable leakage from the Low Pressure Injection System components (which includes valve stems, flanges and pu=p seals) shall not exceed two gallons per hour. 4.5.4.2 Test , During each refueling period the following tests of the Low Pressure Injection System shall be conducted to determine leakage:
- a. The portion of the Low Pressure Injection System, except as
'~ specified in (b), that is outside the containment shall be tested either by use in normal operation or by hydrostatically testing at 350 psig,
- h. Piping from the containment emergency sump to the low pressure injection pump suction isolation valve shall be pressure tested at no less than 59 psis,
- c. Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collection and weighing or by another equivalent method.
pases The leakage rate limit for tSe Low Pressure Injection System is.a judgment value based on assurine .... the components can be expected to operate without mechanical failure for a period on the order of 200 days after a !.oss of Coolant Accident. The test pressure (350 psig) achieved either by normal system operation or ey hydrostatically testing, gives an adequate margin over the highest pressure within the system after-a desian basis accident. Similarly, the pressure test for the return lines from the containment to the Low Pressure Injection System (59 psig) is equivalent to the design pressure of the containment. The dose to the thyroid calculated as a result of this leakage is .76 rem for a 2 hr. exposure at the site boundary. (1) 4.5-9
. s RVVgggycgg I' SAN. Sect: R 14.2,2,4,4, V a l I 9
*.5-10
. i 4.6 EMERGENCY PC'4ER SYSTEM PERIODIC TESTING v Applicability Applies to the periodic testing and surveillance of the emergency power system. Obiective To verify that the emergency power sources and equipment will respond promptly and properly when required. Specification 4.6.1 At intervals not to exceed one month, a test of the Keowee Hydro units shall be performed to verify proper operation of these emergency power sources and associated equipment. This test shall assure that:
- a. Each hydro unit can be automatically started from the Oconee Control Room,
- b. Each hydro unit can be synchronized through the 230 kV overhead circuit to the startup transfor=ers, j l
- c. Each hydro unit can energize the 13.8 kV underground feeder. )
4.6.2 During each calendar year at a refueling outage, a test of the Keowee hydro units and emergency start circuits shall be performed to verify that each hydro unit and associated equipment is available to carry load within 25 seconds of a simulated requirement for engineered safety features. 4.6.3 At intervals not to exceed the normal refueling outage, simulated emerger.cy transfers to the 4160 volt main feeder buses shall be l g9, made to transformer CTl and to the 4160 volt standby buses to verify proper operation. 4.6.4 At intervals not to exceed the normal refueling interval, the External Grid Trouble Protection System Logic shall be tested to demonstrate its ability to provide an isolated power path between Keowee and Oconee.
- 19. l4.6.5 At intervals not to exceed the normal refueling intervals, it shall be demonstrated that a Lee Station combustion turbine can be started and connected to the 100 kV line. It shall be demonstrated that the 100 kV line can be separated from the rest of the system and supply power to the 4160 volt =ain feeder buses.
- 19. l4.6.6 Batteries in the Instrument and Control, Keowee Station, and Switching Station 125 volt DC systems shall be tested as follows:
v 4.6 "1 Rev. 19. 5/5/72
- a. The voltage and temperature of a pilot cell in each bank shall be measured and recorded daily, five days / week, s/
- b. The specified gravity and voltage of each cell shall be measured and recorced every conth,
- c. Before initial operation and at five-year intervals coincident 39' with the refueling outages, a one-hour discharge test at the required emergency load will be made.
- 19. l 4.6.7 The operability of the individual diode monitors in Instrument and Control and Keowee Station 125V DC system shall be verified on a monthly basis by imposing a simulated diode failure signal on the monitor.
- 19. 4.6.8 The peak inverse voltage capability of each auctioneering diode in the Instrument and control; Switchyard, and Keowee Station 125V DC system shall be measured and recorded every six months.
A c , ac1 43 l 19. l4.6.9 The tests specified in 4.6.6, 4.6.7, and 4.6.8 will be considered satisfactory if control room indication and/or visual examination demonstrate that all ecmponents have operated properly. 3asas The Keovee Hydro units, in addition to serving as the emergency power sources for the Oconee Nuclear Station, are power generating sources for the Duke ~/ system requirements. As power generating units, they are operated frequently, normally on a daily basis at loads equal to or greater than required by Table 8.5 of the FSAR for ESF bus loads. Normal as well as emergency startup and operation of these units will be f rom the Oconee Unit 1 Control Room. The frequent starting and loading of these units to meet Duke system l power requirements assures the continuous availability for emergency power for the Oconee auxiliaries and engineered safety features equipment. It will be verified that these units are available to carry load within 25 seconds, including instrumentation lag, af ter a simulated requirement for engineered safety features. To further assure the reliability of these units as emergency power sources, they will be, as specified, tested for auto-
, matic start on a monthly basis from the Oconee control room. These tests will include verification that each unit can be synchroni=ed to the 230 kV bus and that each unit can energize the 13.8 kV underground feeder.
The interval specified for testing of transfer to emergency power sources , is based on maintaining maximum availability of redundant power sources. Starting a Lee Station gas turbine, separation of the 100 kV line from the remainder of the system, and charging of the 4160 volt main feeder buses are specified to assure the continuity and operability of this equipment. REFERENCE FSAR, Section 8 v 4.6-2 Rev. 19. 5/5/72
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t 4.7 REACTOR CONTROL ROD SYSTEM TESTS N- 4.7.1 Control Rod Drive System Functional Tests a Af yllcability Applies to the surveillance of the control red system. objective To assure operability of the control rod system. i Specification
- 4. 7.1.1 The control rod trip insertion time shall be measured for each 1
control rod at either full flow or no flow conditions following each refueling outage prior to return to power. The maximum control rod trip insertion time for an operable control rod drive mechanism, except for the Axial Power Shaping Reds (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconds at reactor coolant full flow conditions or 1.32 seconds for no flow conditions. For i the APSRs it shall be demonstrated that loss of power will not cause rod movement. If the trip insertion time above is not j met, the rod shall be declared inoperable.
'. . / . l . 2 If a control rod is misaligned with its group average position 7 s,, by note than an indicated nine (9) inches, the rod shall be declared inoperable- However, if a safety rod absolute or re-lative position indication is inoperable and an energized out-limit light confirms the rod is fully withdrawn, the safety rod shall not be considered misaligned.
- 4. 7.1.1 If a control rod cannot be exercised, or if neither absolute or relative position indication is operable, the rod shall be de-clared to be inoperable.
niidh The control rod trip insertion time is the total elapsed time frcm power int errupt ion at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The I specified trip time is based upon the safety analysis in FSAR, Section 14. Kach control rod drive mechanism shall be exercised by a movement of approx-imately two (2) inches of travel every two (2) weeks. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions. Exercising the drive mechanisms in this =anner provides assurance j of reliability of the mechanisms. A rod is considered inoperable if it cannot be exercised, if the trip insertion time is greater than the specified allowable time, or if the rod deviates from its group average position by more than nine (9) inches. Conditions for i v 4.7 _ . _ _ . .- __ _. - - .
P oper.ition with an inoperable rod are specified in Technical Specification 3.5.2.(2) y nevE:ue:cEs (1) I'S A R , Section 14 (2) Technical Specification 3.5.2 J l 4.7-2
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4.7.2 Control Rod Proeram Verification (Group vs. Core Positions) b Appifcability
- Applies to surveillance of the control rod systems.
Obtective I To verify that the designated control rod (by core position 1 through 69)
! is operating in its programmed functional position and group. (Rod 1 through 12. Group 1-8) 1 Specificatinn ;
4./.2.1 Whenever the control rod drive patch panel is locked (after in-spection, test, reprogramming, or maintenance) each control rod i drive mechanism shall be selected from the control room and j exercised by a movement of two inches or less to verify that the proper rod has responded as shown on the unit computer printout of that rod. l 4.7.2.2 Whenever power or instrumentation cables to the control rod drive assemblics atop the reactor or at the bulkhead are disconnected or removed, an independent verification check of their reconnection shall be performed. 4.7.2.3 Any rod found to be improperly programmed shall be declared in-operable ontI1 properly programmed. nases Each cont rol rod has a relative and an absolute position indicator system. one set of outputs goes to the plant computer identified by a unique number (i ebroup.h 69) associated with only one core position. The other set of outputu goes to a programmable bank of 69 edgewise meters in the control
, room. In-the event that a patching eYror is made in the patch panel or connectors in the cables leading to the control rod drive assemblies or j
to the control room meter bank are improperly transposed upon reconnection, { these errors and transpositions will be discovered by a comparative check by (1) nelecting a specific rod from one group (e.g. Rod 1 in Regulating ] croup 6) (2) noting that the program-approved core position for this rod of the group (assume the approved core position is No. 53) (3) exercise l the selected rod and (4) note that (a) the computer prints out both i absolute and relative position response for the approved core poisition i (assumed to be position No. 53) (b) the proper =eter in the control room display bank (assumed to be Rod 1 in Group 6) in both . absolute and relative meter positions. This type of comparative check will not assure detection
; of improperly connected cables inside the reactor building. For these, (Spec. 4.7.2.2) it will be necessary for a responsible person, other than tbo one doing the work, to verify by appropriate means that each cable has i
.t.een matched to the proper control rod drive assembly. i l
i I 4.7-3 l
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4.8 MAIN STEAM STOP VALVES b Applicability Appiles to the main steam stop valves. oJQce t l_ve To verily the ability of the main steam stop valves to close upon signal and to verify the leak tightness of the main steam stop valves. Specificatiqn 4.8.1 Using Channels A and B the operation of each of the main steam stop valves shall be tested no less frequently than the normal refueling period interval to demonstrate a closure time of one second or less in Channel A and a closure time of 15 seconds or less for Channel B. 4.H.2 The icak rate through the main steam stop valves shall not exceed
, 25 cubic feet per hour at a pressure of 59 psig and shall be tested no less frequently than the normal refueling period.
n.c m s the main steam stop valves limit the reactor coolant system cooldovn rate and resultant reactivity insertion following a main steam line break accident. s, Th.Ir ability to promptly close upon redundant signals will be verified at each scheduled refueling shutdown. Channel A solenoid valves are designed to close all four turbine stop valves in 240 milliseconds. The backup Channel a solenoid valves are designed to close the turbine stop valves in approximately 12 seconds.(1) Using the maximum 15 second stop valve closing time, the fouled steam generator inventories, and the minimum tripped rod worth with the maximum stuck rod worth, an analysis similar to that presented in FSAR Section 14.1.2.9, (but considering a blowdown of both stean generators) shows that the reactor will remain sub-critical after reactor trip follcwing a double-ended steam line break. The main stop valves would become isolation valves in the unlikely eveat that there should be a rupture of a reactor coolant line concurrent with rupture of the steam generator feedwater header. The allowable leak rate of 25 cubic feet per hour is approximately 25 percent of total allowable containment leakane from all penetrations and isolation valves.(2) REFERENCES ! l (1) FSAR, Supplement 2, Page 2-7 (2) Technical Specifications 4.4.1 4.8-1
4.9 EMERCENCY FEEDWATER PUMP PERIODIC TESTING-ApplIcabi!Ity 1 Applies to the periodic testing of the turbine driven emergency feedwater , pump. Obiective To verify that the emergency feedwater pump and associated valves are operable. Specification 4.9.1 Test On a three-month basis, the turbine driven emergency feedwater pump shall be operated on recirculation to the upper surge tank for a minimum of one hour. 4.9.2 Acceptance Criteria . These tests shall be considered satisfactory if control board j indication and visual observation of the equip =ent demonstrates that all components have operated properly i Iwnen s_. The three (3) month testing frequency will be sufficient to verify that the ,
; turbine driven emergency feedwater pump is operable. Verification of correct ; operation will be made both from the control room' instrumentation and direct ; visual observation of the pump.
REFERENCE l FSAR, Section 10.2.2 FSAR. Section 14.1.2.8.3 i l l+ i w 4.9-1
4.10 REACTIVITY ANOMALIES b., Appilcability Applies to potential reactivity anomalies. Objective To requi re the evaluation of reactivity anomalies of a specified nagnitude occurring during the operation of the unit. Specification Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one percent in reactivity, an evaluation as to the cause of discrepancy shall be made and reported to the Atomic Energy Commission. Bases , To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the cont rol rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power l operation proceeds, the measured boron concentration is compared with the ! predicted concentration and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization abould be completed af ter about 10% of the total core burnup. The reaf ter , actual horon concentration can be compared with prediction, and the reactivity st atus of the core can be continuously evaluated. Any reactivity anomaly nreater than 1% would be unexpected, and its occurrence would be thoroughly invest innteil and evaluated. The value of 1% is considered a safe limit since a shutdown margin of at least II with the most reactive rod in the fully withdrawn position is always maintained. I l v i 4.10-1
4.11 ENVIRONMENTAL SURVEILLANCE b Applicability Applies to the routine testing of the station environs for radiation and radioactive materials attributable to station operation and waste releases. Objective To establish a sampling schedule for the purpose of detecting, measuring and evaluating any significant effects of station operation and waste releases on the environment. Specification 4.11.1 Environmental samples taken in accordance with Table 2-la of the FSAR shall be collected and processed in accordance with Table 4.11-1. 4.11.2 Thermoluminescent dosimeters will be installed at various locations within the exclusion area including shoreline areas, and read quarterly, nases The program will be conducted in accordance with Section 2.7 of the FSAR.
% The sensitivity of counting used in the analyses (90% confidence level when counted for 20 miautes) is based on nominal background levels of 0.5 cpm alpha and 1.0 cpm beta.
Environmental monitoring results will be correlated with information on stat ton radioactive waste releases, site meteorological data and radiological controls, and with information obtained from the installed process radiation monitoring system. The Environmental Surveillance Program will provide means of detecting signi-fleant cha.iges in levels of radioactivity. The resalts will demonstrate the effectiveness of station control over radioactive vaste disposal operations and of compliance with Federal and State regulations for disposal of these ma te ri als . The thermoluminescent dosimeter data will be used in conjunction with routine surveillance of recreational activities on the lake within the exclusion area to assure that the exposure of the public will be maintained to levels within the limits of 10CFR20 for unrestricted areas. y REFERENCES M Ad /aur p.M J FSAR, Section 2.7 jr 3/ 4.11-1 s l}Pl
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t IABLE ' 11-1 ENVIR021 ENTAL SL'RVEILLANCE PROGRMI
- L pe Sanples Schedule Analysis Water (1) (3) Monthly Gross Alpha & Beta Activity Gamma Analysis
- Rain, Settled Dust Wnthly Air Particulate Nnthly
- Vegetation, Terrest rial (2) Quarterly l !
Aquatic Organisms Quarterly Bot tom Sediment Quarterly Radiation Dose &'Duse Rate Quarterly mr & cr/hr Animals Quarterly Sr90 Csl37 Fish Quarterly Sr90 Csl37, K4 0,1131 Milk Quarterly Sr90 Cs137, g40, y131
- Dependent on gross activity ,
(1) Lakes Keowee and Ilartwell will be sampled annually for tritium analysis. (2) Commercial Crops will be substituted when available. (3) The water samples from the Keowee River continuous sac:pler will be analyzed monthly for gross Alpha and Beta and underga an isotopic analysis. l
i 4.12 CONTROL ROOM FILTERING SYSTEM
\-- Applicability Applies to control room filtering system components.
4 Objective To verify that these systems and components will be able to perform their ; design functions. Specification 4.12.1 Operating Tests System tests shall be performed at approximately quarterly intervals. These tests shall consist of visual inspection, a flow measurement using a flow instrument installed at the outlet of each unit and pressure drop measurements across each filter bank. Pressur; drop across pre-filter shall not exceed 1" H O 2and pressure drop across llEPA shall not exceed 2" H20. Fan cocora shall be operated con-tinuously for at least one hour, and all louvers and other mechanical systems shall be proven operable. 4.12.2 Filter Tests s During each refueling period, "in-place" leakage tests using DOP on l llEPA units and Freon-112 (or equivalent) on charcoal units shall be performed at design flew on each filter train. Removal of 99.5% DOP by each entire HEPA filter unit and removal of 99.0% Freon-112 l (or equivalent) by each entire charcoal adsorber unit shall constitute i acceptable performance. These tests must also be performed af ter ] any maintenance which =ay affect the structural integrity of either the filtration system units or of the housing. l Bases The purpose of the control room filtering system is to li=it the particulate l and gaseous fission products to which the centrol area would be subjected l during an accidental radioactive release in or near the Auxiliary Building. The system is designed with two 100 percent capacity filter trains each oi which consists of a prefilter, high efficiency particulate filters, charcoal l filters and a booster fan to pressurize the control room with outside air. Since these systems are not normally operated, a periodic test is required-to insure their operability when needed. Quarterly testing of this system
; will show that the system is available for its safety action. During this test the system will be inspected for such things as water, oil, or other foreign material; gasket deterioration, adhesive deterioration in the llEPA units; and unusual or excessive noise or vibration when the fan motor la running.
s_, . Annual testing will verify the ef ficiency of the charcoal and absolute filters. 4.12 l l
4.13 FUEL SURVEILLANCE 4% Applicability Applies to the fuel surveillance program for fuel rods. Obiective To specify the fuel surveillance program for fuel rods. Specification Measurements of the diameter and length will be made on selected peripheral rods of the following fuel assemblies both prior to operation and after their respective cycles:
- a. The center assembly after the first fuel cycle.
- b. Four assemblics from those removed af ter the second cycle (two containing all pressurized rods and two containing all unpressurized rods).
- c. Two assemblies containing all pressurized rods from those re=oved after the third cycle.
liases Oconce Unit I reactor contains eighty (80) fuel assemblies with initially s,, pressurized fuel rods. The reference schedule for removing these fuel assemblies is as follows: Twelve (12) of these assemblies will be removed from the core and stored in the spent fuel pool at the end of the second fuel cycle. Twenty-eight (28) assemblies will be removed at the end of the third cycle. For some of the second and third fuel cycles in the specifi-cation, the fuel may be irradiated in Oconee Unit 3. The remaining forty (40) fuel assemblies with pressurized fuel rods will be re=oved at the end of the first fuel cycle and are scheduled to be charged to Oconee Unit 3. Eight (8) pressurized rods will be located on the periphery of the center fuel assembly, two in each outer row of rods, and the assembly is to be removed after the first cycle. i This program provides substantiating information for the first core in the present generation of B&W reactors. It provides fuel rods for examination at the end of the first, second, and third cycles to determine the extent, if
; any, of dimensional changes in diameter and length.
f' i 4 4.13-1
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4 4.14 REACTOR BUILDING PURGE SYSTEM s-- Appliest ,t li ty Applies to testing Reactor Building Purge Filters. Objective To verify that the Reactor Building Purge Filters will perform their design func' ion. Specification During each refueling period, leakage tests using DOP on the HEPA filter and Freon-112 (or equivalent) on the charcoal unit shall be performed. Removal of 99.5% DOP by the HEPA filter- unit and re= oval of 99.0% Freon-112 (or equivalent) by the charcoal adsorber unit shall constitute acceptable
- performance. These tests must also be performed after any maintenance which may affect the structural integrity of the filtration units or of the housing.
Bases The reactor building purge filter is constructed with a pref t *er, an absolute filter and a charcoal filtar in series. This test will varify the officiency I of the absolute and charcoal filters. v i 0 i sty -
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5 CESIGN FEATURES, s 5.1 SITE 5.1.1 The Oconee Nuclear Station is approximately eight miles northeast of Seneca, South Carolina. Figure 2-3 of the Oconee FSAR shows the plan of the site. The minimum distance from the reactor center line to the Loundary of the exclusion area and to the outer boundary of the low population zone as defined in 10 CFR 100.3, shall be one mile and six miles respectively. 5.1.2 For the purpose of satisfying 10 CFR Part 20, the " Restricted Area," for gaseous releas purposes only, is the same as the exclusion area as defined above except that the temporary construction quarters (2) in the east southeast section of the exclusion area shall not, when occupied, be deemed to be within the restricted area. REFERENCE (1) FSAR, Section 2.2 (2) Technical Specification 3.9 5.1-1 l
5.2 CONTAINMENT s_ Specification The containment for this unit consists of three systems which are the reactor building, reactor building isolation system, and penetration room ventilation system.
.1 Reactor Building The reactor building completely encloses the reactor and its -
associated reactor coolant system. It is a fully continuous re-enforced concrete structure in the shape of a cylinder with a shallow domed roof and flat foundation slab. The cylindrical portion is prestressed by a post tensioning system consisting of horizontal and vertical tendons. The dome has a three-way post tensioning system. The structure. can withstand. the loaa nf 3 horizontal and 3 vertical tendons in the cylinder wall or adjacent tendons in the dome without loss of function. The foundation slab is conventionally re-enforced with high strength re-enforcing steel. The entire structure is lined with 1/4" welded steel plate to provide vapor tightness. The internal volume of the reactor building is approximately 1.91 x 106 cu. ft. The approximate inside dimensions are: diameter-- 116'; height--208 1/2'. The approximate thickness of the concrete
'- forming the buildings are: cylindrical wall--3 3/4'; dome--3 1/4';
and the foundation slab--8 1/2' . The concrete containment structure provides adequate biological shielding for both normal operation and accident situations. Design pressure and temperature are 59 psig and 286*F, respectively. The reactor building is designed for an external atmospheric This is pressure of 3.0 psi greater than the internal pressure. could be i greater than the dif ferential pressure of 2.5 psig that developed if the building is sealed with an internal temperature of 120 F with a barometric pressure of 29.0 inches of Hg and the building is subsequently cooled to an internal temperature of 80 F with a concurrent rise in barometric pressure to 31.0 inches of Hg. Since - the building is designed for this pressure dif ferential, vacuum breakers are not required. s. 5.2-1
t 8 Penetration assemblies are seal welded to the reactor building liner. Access openings , . electrical penetrations , and fuel transfer tube covers are equipped with double seals. Reactor '" building purge penetrations and reactor building atmosphere sampling penetrations are equipped with double valves having~ resilient seating surfaces. (1) The principal design basis for the structure is that it be capable of withstanding the internal pressure resulting from a loss of coolant accident, as defined in FSAR Section 14 with no loss of integrity. In this event the total energy contained in the water of the reactor coolant system is assumed to be released 1nto the reactor building tbrough a break in the reactor coolant piping. Subsequent pressure behavior is determined by the building volume, engineered safety features, and the combined influence of energy sources and heat sinks. 5.2.2 Reactor Building Isolation System Leakage through all fluid penetrations not serving accident-consequence-limiting systems is to be minimized by a double barrier , so that no single, credible failure or malfunction of an active l component can result in loss-of-isolation or intolerable leakage. ' The installed double barriers take the form of closed piping systems, both inside and outsida the reactor building and various types of isolation valves. (2) 5.2.3 Penetration Room Ventilation System '"#
)
This system is designed to collect, control, and minimize the 1 release of radioactive materials from the reactor building to the ; environment in post-accident conditions. It may also operate l Intermittently during normal conditions as required to maintain ' satisfactory temperature in the penetrations rocms. When the system is in operation, a slight negative pressure will be main-tained in the penetration room to assure inleakage. (3) REFEREN9ES (1) FSAR Section 5.1 (2) FSAR Section 5.2 (1) FSAR Section 5.3
~i l 5.2-2
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5.3 REACTOR b Spec i fic a t ion 5.1.1 Reactor Core t i 5.1.1.1 The reactor core contains approximately 94.1 metric tons of i slightly enriched uranium dioxide pellets. The pellets are i encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 177 fuel assemblies, of which 80 are prepressurized with helium. Each fuel assembly contains i' 208 fuel rods. (1) 5.3.1.2 The fuel assemblies shall form an essentially cylindrical i lattice with an active height of 144 in, and an equivalent l f diameter of 128.9 in. (2) 5.3.1.3 The average enrichment of the initial core for Unit 1 is a nominal 2.1 weight percent of U235 Three fuel enrichments are used in the initial core. The highest enrichment is j less than 2.2 weight percent of U235, 5.3.1.4 Therc are 61 full-length control rod assemblies (CRA) and 3 i axial power shaping rod assemblies (APSR) ' distributed in the reactor core as shown in FSAR Figure 3-46. The full-length j CRA contain a 134 inch length of silver-indiu=-cadmium alloy clad with stainless steel The APSR contain a 36 inch length s,, o f silver-indium-cadmium alloy. (3) 5.1.1.5 Reload fuel assemblies and rods shall conform to design and evaluation described in FSAR and shall not exceed an anrichment j of 3.2 percent of U235, j 5.3.2 Reactor Coolant System 5.3.2.1 Tr.e design of the pressure components in the reactor coolant system shall be in accordance with the code requirements. (4) 5.1.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, shall be designed for a pressure of 2,500 psig and a temperature of 650 0F. The pressurizer and pressurizer surge line shall be designed for a temperature of 6700F. (5) 5,
%m k
.5.3-1
e O 5.3.2.3 The maximum reactor coolant system volume shall be 11,330 f 3 REFERENCES d (1) FSAR Section 3.2.1 (2) FSAR Section 3.2.2 (3) FSAR Section 3.2.4 (4) FSAR Section 4.1.3 (5) FSAR Section 4.1.2 J l l i l 1 d ) l 5.3 1 l 1 i
. t i
5.4 NEW AND SPENT FUEL STORAGE FACILITIES
'- Specification 5.4.1 New Fuel Storage 5.4.1.1 New fuel will normally be stored in the spent fuel pool j
serving the respective unit. The fuel asse=blies are stored in racks in parallel rows, having a nominal center to center 4 distance of 21 inches in both directions. This spacing is sufficient to maintain a K effective of less than .9 when flooded with unborated water, based on fuel with an enrichment
! of 3.5 weight percent U235
$ 5.4.1.2 New fuel may also be stored in the fuel transfer canal. The i fuel assemblies are stored in five racks in a row having a nominal center to center distance of 2' 13/4". One rack I is oversized to receive a failed fuel assembly centainer, The other four racks are normal size and are capable of q receiving new fuel assemblies. i 1 5.4.1.3 New fuel may also be stored in shipping containers. 5.4.2 Spent Fuel Storage 5.4.2.1 Irradiated fuel assemblies will be stored, prior to offsite shipment, in the stainless steel lined spent fuel pool, which s, is located in its respective auxiliary building. Each pool is sized to accommodate a full core of irradiated fuel assemblies in addition to the concurrent storage of the largest quantity of new and spent fuel assemblies predicted by the fuel management program. i 5.4.2.2 Whenever there is fuel in the pool (except the initial core
, loading), the spent fuel pool is filled with water borated to the concentration that is used in the reactor cavity and fuel ! transfer canal during refueling operations.
1 3.4.2.3 Spent fuel may also be stored in storage racks in the fuel transfer canal when the canal is at refueling level. 5.l.2.4 The spent fuel pool and fuel transfer canal racks are designed i for an earthquake force of 0.lg ground motion. I !~ RE FERENCES 1 FSAR, Section 9. 7 {
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- 6. ADMINISTRATIVE CONTROLS L 6.1 ORGANIZATION, REVIEW AND AUDIT Introduction Administrative controls relate to the organization and management proce-dures, record keeping, review and audit systems, and reporting that are considered necessary to provide the assurance and evidence that the station will be managed in a dependable manner.
The administrative controls specify the administrative tools and functions necessary for safe operation. They also define the administrative action to be taken in the event operating it:its or safety limits are exceeded. Specification 4 These administrative controls in regard to operations, review and audit beccme effective at the time of issuance of the facility license by the AEC, and af ter nor=al design and construction activities have been essentially com-pleted. 6.1.1 Organization 6.1.1.1 The Superintendent is directly responsible for the safe operation of the facility. s,, 6.1.1.2 In all matters pertaining to actual operation and maintenance and to these Technical Specifications, the Superintendent shall report to and be directly responsible to the Assistant Vice President, Steam Production. The organization is shown in Figure 6.1-2. 6.1.1.3 The station organization for operation, Technical Support, and Maintenance shall be functionally as shown in Figure 6.1-1. 6.1.1.4 Unit 1 minimum operating shift staffing will be according to j Table 6.1-1. This minimum operating staff shall be permitted to I assist in the pre-licensing activity of Units 2 & 3 only to the extent that it does not affect their full availability for Unit 1 operation. l 6.1.1.5 Incorporated in the staff of the station shall be supervisory and ! professional personnel meeting the minimum requirements specified . in FSAR Section 12.2 encompassing the training and experience des- ! cribed in Section 4 of the " Proposed Standard for Selection and l Training of Personnel for Nuclear Power Plants" prepared by the l ANS-3 committee, Draft 9 dated 7/3/69. ) 6.1.1.6 Retraining and replacement of station personnel shall be in accordance l with Section 5.5 of the Proposed Standard for Selection and Training of Personnel for Nuclear Power Plants, ANS-3, Draft No. 9, dated 7/3/69 s- , 6.1-1
V 6.1.1.7 AEC/DRL licensed operators shall be provided as follows:
- a. At least two licensed reactor operators shall be at the f
station, one of whom shall be in the contro1 room, at all 0 times when there is fuel in the reactor vessel. One of these operators shall hold a Senior Reactor Operator license; f further, y
- b. Two licensed reactor operators shall be in the control t g room during startup and scheduled shutdown of a reactor, and during recovery from reactor trips caused by transients or emergencies.
- c. At least one licensed reactor operator shall be in the reactor building when fuel handling operations are in
- 19. progress in the reactor building. An operator holding a Senior Reactor Operators license shall be in charge i of all fuel handling activities.
6.1.2 Review and Audit In matters of nuclear safety and radiation exposure, review and audit
- 18. of station operation, =aintenance and technical =attera shall be pro-vided by two co=mittees as follows (Reference Figure 6.1.2):
J 6.1.2.1. Station Review Co=mittee
- a. Membership Assistant Superintendent--Chairnan Operating Engineer Technical Support Engineer At least two other members of the station supervisory staff appointed by the Superintendent.
The Superintendent shall appoint an acting chairman in the absence of the Assistant Superintendent.
- b. Meeting Frequency This committee shall meet at least once cach month and as required on call by the chairman.
- e. Quorum The chairman plus two me=bers shall constitute a quorum.
- d. Responsibilities J
6.1-2 Rev. 18. 3/10/72 Rev. 19. 5/5/72
. o The committee shall have the following responsibilities: L (1) Review all new procedures or proposed changes to existinF
- 18. procedures as determined by the Station Superintendent to affect operational safety.
(2) Review station operation and safety considerations. (3) Review abnormal occurrences and violations of Technical Specifications and =ake recommendations to prevent recurrence.
- 18. (4) Review all proposed tests that affect nuclear safety or radiation safety.
(5) Review proposed changes to Technical Specifications and changes or modifications to the Station design.
- e. Authority The Station Review Co=mittee shall make reco==endations to the Superintendent regarding Specification 6.1.2-1-d.
- f. Records Minutes shall be kept at the Station of all meetings of
" tae Committee and copies sent to the Superintendent, Assistant Vice President Steam Production, and the chair =an of the General Office Review Committee.
l 6.1.2.2. General Office Review Co=mittee
- a. The Executive Vice President,and General Manager, shall appoint a General Office Review Committee having the following me=bership:
Chairr c.a Steam Production Department---3 members (including the Superin-tendent or Assistant Superintendent but not other Oconee Nuclear Station personnel) Engineering Department---2 = embers Others deemed advisable (may include consultant from outsida the Company) The Committee shall elect a vice chair =an,
- b. At least one-half of the members of the co=mittee shall have extensive nuclear experience and all members shall be enai-neering or science graduates. Members, except Superin-tendent shall not have direct line responsibility for production and shalt meet the minimum qualifications of Section 4.6 of N-- the proposed standard of Specification 6.1.1.5.
6.1-3 Rev. 18. 3/10/72 Rev. 19 May 5, 1972 (Carry over)
- c. Quorum J
The Chairman or Vice Chairman plus three me=bers shall consti- , tute a quorum. i
- d. As a safety review and audit backup to the normal operating organization, the GORC shall:
(1) Audit procedures or proposed changes to procedures which may affect nuclear safety or radiaticn exposure. (2) Review proposed station design changes er =odifications which affect nuclear safety. (3) Review all requests to the AEC/DRL for changes in Technical Specifications that involve unreviewed safety questions , (10CFR50.59) (4) Review abnormal occurrences and v;olations of Technical Specifications and make reco==endations to prevent recurrence. (5) Periodically, audit station operating records, logs, reports, and tests. (6) Make reco==endations concerning the above items to the appropriate individuals. A copy of these reco==endations s_s/ shall be forwarded to the Executive Vice President and General Manager; Senior Vice President, Engineering and Construction; Senior Vice President Production and Transmission;
- 19. Vice President, Design Enrineering and Asst. Vice President, Steam Production, j (7) Audit Station Re"tew Cc=mittee minutes.
(8) Conduct special reviews or investigations as requested by the Assistant Vice President, Steam Production- ! I
- e. Meeting Frequency )
This committee shall meet at least three times per year at intervals not to exceed five months and as required on call by the chairman.
- f. Records Minutes shall be kept of all =eetings of the cocmittee. Copies of the minutes shall be sent to the Executive Vice President and General Manager; Senior Vice President Engineering and Construction; Senior Vice President, Production and Transmission; l
- 18. Vice President, Design Engineering; Assistant Vice President, Steam Production; and Superintendent. A copy of these minutes shall be kept on file at the station. ,,f 6.1-4 Rev. 18. 3/10/72 Rev. 19 May 5, 1972 (Carry ever)
L
( ( ( . Superintendent Visitors Center Clerical Staff Asst. Supt. . Staff Operating SRO Technical Support Maintenance Engineer Engineer Supervisor { Asst. Operating Engineer SRO* Asst. Maintenanc( Supervisor 4 Perfonnance Instrument Chemistry-Health Engineer Supervisor Physics Supv. Shift SRO Maintenance Supervisor Personnel Instrument Chemis try-IIcalth Personnel Physics Personnel Control RO Operator SRO - Senior Reactor Operator License RO - Reactor Operator License Asst. Control
- License not required for initial operation of Unit 1 Operator RO*
STATION ORGANIZATION CHART Utility sqp OCONEE NUCLEAR STATION Operator Figura 6.1-1
( ( ( Superintendent Visitors enter Clerical Staff Asst. Supt. Operating SRO Technical Support Maintenance Engineer Engineer Supervisor { Asst. Operating Engineer SRO* Asst. Maintenance Supervisor 4 ' Performance Instrument Chemistry-liealth Engineer Supervisor Physics Supv. Shift SRO Haintenance Supervisor Personnel Instrument Chemistry-liealth Personnel Physics Personnel Control RO Operator SRO - Senior Reactor Operator License RO - Reactor Operator License Asst. Control
- License not required for initial operation of Unit 1 Operator RO*
STATION ORGANIZATION CHART ng . Utility OCONEE NUCLEAR STATION 8tyf{'" Operator Figure 6.1-1
.. - _= . . - . - . ._.- - ~ . . -
- u o
TABLE 6.1-1 OCONEE NUCLEAR STATION MINIMUM OPERATING SHIFT REQUIREMENTS UNIT 1 l ! Shift Supervisor 1 SRO l Control Operator 1 RO i Asst. Control Operator
- 1 RO
- l 1
l Utility Operator 2 l Men /Shife 5 .i SRO/ Shift 1 l RO/ Shift 1 l SRO--AEC/DRL Senior U.eactor Operator License l RO--AEC/DRL Reactor Operator License ' l l *Not AEC/DRL licensed for initial operation 1 ( : ' l 4 1 r i l ) 6.1-7 +
6.2 ACTION TO BE TAKEN IN THE EVENT OF AN A3 NORMAL OCCLTLENCd v Specification 6.2.1 Any abnor=al occurrence shall be investigated promptly by the Superintendent. 6.2.2 The Superintendent shall promptly notify the Assistant Vice President, Steam Production, of any abnormal occurrence and shall cause the Station Review Committee to perform a review and prepare a written report which shall describe the circum-stances leading up to and resulting from the occurrence and shall recommend appropriate action to prevent or minimize the probability of a recurrence. 6.2.3 The Station Review Coe=ittee report shall be submitted to the , General Office Review Committee for review and approval of any recommendations. Copies shall also be sent to the Superintendent and the Assistant Vice President, Steam Producticn. 6.2.4 The Senior Vice President, Production-Transmission, shall report
- 19. the circumstances of any abnormal occurrence to the AEC as specified in Secticn 6.6, Station Reporting Requirements.
s
- s. ;
6.2-1 Rev. 19. 5/5/72 ; i l l
6 6.3 ACTION TO 3E TAKEN IN THE EVIST A SAFETY L211T IS EXCEDED v Specification If a safety limit is exceeded: 6.3.1 The reactor shall be shut down i= mediately and maintained in a safe shutdown condition until otherwise authorized by the AEC. 6.3.2 The Superintendent shall cake an i=medicte report to the Assistant Vice President, Steam Production; the Senior Vice
- 19. President, Production and Transmission; and the Chairman of the General Office Review Committee.
e 6.3.3 The circumstances shall be promptly reported to the AEC by the
- 19. j Senior Vice President, Production-Transmission as indicated in l Section 6.6, Station Reporting Requirements.
6.3.4 The Superintendent shall direct the Station Review Cc=mittee to perform an analysis of the circu= stances leading up to and resulting from the situation together with reco==endations to prevent a recurrence. The report covering this analysis shall be sent to the General Office Review Co==ittee for review and approval. Copies of this report shall also be submitted to the Superintendent; Assistant Vice Presidant, Staam Production; the Senior Vice President, Production and Transmission; the l
- 19. l Chairman of the General Office Review Committee; the Senior Vice
- President, Engineering and Construction; Vice President, Design Engineering, and the Executive Vice President and General Manager. Appropriate analyses or reports shall be submitted to the 39, l AEC by the Senior Vice President, Production-Transmission as l indicated in Section 6.6, Station Reporting Requirements.
6.3-1 Rev. 19. 5/5/72
6.4 STATION CPERATING PROCEDURES Specification 6.4.1 The station shall be operated and maintained in accordance with approved procedures. Detailed written procedures with appropriate check-of f lists and instructions shall be provided for the following conditions:
- a. Normal startup, operation and shutdown of the complete facility and of all systems and components involving nuclear safety of the facility.
- b. Refueling operations.
- c. Actions taken to correct specific and foreseen potential malfunctions of systems or components involving nuclear safety
- 18. and radiation levels, including responses to alarms, suspected primary system leaks and abnor=al reactivity changes.
- d. Emergency procedures involving potential or actual release of radioactivity,
- c. Preventive or corrective maintenance which could affect nuclear lo. safety of the reactor or radiation exposure to personnel,
- f. Station survey following an earthquake.
- g. Radiation control procedures.
- h. Operation of radioactive waste canagement systems.
- 1. Control of pH in recirculated coolant after loss of coolant accident. Procedure shall state that pH will be sampled and appropriate caustic added to coolant within 30 minutes after switchover to recirculation mode of core cooling.
6.4.2 All procedures listed in Specification 6.4.1, and changes thereto, shall be reviewed by the Station Review Committee prior to approval l hy the Superintendent for use except as provided in Specification ' 6.4.3 below. l 6.4.3 Written procedures shall be strictly adhered to in all matters relating to nuclear safety. Temporary minor changes which do not change the intent of the original procedure and which are ! not safety related are permitted only on documented approval of the appropriate supervisors. 6.4.4 Quarterly selected drills shall be conducted on site emergency procedures including assembly preparatory to evacuation off site and a check of the adequacy of communications with off-site support groups. 6.4.5 Respiratory protective program approved by AEC shall be in force. 6.4-1 Rev. 18 3/10/72
I 6.4.6 If the Superintendent concludes that a proposed =ajor change in the f acility or operating procedures, tests, or experiments in- ,_j/ volving systems that are nuclear safety or radiation exposure related does not involve a change in the Technical Specifications
- 18. or is an unreviewed safety question, he may order the change, test, or experiment to be =ade, but shall enter a description thereof in the operating records of the facility and shall aend a copy of the pertinent instructions to the Chairman of the General Office Review Committee. If the Chairman of that Committee, upon reviewing such instructions, is of the opinion that the change, test or experiment is of such a nature as to warrant consideration by the Ccmmittee, he shall order such consideration.
6.4.7 If the Superintendent desires to make a change in the facility or operating procedures or to conduct a test or experi=ent which in his opinion might involve a change in the Technical Specifications, might involve an unreviewed safety question or might otherwise not be in accordance with said License, he shall not order such change, test, or experiment until he has referred the matter to the General Office Review Committee for review and report. If that Co==ittee is of the opinion that the proposed change, test, or experiment does not require approval by the Atomic Energy Co= mission under the terms of said License, it shall so report in writing to the Superintendent, together with a statement of the reasons for the Committee decision. The Superintendent may then proceed with the change, test or experi-ment. If, on the other hand, the Committee is of the opinion that approval of the Atomic Energy Commission is required, the Committee _f shall review the request for such approval, including an appropriate safety analysis in support of the request, and forward the report and request to the Senior Vice President, Engineering and Construction,
- 18. and Senior Vice President, Production and Transmission, shall there-up'on forward the report and request to the Atomic Energy Co=nission for approval unless, af ter review, the Vice Presidents either (a) disagree with the opinion of the Com=ittee that approval of the ,
Atomic Energy Com=ission is required, or (b) decide that the j proposed change, test or experiment is not necessary. l J t 1
~/ l 6.4-2 Re- 18. 3/10/72
- 6.5 STATION OPERATING RECORDS N-- Soecification The following records shall be prepared and retained for five (5) years unless a longer period is required by applicable regulations. All records shall be retained in a manner convenient for review.
- a. Records of normal station operation including power levels and periods of operation at each power level,
- b. Records of principal maintenance activities, including inspection, repairs, substitution or replacement of principal items of equipment pertaining to nuclear safety.
- c. Reports of abnormal occurrences and safety limits exceeded.
- d. Records of periodic checks, tests, and calibrations performed to verify that requirements specified under surveillance standards are being met.
All equipment failing to meet surveillance requirements and the corrective action taken shall be recorded.
*e. Records of changes made in the station as described in the ?SAR.
*f. Special nuclear material inventory records.
*g. Routine station radiation surveys and monitoring records.
v Records of environmental monitoring surveys.
*h.
*i. Records of radiation exposure for all station personnel, contractors and visitors to the station as required *oy 10 CFR 20. _j
*j . Records of radioactive waste disposal.
- k. Records of reactor physics tests and other special tests pertaining to nuclear safety,
- 1. Channes to Operating Procedures.
- m. Shift Supervisors log.
- n. By-product material inventory records and source leak test results.
- o. Station Review Committee and General Office Review Conmittee minutes.
*These items will be permanently retained.
v 6.5-1 . s l 3
6.6 STATION REPORTING REQUIREMETIS Specification N-- In addition to reports required by applicable regulations, the following in-formation shall be provided to the Atomic Energy Commission: 6.6.1 Events requiring notification within 24 hours (by telephone or telegraph to the Director of Region II Compliance Office followed by a written report within 10 days to the Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545; with a copy to the Director of Region II Compliance Office:
"~
19.
- a. Abnormal occurrences specified in the Definitions Section of the Technical Specifications.
- b. Any significant variation of measured values in a non-conservative direction fro = corresponding predicted values of safety connected parameters during initial criticality.
The written report, and to the extent possible the preliminary telephone or telegraph report, shall describe, analyze, and evaluate safety implications, and outline the corrective actions and measures taken or planned to prevent recurrence of a. and
- b. above.
6.6.2 Events requiring raports within 30 days (in writing) to the Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545; with a copy to the Director of Region II Compliance s,, Office:
- a. Any change in transient or accident analyses, as described in the Safety Analysis Report, which involves an unreviewed safety question as defined in Paragraph 50.59 (c) of 10 CFR 50.
- b. Any changes in station operating organization which involve positions for which minimum qualifications are specified in the Technical Specifications, or in personnel assigned to these positions.
6.6.3 Events requiring reports within 60 days (in writing) to the Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545; with a copy to the Director of Region II Compliance Office:
- a. Upon receipt of a new operating license or amendment to a facility license involving the planned increase in reactor power level or the installation of a new core, a su= mary report of unit startup and power escalation test programs and evaluations of results thereof shall be submitted within 60 days following completion of testing or commencement of .
commercial power, whichever comes first, i s_, 6.6-1 Rev. 19. 5/5/72
- b. Any changes ~in the station or corporate organizations as shown on Figures 6.1-1 and 6.1-2.
6.6.4 A Semi-Annual Station Operations Report shall be prepared and submitted to the Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545, within 60 days after the end of each reporting period. The report shall provide the following information (summarized on a monthly basis) and shall cover the six month period or fraction thereof, ending June 30 and December 31. The due date for the first report shall be calculated from the date of initial criticality. 6.6.4.1 Operations Summarv
- a. A narrative su= mary of operating experience and of changes in facility design.that relate to' safe operation, perfor=ance characteristics (including fuel performance) and operating pro-
- 19. cedures related to safety occurring during the reporting period. ,
- b. A summary of results of surveillance tests and inspections.
l 6.6.4.2 Power Generation f A summary of the nuclear and electrical output of the unit during the raporting pericd, and the cu=ulative total outputs since initial criticality, including: i
- a. Gross ther=al power generated (in HRH). wM
- b. Cross electrical power generated (in MWH).
i
- c. Net electrical power generated (in MWH).
2 d. Number of hours the reactor was critical.
- e. Number of hours the generator was on line.
I
- f. Histogram of thermal power versus time.
6.6.4.3 Shutdowns Descriptive material covering all outages occurring during the reporting period. The following information shall be provided for each outage: 1 a. The cause of the outage.
- b. The method of shutting down the reactor; e.g., trip, automatic rundown, or manually controlled deliberate-shutdown. .
- c. Duration of the outage in hours.
J 6.6-2 Rev. 19. 5/5/72-
-r-, - -
- , w w w
- d. Unit status during the outage; e.g., cold shutdown, hot' shutdown, or hot standby.
19, e. Corrective and preventive action tsken to preclude re-l currence of each unplanned outage. 6.6.4.4 Maintenance A discussion of electrical, mechanical and general maintenance performed during the report period having potential effects on the safety of the facility. Specific systems involved shall be identified and information shall be provided on:
- a. The nature of the maintenance; e.g., routine or emergency.
)
- b. The effect, if any, on the safe operation of the reactor.
- c. The cause of any malfunction for which corrective maintenance was required.
- d. The results of any such malfunction.
- e. Corrective and preventive action taken to preclude recurrence.
6.6.4.5 Radioactive Effluent Releases s., Information relative to the quantities of liquid, gasecus and solid radioactive effluents released from the facility, and the effluent volumes used in maintaining the releases within the limits of 10 CFR 20 shall be provided as follows:
- a. Liquid Releases (1) Total radioactivity (in curies) released, other than tritium, and concentration at the Keowee Hydro Unit tailrace averaged over a year.
(2) Total tritium oxide (in curies) discharged, and concentration at the Keowee Hydro Unit tailrace averaged over a year. (3) Total volume (in gallons) of liquid waste released into the Keowee Hydro Unit tailrace. (4) Total volume (in gallons) of Keowee Hydro dilution water used. (5) The maxi =um concentration released (averaged over the period of a single release). (6) Estimated total radioactivity (in curies) released, by nuclide (other than tritium), based on representative isotopic analyses performed. 6.6-3 Rev. 19. 5/5/72 l l 4
(7) Percent of applicable limits released, based on nuclide identification performed, s,,/
- b. Gaseous Releases (1) Total radioactivity (in curies) released.
(2) Estimated total radioactivity (in curies) released, by nuclide (other than tritium), based on representative isotopic analyses performed of: (a) noble gassa (b) halogens (c) particulates with half-lives greater than eight days (d) tritium oxide
- c. Solid Waste ,
(1) The total amount of solid waste packaged (in cubic feet). (2) The dates of shipment and disposition (if shipped off-site). J (3) Estimated total radioactivity (in curies):
- a. Packaged
- b. Shipped 6.6.4.6 Environmental Monitoring (1) For each =edium sampled during the si:c-month period, the following information shall be provided.
- a. Number of sampling locations
- b. Total number of samples
- c. Number of locations at which levels are found to be significantly greater than local backgrounds.
V 6.6-4
- d. Highest, lowest, and the average concentrations or levels of radiation for tha sa:pling point with the N-, highest average and descriptier. of the location of that point with respect to the site.
(2) If levels of station contributed radioactive =aterials in environmental media indicate the likelihood of public intakes in excess of 3 percent of those that could result from continuous exposure to the concentration values listed in Appendix B, Table II, Part 20, esti=ates of the likely resultant exposure to individuals and to population groups, and assu=ptions 2pon which estimates are based shall be provided. (These values are comparable to the top of Range I, as defined in FRC Report No. 2.) (3) If statistically significant variations in off-site environmental concentrations with time are observed, and are attributed to stat 10n releases correlation of these results with effluent releases shall be provided. 19, 6.6.5 Special Reports Special test reports shall be prepared and submitted to the Director, Division of Reactor Licensing, USAIC, Washington, D. C. 20545, as follows:
- a. A report
- within 60 days of completion of one year of comnercial operation covering:
w (1) An evaluation of unit performance to date in comparison with design specifications. (2) A reassessment of the validity of prior accident analyses in light of =easured operating characteristics, which may affect consequerces; and system, cocpenent, i and personnel performance which may affect accident probabilities. l l (3) A progress and status report on all items identified i in the operating license review as requiring further effort.
*This report may be incorporated in the Semi-Annual Operations Report due at that time.
- b. The initial reactor building integrated leak rate test shall be the subject of a su= mary technical report and shall be submitted within 90 days to the Director, Division of Reactor Licensing, USAEC, Washington, D. C. 20545, and
- 19. shall include analyses and interpretations of the results which demonstrate cenpliance in meeting the leak rate limits ,
specified in the Technical Specificaricns. Other containment l leak rate tests that fail to meet the acceptance criteria s_, shall be the subject of a special sunmary report. 6.6-5 Rev. 19. 5/5/72
i l i
- c. A report covering single loop operation, permitted by Specification 3.1.8, within 90 days after completion of s/
testing. This report shall include the data obtained as noted in 3.1.8. together with analyses and interpretations of these data which demonstrate: (1) Coolant flows in the idle loop and operating loop are 4 as predicted in FSAR Supplement 7, Tables 1-1 and 1-2. (2) Relative incore flux and temperature profiles remain essentially the same as for four pump operation at each power level taking into account the reduced ficw in single loop operation. (3) Operating loop temperatures and flows are obtained which justify the revised safety system setting prescribed for the temperature and flow instruments located in the operating loop (which must sense the combined core flow plus the cooler bypass flow of the idle loop).
- d. The results of the initial reactor building structural tests (as specified in FSAR Section 5.6.1.2 including the Structural i Instrumentation Report contained in Amendment No. 25, dated December 30, 1970) shall be reported within 90 days following completion of the test.
- e. A reactor building structural integrity report shall be '"#
submitted within 90 days of completion of each of the following tests covered by Technical Specification 4.4* (the integrated leak rate test is covered in 6.6.4.7b above): *May be included in Semi-Annual Operations Report. (1) Annual Inspection (2) Tendon Stress Surveillance (3) End Anchorage Concrete Surveillance (4) Liner Plate Surveillance
- f. Inservice Inspection Program
- g. Report to be submitted upon completion of program on fuel surveillance per Specification 4.14.
s.d - 6.6-6
r 6.7 RADIOLOGICAL CONTROLS 1'
\*- Specification 6.7.1 The radiation protection program shall be organi*ed,with the following exceptions, to meet the requirements of 10 CFR 2v.
- a. Pursuant to 10 C7R 20.103(c)(1) and (3), allowance may be =ade for the use of respiratory protective equipment in conjunction with activities authorized by the operating license for this station in determining whether individuals in the Restricted Area are exposed to concentrations in excess of the limits specified :ba Appendix B, Table I, Column 1, of 10 CFR 20, subject to the
-i following conditions and limitations: a
- 1. Notwithstanding any exposure limit provided herein, the licensee shall, as a precautionary procedure, use process or other engineering controls, to the extent practicable, to limit con-centrations of radioactive materials in air to levels below those which delimit an airborne radioactivity area as defined in 20. 203 (d) (1) .
- 2. When it is impracticable to apply process or other engineering controls to limit concentrations of radioactive =aterials to levels below those which delimit an airborne radioactivity area as defined in 20.203(d)(1), and respiratory protective equipment is used to limit the inhalation of airborne radio-s,, active material, the licensee may make allowance for such use
' in estimating exposures of individuals to such materials provided: , (a) Intake of radioactive material by any individual within any period of seven consecutive days will not exceed that which would result from inhalation 1/2/3/ of such material 40 hours per week, at uniform concentrations specified in Appendix B, Table I, Column 1 of 10 CFR Part 20.
1/ Since the concentration specified for tritium oxide vapor assumes equal intakes by skin absorption and inhalation, the total intake permitted is twice that which would result from inhalation alone at the concentration specified for H 3, S in Appendix B, Table I, Column 1 for 40 hours 2/ For radioactive materials designated "Sub" in the " Isotope" column of the table, the concentration value specified is based upon exposure to the material as an external radiation source. Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in 20.101. These materials shall be subject to applicable precautionary proceduret of Paragraph
; 6.7.1.a.1 above.
3/ For modes of intake other than inhalation, such intakes must be controlled, evaluated, and accounted for by techniques and procedures as =ay be appropriate to the circumstances of the occurrence with proper consideration of critical organs and limiting doses. 6.7-1
(b) Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive '" material inhaled by an individual wearing the equipment
, do not exceed the pertinent values specified in Appendix B, Table I, Column 1 of 10 CFR Part 20. For the purposes of this subparagraph, the concentration of radioactive material that is inhaled when respirators are worn may be initially estimated by dividing the ambient airborne concentration by the protection factor specified in Table 6.7-1 attached hereto for the respiratory protective equipment worn. If the intake of radioactivity is later determined by other measurements to have been greater than that initially estimated, the greater quantity shall be used in evaluating exposures; if it is less than that initially estimated, the lesser quantity may be used in evaluating exposures.
(c) The licensee advises each respirator user that he =ay leave the area at any time for relief from respirator use in case of equipment malfunction, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer. (d) The licensee maintains a respiratory protective program adequate to assure that the requirements of paragraphs 1 and 2 above are met. Such a program shall include: J (1) Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equip-ment. (2) Written procedures to assure proper selection, super-vision, and training of personnel using such protective equipment. (3) Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment for operability immediately prior to use. (4) Written procedures for maintenance to assure full effectiveness of respiratory protective equipment, in-cluding issuance, cleaning and decontamination, in-spection, repair, and storage. (5) Written operational and administrative procedures for proper use of respiratory protective equipment including provisions for planned limitations on working times as necessitated by operational conditions. J 6.7-2
(6) Bicassays and/or whole body counts of individuals,
- and other surveys, as appropriate, to evaluate in-N-- dividual exposures and to assess protection actually provided.
(7) Records sufficient to permit periodic evaluation of the adequacy of the respiratory protective program. I (e) The licensee uses equipment approved by the U. S. Bureau of Mines under its appropriate Approval Schedules as set forth in Table 6.7-1 below. Equipment not approved under U. S. Bureau of Mines Approval Schedules may be used only if the licensee has avaluated the equipment and can demonstrate i by testing, or on the basis of reliable test information, that the material and performance characteristics of the ] equipment are at least equal to those afforded by U. S. 1 Bureau of Mines approved equipment of the same type, as specified in Table 6.7-1 below. (f) Unless otherwise authorized by the Commission, the licensee-does not assign protection factors in excess of those specified in Table 6.7-1 below in selecting and using respiratory protective equipment. (g) These specifications with respect to the provisions of 20.103 shall be superseded by adoption of proposed changes to 10 CFR 20, Section 20.103, which would make this specifi-s,, cation unnecessary. 6.7.2 Exposure of individuals to concentrations of Argon-41 in the reactor building may be controlled in accordance with the dose limits and requirements of Section 20.101, instead of 20.103(b). l N%. 6.7-3 l N
e TABLE 6.7-1 em PROTECTION FACTORS FOR RESPIRATORS , PROTECTION FACTORS 2/ GUIDES TO SELECTION OF EQUIPMENT PARIICULATES BUREAl' OF MINES APPROVAL SCilEDULES* AND VAPORS AND FOR EQUIPMENT CAPABLE OF PROVIDING AT CASES EXCEPT LEAST EQJIVALENT PROTECTION FACTORS TRITIUM OXIDg3/ *or schedule superseding for equipment MODES 1/ of type listed DESCRIPTION
- 1. AIR-PURIFYING RESPIRATORS Facepiece, half-mask 4/ 2/ NP 5 218 30 CFR 14.4(b)(4)
NP 100 21B 30 CFR 14.4(b)(5); 14F 30 CFR 13 Facepiece, full Z/
- 11. ATMOSPilER E-SU t*Pl.YING RESPIRATOR
- 1. Airline respirator CF 100 19B 30 CFR 12.2(c)(2) Type C(1)
Facepiece, half-mask 12.2(c)(2) Type C(i) CF 1,000 19B 30 CFR Facepiece, full 12.2(c)(2) Type C(ii) 500 19B 30 CFR Facepiece, full J/ D 19B 30 CFR 12.2(c)(2) .; je 'C(iii) Facepiece, full PD 1,000 P' CF _5_/ See note 6/ 7 Ilood CF 5/ See note 6/ Suit
- 2. Self-contained breathing apparatus (SCBA)
Facepiece, full J/ D 500 13E 3C CFR 11.4 (b)(2)(i) Facepiece, full PD 1,000 13E 30 CFR 11.4 (b)(2)(ii) R 1,000 13E 30 CFR 11.4 (b) (1) Facepiece, full III. COMBINATION RESPIRATOR 19 B CFR 12.2(e) or applicable Any combination of air- Protection factor for purifying and atmosphere- type and mode of opera- schedules as listed above supplying respirator tion as listed above 1/, 2/, 3/, 4/, 5/, 6/, J/, [These notes are on the following pages]
j , o l 1/ See the following symbols:
\'- CF: continuous flow D: demand NP: negative pressure (i.e., negative phase during inhalation)
PD: pressure demand (i.e., always positive pressure) PP: positive pressure R: recirculating (closed circuit) 2/ (a) For purposes of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radio-active material outside the respiratory protective equipment to that inside the equijaent (usually inside the facepiece) under conditions of use. It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following formula: i Ambient Airborne Concentration L Concentration Inhaled = Protection Faerer
)
y (b) The protection factors apply: (1) only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program. x,, (ii) for air-purifying respirators only when high efficiency [above 99.9% re= oval efficiency by U. S. Bureau of Mines type dioccyl phthalate (DOP) test] particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen. (iii) for atmosphere-supplying respirators only when supplied with adequate respirable air. 3/ Excluding radioactive contaminants that present an absorption or sub-mersion hazard. For tritium oxide approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide. See also footnote 5/ below, concerning supplied-air suits and - hoods. if Under chin type only. Not .acommended for use where it might be possible for the ambient airborne concentration to reach instantaneous values greater than 50 times the pertinent values in Appendix 3, Table I, Column 1 of 10 CFR, Part 20. N.- 6.7-5 l
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