ML19316A087

Tech Specs,App a

ML19316A087
Person / Time
Site:	Oconee
Issue date:	02/06/1973
From:	DUKE POWER CO.
To:
References
NUDOCS 7911280591
Download: ML19316A087 (200)

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{{#Wiki_filter:9 = Appendix A Technical Specifications For Oconee Nuclear Station Unit No.1 Duke Power Compny Seneca, South Carolina Docket No. 50-269 1 DUKE POWER 1 i-a

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t 4 TABLE OF CONTENTS Section Page TECHNICAL SPECIFICATIONS 1 DEFINITIONS 1-1 1.1 RATED P0h'ER l-1 1.2 REACTOR OPERATING CONDITIONS 1-1 1.2.1 Cold Shutdewn 1-1 1.2.2 Hot Shutdown 1-1 1.2.3 Reactor control 1-1 1.2.4 Hot Standhv l-1 1.2.5 Power Goeration 1-1

1.2.6 Refuelin

Shutdcun 1-1 1.2.7 "efuelin2 Oper?. tion 1-2 1.2.8 Refueling Period 1-2 1.2.9 Startuo 1-2 1.3 OPERABLE l-2 1.1 PROTECTIVE INSTRCMENTATION LOGIC 1-2 1.4.1 Instrument Channel 1-2 1.4.2 Reactor Prctective Svstem 1-2 1.4.3 Protective Channel 1-2 1.4.4 Reactor Protective Svstem Locic 1-3 1.4.5 Encineered Safetv Features Svstem 1-3 1.4.6 Decree of Redundance 1-3 1.5 INSTRUMENIATION SUR'IEILLANCE 1-3 1.5.1 Trio Test 1-3 1.5.2 Channel Test 1-3 1.5.3 Instrument Channel check 1-3 i

.e s Section Page 1.5.4 Instrument channel Calibration 1-3 1.5.5 Heat 3alance Check 1-4 1.5.6 Heat Balance Calibration 1-4 4 1.6 QUADRANT POWER TILT l-4 t 1.7 CONTAINMENT INTEGRITY l-4 1.8 A3 NORMAL OCCURRENCE l-4 1.9 UNUSUAL EVENTS 1-5 2 SAFETY LU!ITS AND LHiITING SAFETY SYSTDi SETTINGS 2.1-1 2.1 SAFETY LIMITS, REACTOR CORE 2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE 2.2-1 i 4 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE 2.3-1 INSTRUMENTATION 4 3 LIMITING CONDITIONS FOR OPERATION 3.1-1 3.1 REACTOR COOLANT SYSTEM 3.1-1 3.1.1 Geerational Conconents 3.1-1 3.1.2 Pressurization. Heatuo, and Cooldown Limitations 3.1-3 3.1.3 Minimum Conditions for Criticality 3.1-8 3.1.4 Reactor Coolant Svstes Activitv 3.1-10 3.1.5 Chemistrv 3.1-12 3.1.6 Leakage 3.1-14 3.1.7 Modarator Temeerature Coefficient of Reactivity 3.1-17 3.1.8 Single Loon Restrictions 3.1-19 3.1.9 Low Power Physics Testing Restrictions 3.1-20 l i 3.1.10 control Rod coeration 3.1-21 i 3.2 HIGH PRESSURE INJECTION AND CHDiICAL ADDITION SYSTDIS 3.2-1 3.3 DIERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR 3.3-1 BUILDING SPRAY, AND PENETRATION ROOM VESTILATION SYSTDiS 'il i ~. -, - - ,v., ~, -. - -,,, - - -

s Section Pace 3.4 STEAM AND PCNER CO:NERSION SYSTE! 3.4-1 3.5 INSTRUMENTATION SYST2iS 3.5-1 3.5.1 Goerational S..?etr Instrumentation 3.5-1 3.5.2 Control Rod Group and Power Distribution Limits 3.5-6 3.5.3 Enzineered Safet7 Features Protective Svsten 3.5-9 Actuation Setcoints 3.5.4 Incora Instrumentation 3.5-11 3.6 REACTOR SUILDING 3.6-1 3.7 AUXILIARY ELECTRICAL SYSTCIS 3.7-1 3.8 FUEL LOADING AND REFUELING 3.3-1 3.9 RELEASE OF LICUID RADICACTIVE WASTE 3.9-1 3.10 RELEASE OF GASEQUS RADI0 ACTIVE WASTE 3.10-1 3.11

1AXIMUM POWER REEIRICTIONS 3.11-1 3.12 REACTOR 3UILDING ?OLAR CPANE AND AUXILLiRY HCIST 3.12-1 3.13 SECONDARY SYSTDI ACTIVITY 3.13-1 4

SURVEILLANCE STANDARDS 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 3YSTDI 4.3-1 4.4 REACTOR SUILDING 4.4-1 4.4.1 Containment Leakage Tests 4.4-1 4.4.2 Structural Integritv 4.4-7 4.4.3 Hvdrogen Purze Svsten 4.4-11 l 4.5 DIERGENCY CORE C0 CLING SYSTE!S AN REACTCR SUILDING 4.5-1 COOLING SYSTE!S PERICDIC TESTING ] 4.5.1 Enervence Core Ccoling Sectens a.3-1 l't

Section Page 4.5.2 Reactor Building Cooling Systems 4.5-4 4.5.3 Penetration Room Ventilation Svstem 4.5-7 4.5.4 Low Pressure Injection Svstem 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 Procram Verification (Grous vs. Core 4.7-3 Position) 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 SUFVEILLANCE 4.13-1 4.14 REACTOR BUILDING PURGE FILTERING SYSTEM 4.14-1 4.15 IODINE RADIATION MONITORING FILTERS 4.15-1 5 DESIGN FEATURES 5.1-1 5.1 SITE 3.1-1 5.2 CONTAINMENT 5.2-1 5.3 REACTOR 5.3-1 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5.4-1 6 ADMINISTRATIVE CONTROLS 6.1-1 6.l' ORGANIZATION, REVIEW, AND AUDIT 6.1-1 6.1.1 Oraanization 6.1-1 .6.1.2 Review and Audit 6.1-2 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN ABSORMAL OCCURRENCE 6.2-1 iv

a e b Section Pace 6.3 ACTION TO 3E TAKEN IN TiiE EVENT A SAFETY LD1IT IS ENCEEDED 6.3-1 6.4 STATION OPERATING PROCEDURES 6.4-1 6.5 STATION OPERATING RECORDS 6.5-1 6.6 STATION REPORTING REQUIRDIENTS 6.6-1 6.6.1 Routine Recorts 6.6-1 6.6.2 Non-Routine Reports 6.6-6 6.6.3 Special Reports 6.6-7 6.7 RADIOLOGICAL CONTROLS 6.7-1 1 v

a e o LIST OF TABLES Table No. Pane 2.3-1 Reactor Protective System Trip Setting Limits 2.3-7 3.5.1-1 Instrument operating Conditions 3.5-3 4.1-1 Instrument Surveillance Requirements 4.1-3 4.1-2 Minimum Equipment Check Frequency 4.1-8 4.1-3 Minimum sampling Frequency 4.1-9 4.11-1 Environ = ental Surveillance Program 4.11-2 6.1-1 Minimum Operating Shift Requirements 6.1-9 6.6-1 Report of Radioactive Effluents 6.6-9 1 vi

O 9 e LIST OF FIGURES Figure Pace 2.1-1 Core Protection Safety Limits 2.1-4 2.1-2 Core Protection Safety Limits 2.1-5 2.1-3 Core Protection Safaty Limits 2.1-6 2.3-1 Protective System Maximum Allowable Set Points 2.3-5 2.3-2 Protective Systen Maximus Allowable Set Points 2.3-6 3.1.2-1 Reactor Coolant System Heatup Limitations 3.1-6 3.1.2-2 Reactor Coolant System Cooldown Limitations 3.1-7 3.5.4-1 Incore Instrumentation Specification Axial 3.5-13 Imbalance Indir ation 3.5.4-2 Incore Instrumentation Specification Radial 3.5-14 Flux Indication 3.5.4-3 Incore Instrumentation Specification 3.5-15 6.1-1 Station Organization Chart 6.1-7 6.1-2 Management Organization Chart 6.1-8 vii

e INTRODUCTION These Technical Specifications apply to the Oconee Nuclear Station, Unit 1 and are in accordance with the requirements of 10 CFR 50, Section 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 intent of the spec-ificatien. These bases are not part of the Technical Specifications, and they do not constitute limitations or requirements for the licensee. viii

s TECHNICAL SPECIFICATIONS 1 DEFINITIONS The following terms 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 MWt. 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 T is no more than 2000F. Pressure is defined by avg Specification 3.1.2. 1.2.2 Hot Shutdown The reactor is in the hot shutdcwn condition when it is subcritical by at least 1 percent ak/k and Tavg is at or greater than 530F. 1.2.3 Reactor Critical The reactor is critical when the neutron chain reaction is self-sustaining and Keff = 1.0. 1.2.4 Hot Standby The reactor is in the hot standby condition when all of the following conditions exist: Tavg is greater than 5250F. a. 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 Ooeration The reactor is in a power operating condition when the indicated neutron power is above 2 percent of rated pcwer as indicated on the power range channels. 1.2.6. Refueling Shutdown The reactor is in the refueling shutdown condition when,even with all rods recoved, the reactor would be suberitical by at least 1 percent ak/k and the coolant temperature at the low pressure injection pu=p suction is no more than 140 F. Pressure is defined by Specificatica 3.1.2. A refueling shutdown re-fers to a shutdown to replace or rearrange all or a portion of the fuel assem-blies and/or control rods. 1-1

1.2.7 Refueling Operation - g An operation involving a change in core geometry by manipulation of fuel or control rods when the reactor vessel head is removed. 1.2.8 Refueling Period Time between normal refuelings of the reactor, not to exceed 18 months without prior approval of the AEC. 1.2.9 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical. 1.3 OPERABLE A component or system is operable when it is capable of performing its intended function within the required range. The component or system shall be considered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Specification 3, and (2) it has been tested periodically in accordance with Specification 4 and has met its performance requirements. 1.4 PROTECTIVE INSTRWENTATION LOGIC 1.4.1 Instrument Chennel ) .) An instrument channel is the combination of sensor, wires, amplifiers and output devices which are connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/or protection. An instrument channel may be either analog or digital in nature. 1.4.2 Reactor Protective Svstem The reactor protective system is shown in Figures 7-1 and 7-6 of the FSAR. It is that combination of protective channels and associated circuitry which forms the automatic syste= that protects the reactor by control rod trip. It includes the four protective channels, their associated instrument channel inputs, manual trip switch, all rod drive protective trip breakers and activating relays or coils. 1.4.3 Protective Channel 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, amplifiers and bistable modules provided for every reactor protective safety parameter) is a combination of instrument channels forming a single digital output to the protective system's coincidence logic. It includes a shutdown bypass circuit, a protective channel bypass circuit and reactor trip module and provision for insertion of a dummy bistable. 1-2

1.4.4 Reactor Protective System Logic This system utilizes reactor trip module relays (coils and contacts) in all four of the protective channels as shown in Figure 7-1 of the FSAR, to pro-vide reactor trip siguals for de-energizing tne six control rod drive trip breakers. The control rol drive trip breakers are arranged to provide a one out of two times two logic. Each element of the one out of two times two logic is controlled by a separate set of two out of four logic contacts from the four rea-protective channels. 1.4.5 Engineered Safety Features System This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as shown in Figure 7-3 of the FSAR. 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. 1.4.6 Degree of Redundancy s '- The'cifference between the number of operable channels and the number of channels which when tripped will cause an automatic system trip. l 1.5 INSTRCIENTATION SURVEILLANCE 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 channels measuring the same variable. 1.5.4 Instrument Channel Calibration An instrument channel calibration la a test, and adjustment (if necessary), to establish that the channel output responds alth acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include the channel test. 1-3

4 s 1.5.5 Heat Balance Check ' ~h A heat balance check is a comparison of the indicated neutron power and core j thermal power. a { l.5.6 Heat Balance Calibration An adjustment of the power range channel amplifiers output to agree with the core thermal power as determined by a heat balance on the secondary side of the steam generator considering all heat losses and additions. 1.6 QUADRANT POWER TILT 4 Quadrant to average power tilt is expressed in percent as defined by the following equation: 1, I / Power in any core quadrant _ \\g verage for all quadrants A l To obtain quandrant 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. 'T 1.7 CONTAIMiENT INTEGRITY / Containment integrity exists when the following conditions are satisfied: 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. i b. At least one door on each of the personnel hatch and emergency hatch is closed and sealed during refueling or personnel passage through these hatches. f c. All non-automatic containment isolation valves and blind flanges are closed as required. j d. All automatic containment isolation valves are operable or locked 4 closed. e. The containment leakage determined at the last testing interval satisfies Specification 4.4.1. 1.8 ABNORMAL OCCURRENCE An abnormal occurrence means the occurrence of any-plant condition that: s l-4

a. Results in a protective instrumentation setting in excess of a Limiting Safety System Setting as established in the Technical Specifications, or b. Exceeds a Limiting Condition for Operation as established in the Technical Specifications, or Causes, any significant uncontrolled or unplanned release of c. radioactive material from the site, or d. Results in abnormal degradation of one of the several boundaries which are designed to contain the radioactive materials resulting frca the fission process, or e. Results in uncontrolled or unanticipated changes in reactivity greater than 1% ak/k except for trip. 1.9 UNUSUAT. EVENTS An unusual event is: (a) Discovery of any substantial errors in the transient or accident analyses, or in the methods used for such analyses, as described in the Safety Analysis Report or in the bases for the Technical F ecifications. (b) Any substantial variance from performance specifications contained in the Technical Specifications or the Safety Analysis Report. (c) Any observed inadequacy in the implementation of administrative or procedural controls during the operation of the facility which could significantly affect the safety of operations. (d) Any occurence resulting in an Engineered Safety System or Reactor Protective System component malfunction or system or component malfunction which could render a safety system incapable of performing its intended safety function. (e) Any occurrence arising from natural or offsite man-made events that affect or threaten to affect the safe operation of the plant. 1-5

~ a 2 SAFETY LDfITS AND LIMITING SAFETY SYSTDI SE'TZNGS 2.1 SAFETY LIMITS, REACTOR CORE Apolicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant. Obiective To maintain the integrity of the fuel cladding. Specification The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1. If the actual pressure / temperature point is below and to the right of the line, the safety limit is exceeded. The combination of reactor thermal power and reactor power imbalance (power fraction in top half of core minus power fraction in the bottom half of the core) shall not exceed the safety limit as defined by the locus of points (solid line) for t specified flow set forth in Figure 2.1-2. If the 5 ' actual-reactor-thermal-power / reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded. Bases To maintain the integrity of the fuel cladding and to prevent fissien product release, it is necessary to prevent overheating of the cladding under normal operation conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is 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" (DNB). 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 power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the W-3 correlation.fl) ~ The W-3 correlation has been d'eveloped to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the' actual heat flux, is indicative of the margin to DNB. The minimum value of the 7NBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.3. A DNBR 1 of 1.3 corresponds to a 94.3% probability at a 99% confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core-protection safety limits. The difference in these two pressures is nominally 45 psi;- however,' only a 30 psi drop was assumed in 2.1-1

reducing the pressure trip set pointo to correspond to the elevated location g s where the pressure is actually measured. / The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.3 is predicted for the maximum possible thermal power (114%) when four reactor coolant pumps are operating (rcactor coolant flev is 131.3 x 106 lbs hr). This curve is based on the following nuclear power peaking factors (2 ; N N N F 2.67; FAH = 1.78; F 1.50 = = q 2 The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion and are the core DNBR design bssis. The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits: 1. The 1.3 DNBR limit produced by a nuclear power peaking factor of FN = 2.67 or the combination of the radial peak, axial peak and p0sitionoftheaxialpeakthatyieldsnolessthana1.3DNBR. 2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 22.2 kw/ft. Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking. The specified flow rates for curves 1, 2, 3, and 4 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, one pump in each loop and two pumps in one loop, respectively. The curve of Figure 2.1-1 is tne most restrictive of all possible reactor coolant pump--maximum thermal power combinations shawn in Figure 2.1-3. The curves of Figure 2.1-3 represent the conditions at which a minimus DNBR of 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 15%,(3) whichever condition is more restrictive. Using a local quality limit of 15% at the point of minimum DNBR as a basis for curves 2 and 4 of Figure 2.1-3 is a conservative criteron even though the quality of the exit is higher than the quality at the point of minimum DNBR. The DNBR as calculated by the W-3 correlation continually increases from point of minimum DNBR, so that the exit DNBR is 1.7 or higher, depending on the pressure. Extrapolation of the W-3 correlation beyond its published quality range of +15% is justified on the basis of experimental data.(4) The maximum thermal power for three pump operation is 88.5% due to a pcwer level trip produced by the flux-flow ration (75% flow x 1.1 - 82.5% power) plus the maxi =um calibration and instrumentation error. The maximum ther=al power for other reactor coolant pump conditions are produced in a similar manner. 2.1-2

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 DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 15% for that particulcr reactor coolant pump situation. The 1.3 DNBR curve for four pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of the four pump curve will be above and to the left of the other curves. REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) FSAR, Section 3.2.3.1.1.c (3) FSAR, Section 3.2.3.1.1.k (4) The following papers which were presented at the Winter Annual Meeting, ASME, November 18, 1969, during the "Two-phase Flow and Heat Transfer in Rod Bundles Symposium": (a) Wilson, et.al. " Critical Heat Flux in Non-Uniform Heater Rod Bundles." (b) Gellerstedt, et.al. " Correlation of a Critical Heat Flux in a Bundle Cooled by Pressurized Water." i i i 2.1-3 1

i -s 2600 2400 na 3 2200 r a; E a. 2000 5 = ~ 3 / ^ 1800 / 1600 560 580 600 620 640 660 Reactor Outlet Temperature, F CORE PROTECTION SAFETY LIMIT '\\ i q OCONEE NUCLEAR STATION Figure 2.1 - 1 2.1-a

Tmm'AL Fwia LE/EL, 3 f 120 / f' - DNBR Limit ~~ mt ,[ (1) 80 (2) 60 (3 ) (4) 40 L i i t i i -60 -40 -20 0 +2 0 +4 0 +60 Reactor Power Imualance, 's i CURVE REACTOR COOLANT FLOW (LB/HR) 1 131.3 x 106 2 98.1 x 106 3 64.4 x 106 4 60.1 x 106 CORE PROTECil0N SAFETY LIMITS .b\\ Pi rcat s, OCONEE NUCLEAR STATION M' Figure 2.1 2 2.1-5 ~

s' e' 2600 3 2400 4 2200 E d 5 ) y 2000 / a. O; E b 1800 '7 // D 1600 560 580 600 620 640 660 Reactor Outlet Temperature F REACTOR COOLANT FLOW CURVE (L8S/HR) POWER PUMPS OPERATING (TYPE OF LIMIT) 1 131.3 x 106 (1005 ) 114", FOUR PUMPS (DNBR LIMIT) 2 60.1 x 106 (45.85) 665 TWO PUMPS 'N ONE LOOP (QUAI.lTY LIMIT) 3 98.1 x 106 (74.75) 88.55 THREE PUMPS (DNBR LIMIT) 4 64.4 x 106 (49, g,) 605 ONE PUMP IN EACH LOOP (QUALITY LIMIT) CORE PROTECil0N SAFETY LIVITS ..J 'h i\\ OCONEE NUCLEAR STATION Figure 3.2.1 - 3 i

o 2.2 SAFETY Lri1TS - REACTOR C001E T SYSTEM PRESSURE Aoplicability, Applies to the limit on reactor coolant system pressure. Obiective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity. Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel. 2.2.2 The setpoint of the pressurizer code safety valves shall be in acccrdance with ASME, Boiler and Pressurizer Vessel Code, Section III, Article 9, Summer 1967. Bases The reactor coolant system (l) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME code, Section III, is 110% of design pressure.(2) The maximum 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 settings, the reactor high pressure trip (2355 psig) and the pressurizer safety valves (2500 psig)(3) have been established to assure never reaching the reactor cociant 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 safety limit is provided by setting the pressurizer electromatic relief valve at 2255 psig. REFERENCES (1) FSAR, Section 4 (2) FSAR, Section 4.3.10.1 (3) FSAR, Section 4.2.4 2.2_1

2.3 LD11 TING SAFETY SYSTE 1 SETTINGS, PROTECIlVE INSTRDIENTATION Aeolicability Applies to instruments nonitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure. Obiective To provide automatic protective action to prevent any combination of process variables from exceeding a safety limit. Specification The reactor protective systea trip setting linits and the permissible by-passes 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: Loss of two pumps and reactor power level is greater than 55% a. of rated power. b. Loss of two pumps in one reactor ecolant loop and reactor power level is greater than 0.0% of rated power. (Reactor power level trip setpoint is reset to 55% of rated power for single loop operation.) c. Loss of one or two famps during two pump operation. Eases The reactor protective system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to 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 errors. Nuclear Overoower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements. During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 107.5% of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actu-ated could be 114%, which was used in the safety analysis.(E) 2.3-1

s Overpower Trio Based on Flow and Imbalance, / The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to acco=modate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. The analysis in Section 14 demonstrates that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any electrical malfunction. The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increcses or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB pro-tc: tion for all modes of pump operation. For every flow rate there is a maxi-mum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows: 1 1. Trip would occur when four reactor coolant pumps are operating if power is 110% and reactor flow rate is 100%, or flow rate is 90.9% and power level is 100%. 2. Trip would occur when three reactor coolant pumps are operating if pcwer is 82.5% and reactor flow rate is 75%, or flow race is 68.2% and power level is 75%. m I 3. Trip would occur when two reactor coolant pumps are operating in a single loop if power is 55.0% and the operating loop flow rate is 50.0% or flow rate is 45.5% and power level is 50%. 1 4. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 54.0% and reactor flow rate is 49.0% or flow rate is 45.5% and the power level is 50%. For safety calculations t'..e maximum calibration and instrumentation errors for the power level were used. The power-imbalance boundaries are established in order to prevent reactor } thermal limits from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio such that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbalance boundaries by 1.10% for a 1% flow reduction. Pumo Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio. The pump monitors also restrict the power level for the number of pumps in operation. l 2.3-2 I l

Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from 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 transient.(1) The low pressure (1800 psig) and variable low pressure (13.26T - 5989) trip out setting limits shown in Figure 2.3-1 have been established to maintain the DN3 ratio greater than or egual to 1.3 for those design accidents that result in a pressure reduction.(2,3) Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip set point of (13.26T -6029). out Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F. Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip. Shutdown Evnass In order te provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shcan in Table 2.3-1. Two conditions are imposed when the bypass is used: 1. By administrative control the nuclear overpower trip set point must be reduced to a value 5,5.0% of rated power during reactor shutdown. 2. A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed. i The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed. This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The over power trip set point of 5,5.0% prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circu-lation(5) would be available to remove 5.0% of rated power if none of the reactor coolant pumps were operating. 2.3-3

's _S_ ingle Loop Operation. Single loop operation is permitted only after the reactor has been tripped. After 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 one of the two protective channels receiving outlet temperature information from sensors in the idle loop. Tripping one of the two protection channels receiving outlet temperature information from the idle loop assures a protective system trip logic of one out of two. REFERENCES (1) FSAR, Section 14.1.2.2 /;) FSAR, Section 14.1.2.7 (3) FSAR, Section 14.1.2.8 (4) FSAR, Section 14.1.2.3 'N (5) FSAR, Section 14.1.2.6 ,\\ 2.3-4

4 2500 en 2300 a 2, m A C# '00 C O O 1900 O a un v m 1700 1500 t i i 5% 560 580 600 620 6'40 Reactor Outlet Temperature, F PROTECTIVE SYSTEM MAXIMUM ALLO 1ABLE SET POINTS bh '.ocat ecau. OCONEE NUCLEAR STATION Y Figure 2.3 - 1 4.3-3

' PGTER LE/IL, fa ) -- 120 Faur Pump Set Points - - 100 Three Pump 80 Set Points 60 (2)x Tea Pump Set Points 40 (1) ONE PUMP IN EACH LOOP - (2) TWO PUMPS IN ONE LOOP 20 e i J -40 -20 0 +20 +40 +60 Pane r Imca lanc e, '. PROTECTIVE SYSTEM MAXIMUM AlL0 FABLE SET POINTS l l cat e an% . OCONEE NUCLEAR STATION Figure 2.3 2 2.3-6

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3 LD1ITING CONDITIONS FOR OPERATICN 3.1 REACTOR C00I. ANT SYSTDi Applicability Applies to the operating status of the reactor coolant syste=. Obiective To specify those limiting conditions for operation of the reactor coolant system components which must be met to ensure safe reactor operation. Soecification 3.1.1 Operational Components a. Reactor Coolant Pumps 1. Whenever the reactor is critical, single pump operation shall be prohibited, single loop operation shall be restricted to testing, and ot".ter pump combinations permissible for given power levels shall be as shown in Table 2-3.1. 2. The boran concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one low pressure injection pump is circulating reactor coolant. b. Steam Generator 1. One steam generator shall be operable whenever the reactor coolant average temperature is above 250 F. c. Pressurizer Safetv Valves 1. All pressurizer code safety valves shall be operable 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 Boiler and Pressure Vessel Code. Bases A reactor coolant pump or low pressure inj ection pump is required to be in operation 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) 3.1-1

= o. I i i The low pressure injection system suction piping is designed for 3000F ') i and 370 psig; thus the system with its redundant c sponents can remove decay heat when the reactor coolant system is below this temperature. (2,3) 1 One pressurizer code safety valve is capable of preventing overpressurization when the reactor is nct critical since its relieving capacity is greater than that required by the ses of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. (4) Both pressurizer code safety i valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure i i for a rod withdrawal accident at hot shutdown. (5) The pressurizer code safety valve lif t set point shall be set at 2500 psig i 1% allowance for error,and 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. REFERENCES j (1) FSAR Tables 9-11 and 4-3 through 4-7. (2) FSAR Sections 4.2.5.1 and 9.5.2.3. (3) FSAR Section 4.2.5.4. (4) FSAR Sections 4.3.10.4 and 4.2.4. (5) FSAR Sections 4.3.7 and 14.1.2.2.3. ) i i i i 's 3.1-2

3.1.2 Pressurization. Heatuo. and Cooldew, Limitations Soecification 3.1.2.1 Hydro Tests For thermal steady state system hydro test the system may be pressurized to the limits set forth in Specification 2.2 when there are fuel as-semblies in the core and to ASME Code Section III limits when no fuel assemblies are present provided: Prior to initial criticality the reactor coolant system a. temperatuiu is ll80F or greater or b. After initial criticality and during the first two years of operation the reactor coolant system temperature is 215 F or greater. 3.1.2.2 Leak Tests Leak tests may be conducted under the provisions of 3.1.2.1 a and a. b above or b. After initial criticality and during the first two years of operation the system may be tested to a pressure of 1150 psig provided that the system temperature is 175 F or greater. 6 thermal 3.1.2.3 For the first two years of power operation (1.7 x 10 megawatt days) the reactor ecolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance 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 below the linit line in Figure 3.1.2-1. The heatup rates shall not exceed those shown on Figure 3.1.2-1. Cooldown: Allowable combinations of pressure and temperature far a specific cooldown shall be to the left of and below the limit line in Figure 3.1.2-2. Cooldown rates shall not exceed those shoun 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 belew 1000F, 3.1.2.5 The preasurizer heatup and cooldewn rates shall not exceed 100'F/hr. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 4100F. 3.1.2.6 Within two years of power operation, Figures 3.1.2-1 and 3.1.2-2 shall be updated in accordance with appropriate criteria accepted by the AEC. 3.1-3

m Bases ) All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. (1) These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR. The maximum 0 unit heatup and cooldown rate of 100 F per hour satisfies stress limits for cyclic operation. (2) The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 1000F satisfies stress levels for temperatures below the DTT. (3) The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maxicum NDTT value of 200F has been determined based on Charpy V-Notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 400F. I Figures 3.1.2-1 and 3.1.2-2 contain the limiting 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 thei FSAR. As a result of fast neutron 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 40-year exposure is shown on Figure 4.10:(4) The actual shift in NDTT will be determined periodically during plant opera-tion by testing of irrt.diated vessel material samples located in this reactor vessel.(5) The results of the irradiated sample testing will be 1 evaluated and compared to the design curve (Figure 4-11 of FSAR) being used j to predict the increase in transition temperature. Thedesignyaluefgrfastneutron(E>1MeV)exposureofthereactorvessel 0 n/cm{cm--sat 2,568MWtratedpowerandanintegratedexposureof i is 3.0 x 10 n i 3.0 x 1019 for 40 years operation. (6) The calculated maximum values 2 1 are 2.2 x 1010 n/cm --s and 2.2 x 1019 n/cm-integrated exposure for 40 years operation at 80 percent load. (4) Figure 3.1.2-1 is based on the design a I value which is considerably higher than the calculated value. The DTT value l for Figure 3.1.2-1 is based on thm projected NDTT at the end of the first two years of operation. During these two years, the energy cutput has been 6 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 fast neutron exposure of the reactor vessel for the two years is 1.7 x 1018 n/cm2 which is based on the 1.7 x 106 thermal megawatt days and the design value for fast neutron exposure. The actual shift 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.- 3.1-4

The NDTT shift and the magnitudes of the thermal and pressure stresses are sensitive to integrated reactor power and not to instantaneous power level. Figure 3.1.2-1 and 3.1.2-2 are applicable to reactor core thermal ratings up to 2,568 MWt. The pressure limit line on Figure 3.1.3-1 has been selected such that the reacter 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. Maximum differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pump ccmbinations. For adequate conservatism in fractare taughness including size '. thickness) effect, a maximum pressure of 550 psig balow 2750F with a maximum heatup and cooldown rate of 500F/hr has been imposed for the initial tvo year period as shown on Figure 3.1.2-1. During this two year period, a fracture toughness criterion applicable to Oconee Unit 1 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 irrad'ated and unirradiated, f or the AEC-HSST program and the PVRC program, and t.f th considerations of test results of the Oconee Unit 1 reactor surveillance program. The spray temperature difference restriction is imposed to maintain the ther.al stresses at the pressurizer spray line nozzle below the design Temperature requirements for the steam generator correspond with the limit. measured NDTT for the shell. REFERENCES (1) FSAR Section 4.1.2.4 (2) ASME Boiler and Pressure Code, Section III, N-415 (3) FSAR Section 4.3.10.5 (4) FSAR Section 4.3.3 (5) FSAR Section 4.4.6 (6) FSAR Sections 4.1.2.8 and 4.3.3 3.1-5

POINT TEMP. PRESS. A 40 550 B 275 550 C 275 1400 2400 0 380 2275 2200 2000 a 1800 S 5 - UPPER PRESSURIZATION a LIMIT 14I0 <-C E E 1200 2 h E J 1000 a O 0 800 2 3 600 j A' 400 MAXIMUM HEATUP RATE,*F/HR 50 100 200 = = I I I I I 0 O 100 200 300 400 500 275 Indicated Reactor Coolant System Temperature.'F f s\\ w; OCONEE NUCLEAR STATION REACTOR COOLANT SYSTEM HEATUP LIMITATIONS ( A FPLICABLE UP TO AN INTEGRATE 0 EXPOSURE ~ 0F 1.7 x 10 38 n/cm2 OR DTT 144 'F) l F i gu r e 3.1.2 - 1 3.1-6

4 = PolNT TEMP PRESS A 350 2275 8 275 1400 C 275 550 0 250 550 E 250 450 F 175 450 2400 F / G 175 200 1 3 H 120 200 2200 RC PUMP C093tN AT10NS ALL0s ABLE. ABOVE 185F ALL 1 1.1 -B j 0 A.2 -B;1 - A.0-8;0 A,1 8 y 2000 BEL 0s IB5F (1)

• HEN DECAY HE AT REWOVAL SYSTEM (OH) 15 j1800 CPERATING tlTH00T ANY RC PU4PS OPERATING, INDICATED CH RETURN TEMP. TO THE REACTOR 3r 1600 VESSEL SH ALL SE USEO.

S / (2) IN THE TEMPERATURE RANGE 250F TC 175F. A e 5 1400 WA114;w STEP T!WPERAT::RE CHANGE OF 75F y IS ALL0 TABLE FOLL0 SED SY A CNE HCUR 3 1200 - uisivty HOLO CN TEMPERATURE. IF THE STEP CHANGE IS TAKEN BELO4 250F RC TEMPERATURE, THE M A114LM ALLO *ABLE STEP SH ALL BE TH AT 1000 WHICH YlELOS A FIN AL TEV8ER ATURE OF 115F, y THE STEP TEMPERATURE CHANGE IS DEFlhEO AS J 800 - RC TE*PERATURE(BEFORE ST0FF1hG ALL RC PUMPS) o WINUS THE CH 4ETURN TEMPERATURE TO THE REACTOR 5 600 VESSEL. C, I -F s 400 p 4 E g g / 200 UPPER PRESSURI Z ATION.__ ll"lI 0 MAllMUM C00L00nN RATE. 'F HR - ( 2 ).- l I 100 l 50 ', = I i i., i 6 I-260 6 175 la i 530 275 600 500 400 300 230 100 incicates Reactor Coolant System Temaer 3ture *F( U REACTOR COOLANT STSTEW COOL 004N LlWITATIONS ( APPLICAdLi UP TO OTT = 135'F) .\\

71. Pw a. OCONEE NUCLEAR STATION Figure 3.1.2 2

3.1 3.1.3 Minimum Conditions for Criticality Specification 3.1.3.1 The reactor coolant temperature shall be above 5250F except for portions of low power physics testing when the requirements of specification 3.1.9 shall apply. 3.1.3.2 Reactor coolant temperature shall be above DTT + 100F, 3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified in 3.1.3.1 above, 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. 3.1.3.4 The reactor shall be maintained suberitical by at least 1% Ak/k until a steam bubble is formed and a water level between 80 and 396 inches is established in the pressurizer. Bases At the beginning of the initial fuel cycle, the ecderator temperature co-efficient is expected to be slightly positive at operating temperatures with the operating configuration of control rods. l') Calculations show that above 525 F, the consequences are acceptable. Since the moderator temperature coefficient at lower temp negativeormorepositivethanatoperatingtemperature,gratureswillbeless

2) startup and eperation of the reactor when reactor coolant temperature is less than 525 F is prohibited except where necessary for low power physics tests.

The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1% ak/k. During physics tests, special operating ( recautions will be taken. In addition, the strong negative Doppler coefficient ) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density. The requirement that the reactor is not to be made critical below DTT + 10*F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NDIT of the primary coolant system. Heatup to this temperature will be acccmplished 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. 3.1-8

e The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% suberitical will assure that the reactor coolant system cannot become solid in the event of a red withdrawal accident 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 '\\) i I 3.1-9 4 l

~_ l t i 1 3.1.4 Reactor Coolant Mystem Activity 4 Soecification The total activity of the reactor coolant due to nuclides with half lives longer than 30 minutes -hall not exceed 224/E microcuries per ml whenever j the reactor is crir* E is the average (mean) beta and ga=ma energies j per disintegration, in MeV, weighted in proportion to the measured activity of the radionuclides in reactor coolant samples. i j Bases The above specification is based on limiting the consequences of a postulated accident involving the double-ended rupture of a steam generator tube. The i l rupture of a steam generator tube enables reactor coolant and its associated j activity to enter the secondary system where volatile isotopes could be dis-charged to the atmosphere through condenser air-ejectors and through steam safety valves (which may lift momentarily). Since the major portion of the activity entering the secondary system is due to noble gases, the bulk of the activity would be discharged to the atrasphere. The activity release continues until the operator stops the leakage by reducing the reactor 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 N16 detectors on the steam lines in conjunction with the off-gas 3 monitors on the condenser air ejector lines; thus he can isolate the faulty i steam generator within 34 minutes after the tube break occurred. During that 34 minute period, a maximum of 2760 ft3 of 580 F reactor coolant leaked into the j secondary system. (This is equivalent to a cold makeup volume of 1980 ft3). I 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 cond(nser cooling water flow can be maintained by gravity flow from Lake Keowee through the emergency condenser cooling water discharge to the Keewee Hydro ta11 race. The controlling dose for the steam generator tube rupture accident is the whole-body dose resulting from immersion in the cloud of released activity. 1 4 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 l tube rupture accident occur. i l Although only volatile isotopes will be released from the secondary system, the following whole-body dose calculation conservatively assumes that all of the radioactivity which enters the secondary system with the reactor coolant l 1s released to the atmosphere. -Both the beta and gama 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 t potential ga=ma dose. The semi-infinite. cloud model is applicable to the beta dose because of the short range of beta radi. ion in It is further assumed that meteorological conditions du-ing the ( of the accident correspond to Pasquill Type F and 1 meter per seconc -d speed, resulting in a 3 X/Q value of l.16 x 10-4 sec/m, which includes a coccection factor of 2.2 to the dilution ~ calculated by the Pasquill method. This correction factor was shown 3.1-10

r appropriate by onon measurements using SF6 (sulfur hexafluoride) as a gas tracer. The combined gamcole body dose from a semi-infinite cloud is given by: i 3 Dose (Rem) =.* (3.7 x 1010 dps/C1). (1.33 x 10-11 Rem /MeV/m )] Dose (Rem) = 0.21. 0.5 Amax(pc/cc) =. 0.246 x E x 78.25 x 1.16 x 10-4 Amax(pC[CC) Where 3 A = Reactor cty (pCi/ml = Ci/m ) V = Reactor c at 580*F leaked into secondary system (2763 ft3-3 78.25 m ) X/Q = Atmosphes coef fieient at site beundary for a two hour 3 period (%/m ) 5 = Average I cnergies per disintegration (MeV) corrected to operatinpnd pr_ essure. Calculations reqcine E will consist of the following: 1. Quantitativef the specific activity (in units of pCi/ce) of ~ radionuclideries longer than 30 minutes, which =cke up at least 95% ofivity in reactor coolant samples. 2. A determinatrage beta and ga=ma decay energies per disinte-gration for ceasured in (1) above, by utilizing known decay energies and (e.g., Table of Isotopes, Sixth Edition, March 1968). 3. A calculatioaverage beta and gamma energy for each radionuc-lide in propspecific activity, as measured in (1) above. REFERENCES FSAR, Section 14 ..] 3.1 -.

t 3.1.5 chemistry Scecification 3.1.5.1 If the concentration of oxygen in the primary coolant exceeds 0.1 ppm during power operation, corrective action shall be initiated within eight hours to return oxygen levels to 1 0.1 ppm. 3.1.5.2 If the concentration of chloride in the primary coolant exceeds 0.10 ppm during power operation, corrective action shall be initiated within eight hours to return chloride levels to 1 0.10 ppe. 3.1.5.3 If the concentration of fluorides in the primary coolant exceeds 0.10 ppm following modifications or repair to the primary system involving welding, corrective action shall be initiated within eight hours to return fluoride levels to 1 0.10 pps. 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 above are not restored within 24 hours the reactor shall be placed in a hot shutdown conditica within 12 hours there-after. If the normal operational limits are not restored within an additional 24-hour period, the reactor shall be placed in a cold shut-down condition within 24-hours thereafter. 3.1.5.5 If the oxygen cencentration and the chloride or fluoride concentra-tion of the primary coolant system individually exceed 1.0 ppa, the reactor shall be immediately brought to the hot shutdown condition using normal shutdown procedure and action is to be taken immediately to return the system to within normal operation specifications. If normal operating specifications have not been reached in 12 hours, the reactor shall be brought to a cold shutdown condition using normal procedure. Bases By maintaining the chloride, fluoride and oxygen concentration in 7he reactcc coolant within the specifications, the integrity of the reactor coolant system is protected against potential stress corrosion attack. (1,2) The oxygen concentration in the reactor coolant system is normally expected to be below detectable limits since dissolved hydrogen is used when the reactor is critical and a residual of hydrazine is used when the reactor is suberitical to control the oxygen. The requirement that the oxygen concentration not exceed 0.1 ppe during power operation is added assurance that stress corrosion (4) 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, replace the ion exchange resin, increase the hydrogen concentration in the makeup tank, etc.) and further because of the time dependent nature of any adverse effects ariaing from chlorides or oxygen concentrations in excess of the limits, it is unnecessary to shutdown immediately, since the condition can be corrected. 3.1-12

p. a ^ The oxygen and halogen limits specified are at least an order of magnitude below concentraticas which could result in damage to materials found in the "'\\ reactor coolant system even if maintained for an extended period of time. (4) j Thus, the 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 contamiration. If corrective action has not been effective at the end of 24 hours, taen the reactor coolant system will be brought to the hot shutdown condition within 12 hours and corrective l action will continue. If the operational limits are not restored within an j additional 24 hour period, the reactor shall be placed in a cold shutdown condition within 24 hours thereafter. j The maximum limit of 1 ppm for the oxygen and halogen concentration that will l not be exceeded was selected as the hot shutdown limit because these values j have been shown to be safe at 500*F. (3) l REFERENCES (1) FSAR, Section 4.1.2.7 j. (2) FSAR, Section 9.2.2 j (3) Stress Corrosion of Metals, Logan f (4) Corrosion and Wear Handbook, O. J. DePaul, Editor i a 4 J l N n 3.1-13

3.1.6 Leakage Soecification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpu, 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 gpm or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be shutdcwn within 24 hours of detection. 3.1.6.3 If any reactor coolant leakage exists through a non-isolable fault in a RCS strength boundary (such as the reactor vessel, piping, valve body, etc., except the steam generator tubes), 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 cf shutdown and the conditions of shutdown shall be determined by the safety evaluation for each case and justified in writing as soon thereafter as practicable. 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 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 guidelines of 10CFR20. 3.1.6.6 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2 or 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 leak detection systems of different operating principles shall be in operation, with cne of the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for 48 hours provided two other means are available to detect leakage. 3.1.6.3 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 not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1.6.1, 3.1.6.2, 3.1.6.3, 3.1.6.4, 3.1.6.5, or 3.1.6.6 except that such losses when added to leakage shall not exceed 30 gpm. Bases Every reasonable effort 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 belcw 1 gpm in order to prevent a large leak from masking the presence of a smaller leak. Water inventory balances, radiation monitoring equipment, boric acid crystalline deposits, and physical inspections can dis-close reactor coolant leaks. Any leak of radioactive fluid, whether from the 3.1-14

reactor coolant system primary boundary or not can be a serious problem with respect to in-plant radioactivity contamination and cleanup or it could s develop into a still more serious problem; and therefore, first indications of such leakage will be followed up as soon as practicable. Although mome leak -stes en the a-der af CDM ~2 'n *elarable from a dose point of view, especially if they are to closed systems, it must be recog-nized that leaks in the order of drops per minute through any of the walls of the primary system could be indicative of materials 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 be 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 be per-formed by the Operating Staff and will be documented in writing and approved by the Superintendent. Under these conditions, an allowable reactor coolanc system leakage rate of 10 gpm has been established. This explained leakage rate of 10 gpm is also well within the capacity of one high pressure injection pump and makeup would be available even under the loss of off-site power condition. If leakage is to the reactor building it may be identified by one or more of the following methods: a. The reactor building air particulate monitor is sensitive to low leak rates. The rates of reactor coolant leakage to which the instrument is sensitive are.10 gpm to greater than 30 gpm, assuming corrosion product activity and no fuel cladding lealage. Under these conditions, an increase in coolant leakage of 1 gpm is detectable within 10 minutes after it occurs, b. The iodine monitor, gaseous mo itor and area monitor are not as sensitive to corrosion product activity. 1) It is calculated that the iodine monitor is sensitive to an 8 gpm leak and the gaseous monitor is sen-sitive to a 230 gpm leak based on the presence of tramp uranium (no fission products from tramp uranium are assumed to be present). However, any fission products in the coolant will make these monitors more sensitive to coolant leakage. 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 from any of the systems located inside the reactor building such as reactor coolant system, low pressure service water system, component cooling system 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 on the level indicates 1/2 inch of sump height. This indicator is capable of detecting changes on the order of 7.5 gallons of leakage into the sump. A 1 gpm leak would therefore be detectable within less than 10 minutes. 3.1-15

Total reactor coclant system leakage rate is periodically determined by temperature, pressuricer d. comparing indications of reactor power, coolant All vater level and letdown storage tank level over a time interval. Since the pressurizer level is cain-of these indications are recorded. tained essentially constant by the pressurizer level controller, any from the letdown storage tank ecolant leakage is replaced by coolantThe letdown storage tank capacity resulting in a tank level decrease. level recorder .lons per inch of height and each graduation on the is 31 g6 This inventory monitoring method is represents 1 inch of tank height. A 1 gpn leak capaole of detecting changes on the order of 31 gallons. would therefore be detectable within approximately one half hour. in addition to direct observation, the means of As described above, principles, detecting reactor coolant leakage are based on 2 different inventory measurements. i.e., activity, sump level and reactor constant Two systems of dif f erent 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 A 30 gpm loss of water in conjunction with a complete loss of station power. station power and subsecuent cooldown of the reactor coolant system by the turbine bypass system (set at 1,040 psia) and steam driven emergency feedwater pump <truld r2 quire more than 60 minutes to empty the pressurizer from the com-This will be ample time to biced effect of system leakage and contraction. restcra electrical power to the station and makeup flow to the reactor coolant system. REFERENCES FSAR Section 11.1.2.4.1 3.1-16

of Reactivity Mcderator Tencerature Coefficient 3.1.7 Sodeification full ptwer shall not exceed The maximum moderator temperature coefficient at +0.9 x 10-4 ak/k/*F. Bases analysis described in the FSAR (1) was done for a range of 4 ak/k/*F. The accident moderator temperature coefficient values including +0.9 x 10-The maximum positive value of the moderator coef ficient used in the calcu-(2) for the loss-of-coolant accident was +0.9 lation of clad temperature The threshold value of the positive moderator coefficient x 10 " ik/k/*F. allows azimuthal xenon oscillations is greater than +0.9 x 10-" that ik/k/*F. full power zero power value is corrected to obtain the hot 'ihen the hot value, the following corrections will be applied. A. Uncertainty in isothermal measurement The reasured moderator tcmperature coefficient will contain uncertainty on the account of the following: 10.2*F in the iT of the base and perturbed conditions. 1. of to.1 x 10-4 2. Uncertainty in the reactivity measurement ak/k. Proper corrections will be added for the above conditions to conservative coderator coefficient. result in 1 3. Doppler coefficient at hot zero power During the isothernal nederator coefficient measurement at hot zero power, the fuel tenperature will increase by the same amount The measured temperature coefficient must be as the moderator. increased by 0.16 x 10-"(Ak/k)/*F to obtain a pure moderator temperature coefficient. C. Moderator t ?.perature change zero power measurement must be reduced by.09 x 10-4 The hot This corrects for the difference in water temperature (ak/k)/*F. zero power (532*F) and 13.. pcwer (580*F) and for the increased atfuel te=perature effects at 15% power. Above this pcwer, the However, the co-average coderator tenperature recains 580*F. also be adjusted for the interaction of an efficient, an, must average moderator temoerature with increased fuel temperatures. This correction is.001 x 10-4 ca /1% power. It adjusts the 15% n to the moderator coefficient at any power level above 15 is adjusted by power an pcwer. For example to correct to 100% power, 23 (.001 x 10-*) (35%), which is .085 x 10-4 23 3.1-17 e

D. Dissolved bocon concentration i This correction is for any difference in bocon concentration, if required, between zero and full power. Since the moderator coefficient is more positive for greater dissolved boren concen-trations, the sign of the correction depends on whether boron is added or removed. The correction is 0.16 x 10-6 aa /aPPM. (The m magnitude of the correction varies slightly with boron concen-tration; the value presented above, however, is valid for a range in boron concentrations from 1000 to 1400 ppm.) E. Control rod insertion This correction is for the difference in control rod worth (% ak/k) in the core between zero and full power. The correction is 0.17 x 10-4 aa /%Ak/k, where the sign for rod worth change is m negative for rod insertion, because the moderator coefficient is more negative for a larger rod worth in the cora. F. Isothermal to distributed temperature The correction for spatially distributed moderator temperature has been found to be 1 css than or equal to zero. Therefore, zero is a conservative correction value for distributed effects. G. Azimuthal xenon stability Before commercial operation a test will be performed to verify that divergent azimuthal xenon oscillstions do not occur. REFERENCES (1) FSAR, Section 14 (2) FSAR, Section 3 (3) FSAR, Section 14.2.2.3.4 1 i i i 1 I 3.1-18 L

3.1.8 single Loon Restrictions Soecification The following special limitations are placed on single loop operation in addition to the lititations set forth in Specification-2.3. 3.1.8.1 Single loop operation is authorized for test purposes only. At least 23 incore detectors meeting the requirements of Technical 3.1.8.2 Specification 3.5.4.1 and 3.5.4.2 shall be available throughout this test to check gross core power distribution. 3.1.8.3 The pump monitor trip set point shall be set at no greater than 50% of rated power. 3.1.8.4 The outlet reactor coolant temperature trip set point shall be at no greater than 610*F. set At 15% of rated power and every 10% of rated power above 15%, 3.1.8.5 measurements shall be taken of each operable incore neut-on detector and each operable incore thermocouple, reactor coolant icop flow rates and vessel inlet and outlet temperature, and evaluation of this data determined to be acceptable before pro-ceeding to higher power levels. DOL shall be notified of the scheduled date of single loop testing. 3.1.8.6 Upon complation of test, results shall be reported to AEC/ DOL. Subsequent single loop operation shall be contingent upon written appro/al by DOL. Bases the 1/6 scale model The purpose of single loop testing is to (1) supplement test information, (2) verify predicted flow through the idle loop, (3) verify that 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 temperature 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 reactor f temperature trip set point to 610*F to perform this confirmatory coolant outlet testing assures operation well within the core protective safety limits shown in Figure 2.1-3, curve 2. Incore thermocouples will be installed and data will be taken to check outlet core temperature profiles. This data will be used in evaluating test results. 3.1-19 m

Low Power Physics Testine Restrictions 3.1.9 Specification The following special limitations are placed on low power physics testing. 3.1.9.1 Reactor Protective System Requirements Below 1720 psig Shutdown Bypass trip setting limits shall apply in a. accordance with Table 2.3-1. less than Above 1800 psig nuclear overpower trip shall be set at b. Other settings shall be in accordance with Table 2.3-1. 5.0%. at all Startup rate red withdrawal hold shall be in effect 3.1.9.2 This applies to both the source and intermediate ranges. times. Bases 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. 3.1-20 w-

3.1.10 Control Rod Goeration Specification Allowable co=binations of pressure and te perature for centrol rod 3.1.10.1 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 cc/ liter. If either the limits of 3.1.10.1 or 3.1.10.2 are exceeded, the 3.1.10.3 center control rod drive mechanism shall be checked for accumulation of undissolved gases. Bases The limiting pressure versus temperature curve for dissolved gases is determined by the equilibrium pressure versus temperature curve f or the dissolved gas con-centration of 100 std. cc/ liter of water. This equilibrium total pressure is the sum of the partial pressure of the dissolved gases plus the partial pressure of water at a given temperature. temperature and pressura as specified above, By maintaining the reactor coolant any dissolved' gases in the reactor coolant system are maintained in solution. Although the dissolved gas concentration is expected to be approximately 20-40 cc/ liter of water, the dissolved gas concentration is conservatively std. 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 limit-ing pressure versus temperature curve, the center CRDM should be checked for accumulation of undissolved gases. } j 3.1-21

2000 J 1800 4 1600 $ 1400 = 3 0 1200 E 5; J 1009 2 O3 800 7 5 0 a: s 3 600 ) 0 2 400 e a 200 f 0 0 100 200 300 400 500 600 700 Inaicaten Reactor Coolant System Temperature,*F LLMITING PRESSURE VS TEMPERATURE CURVE FOR 100 STO CC/ LITER H O 2 l Ot J t ro OCONEE NUCLEAR STATION Figure 3.1.10 - 1 Page 3.1 22 I

=. ~. i i i 3.2 HIGH PRESSURE INJECTION ASD CHEMICAL ADDITICN SYSTEMS Aeplicab111tv 8 i Applies to the high pressure injection and the chemical cdditi:n systems. I d Obiective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition. l l j Specification The reactor shall not be critical unless the following conditions are met: i 3.2.1 Two high pressure injection pu=ps per unit are operable except as specified in 3.3. l 3.2.2 One source per unit of concentrated soluble boric acid in addition to the borated water storage tank is available and operable. This can be either: b a. The boric acid mix tank containing at least the equivalent of 450 ft3 of 10,600 ppm boron as boric acid solution at a temperature of at least 10*F above the crystallization temperature. System j piping and valves necessary to establish a flow path from the tank to the high pressure injection system shall also be operable and shall have at least the same temperature requirement as the i boric acid mix tank. One associated boric acid pump shall be l operable. If the daily average air temperatare in the vicinity of i this tank and associated flow path piping is less than 85'F, at I least one channel of heat tracing shall be in operation for this tat.k and piping. a I i b. The concentrated boric acid storage tank containing at least the 3 equivalent of 550 ft of 8700 ppm boron as boric acid solution l with a temperature of et least 10*F above the crystalli:ation I temperature. System piping and valvea necessary to establish a flev 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. 11 the daily average 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 i The high pressure injection system and chemical addition system provide j . control of: the reactor coolsnt system boron concentration.(1) This is j normally accomplished by using any of the three high pressure injection pumps in series with a boric acid pump associated with either the boric acid mix tank or the concentrated boric acid storage tank. An alternate method of 3,. 2-1

1 boration will be use of the high. pressure injection pumps taking suction ) directly from the borated water storage tank.(2) The quantity of boric acid in storage from any of the 3 above mentioned sources is sufficient to borate the reactor coolant system to a 1% sub-critical margin in the cold condition at the end of core life. The maximum required is the equivalent of 396 ft3 of 10,600 ppm boron as boric acid solution. A minimum of 450 ft3 of 10,600 ppm boron as boric acid solution in the boric acid mix tank, a minimum of 550 ft3 of 8,700 ppm boron as boric acid solution in the' concentrated boric acid storage tank or a minimum of 350,000 gallons of 1800 ppm boron as boric acid solution in the borated water storage tank (3) will satisfy the requirements. The specification assures that at least two 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. The quickest method allows for the necessary boron addition in less than one hour. The slowest method (using one 10 gym pump taking suction from the boric acid storage tank) would require approximately 3 hours to inject enough boron to keep the reactor 1% suberitical with xenon in the core. As xenon decays out, more boron would have to be added. Therefore, in order to account for xenon decay, the 10 gpm pump would pump for something less than 5 hours. At this time, the reactor coolant system would be at a temperature of approximately 175'F and the core would be more than 1% suberitical. J The concentration of boron in the boric acid mix tank and concentrated boric l acid storage tank may be higher than the concentration which would crystallize at ambient conditions. For this reason and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept 10*F above the crystallization temperature for the concentration present. Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility. 4 The boric acid mix tank concentration of 10,600 ppm boren corresponds to a precipitation temperature of 80*F, and the concentrated boric acid storage tank concentration of 8700 ppm corresponds to a precipitation temperature of 68*F. It is expected that the surface temperatures of these tanks and associated piping will be 10*F above the precipitation temperatures. 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 will be made up to =aintain this 10*F =argin. REFERENCES (1) FSAR, Section 9.1; 9.2 (2) FSAR Figure 6.2 2 (3) Technical Specification 3.3 ): 3.2 ~

4 3.3 DiERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR BUILDING SPRAY AND PENETRATION ROOM VENTILATION SYSTC45 Applicability Applies to the emergency core cooling, reactor building cooling, reactor building spray and reactor building penetration room ventilation systems. Obiective To define the conditions necessary to assure i= mediate availability of the energency core cooling, reactor building cooling, reactor building spray and reactor building penetration room ventilation systems. i Specification 4 3.3.1 The reactor shall not be made critical, unless the following j 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 concen-I tration of 1,800 ppm boron at a temperature not less than 40*F. The manual valve, LP-23,.cni the discharge line from the borated water storage tank shall be locked open. (2) Two out of three high pressure injection pumps shall be operable. a' (3) Two Engineered Safety feature low pressure injection pumps shall be operable. (4) Two low pressure injection coolers shall be operable. I (5) Two BWST level instrument channels shall be operable. (6) The two reactor building emergency sump isolation valves shall be either manually or remote-manually i operable. j b. Core Flooding System (1) The two core flooding tanks shall each contain a min-3 of borated water at 600 i 25 psig. inum of 104Qt 30 ft i (2) Core Flooding tank baron concentration shall not be less than 1,800 ppm boron. (3) The electrically operated discharge valves from the core flood tanks shall be open and breakers locked open'and tagged. i 3.3-1 ? l L 1

4 i 'S (4) One pressure instrument channel and one level instru-ment channel per tank shall be operable. c. Reactor Building Spray System and Reactor Building Cooling System The following subsystems shall be operable: (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 pumps shall be operable. (2) The valve in the LPSW discharge from the reactor building cooler (LPSW 108) shall be locked open. 1 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 PR-18) shall be locked open. f. Engineered Safety feature valves and interlocks associated with each of the above systems shall be operable. 3.3.2 Except as noted in 3.3.3 below, maintenance shall be allowed during power operation on any component (s) in the high pressure injection, core flooding, low pr.ssure injection, low pressure service water, reactor building spray, or penetration room ventilation systems which will not remove more than one train of each system from service. Components shall not be removed from service so that the affected system train is inoperable for more than 24 consecutive hours. If the system is not restored to meet the requirements of Specification 3.3.1 within 24 hours, the reactor shall be placed in a hot shutdown condition within 12 hours. If the requirements of Specification 3.3.1 are not met within an additional 48 hours, the reactor shall be placed in a cold shutdown condition within 24 hours. 3.3.3 Exceptions to 3.3.2 shall be as follows: a. Both core flooding tanks shall be operational at all times. b. Both motor operated valves associated with the core flooding tanks shall be fully open at all times. \\ 3.3-2 3

i i l i l_ c. One reactor building cooling fan and associated cooling unit shall be per=itted to be out-of-service for seven days provided no more than one reactor building spray pu=p and associated spray nozzle header is out of service at the sa=e time per Specification j 3.3.2. j 3.3.4 Prior to initiating maintenance on any of the components, the duplicate (redundant) component shall be tested to assure operability. Bases The requirements of Srecification 3.3.1 assure that, before the reactor can be made critical, adequnte engineered safety features are operable. Two high pressure injection ptmps and two low pressure injection pumps are specified. However, only one of each is necessary to supply emergency coolant to the j reactor in the event of a loss-of-coolant accident. Both core flooding tarks are required as a single core flood tank has insufficient inventory to reflood i the core. (1) 1 The borated water storage tanks are used for two purposes: i A. As a supply of borated water for accident conditions. l 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 emergency core cooling s and reactor building sprav '.s the event of a loss-of-core cooling accident. j

• ihis amount fulfills rzquirements for emergency core cooling. The borated water storage tank capacity of 388,000 gallons is based u refueling volu=e l

requirements. Heaters maintain the borated water supply at a temperature to j prevent freezing. The boron concentration is set at the amount of baron j required to maintain the core 1 percent suberitical at 70*F without any control rods in the core. This concentration is 1,338 ppm boron while the j mirauum value specified in the tanks is 1,800 ppm boron. J The post accident reactor building cooling may be accomplished by three cool-ing units, by two spray units or by a cc=bination of two cooling units and 3 one spray unit. The specified requirements assure that the required post accident components are available.(3) The spray system utilizes co= mon suction lines with the low pressure injection system. If a single train of equipment is removed from either system, the .other train must be assured to be operable in each system. When the reactor is critical, maintenance is alleved per Specification 3.3.2 { and 3.3.3 provided requirements in Specification 3.3.4 are met which assure operability of the duplicate components. Ope. ability of the specified com-ponnts shall be based on the results of testi y as required by Technical 4 Specification 4.5. The maintenance period of up to 24 hours is acceptable if 4 i the operability of equipment redundant to that removed from service is demon-i' stra6ed i= mediately prior to removal. The basis of acceptability is a low likelihood of failure within 24 hours following such demonstration. 3.3-3 ~

In the event that the need for emergency core cooling should occur, functioning ~S of one train (one high pressure injection pump, 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 the peak clad temperature to less than 2,300*F and the metal-water reaction to that representing less than 1 percent of the clad. Three low pressure service water pumps serve Oconee Units 1 and 2 and two low pressure service water pumps serve Oconee L' nit 3. There is a manual cross-connection on the supply headers for Units 1, 2 and 3. One law pressure service water pump per unit is required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of-coolant accident. A single train of reactor building penetration room 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. REFERENCES (1) FSAR, Section 14.2.2.3 (2) FSA2, Section 9.5.2 (3) FSAR, Sectiqn 14.2.2.3.5 (4) FSAR, Section 6.4 i 3.3-4 l l 1 l

3.4 STEAM & P0h'ER CONVERSION SYSTEM Aeolicability Applies to the turbine cycle components for removal 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*F unless the following conditions are met: 3.4.1 Capability to remove a decay heat load of 5 percent full reactor power j from at least one of the follcwing means: A hotwell pump, condensate booster pu=p, and a main feedwater pump. a. b. The emergency feedwater pu=p. A hotwell pump and a condensa*.a booster pump. c. 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 avail-able in the upper surge tank, condensate storage tank, and 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 main feedwater pump. The feedwater flow required to remove decay heat corresponding to 5 percent full power with saturated steam in the pressure range from 30 psia (saturation pressure at 250*F) to 1065 psia (lowest setting of steam safety valve) as a function of feedwater temperature is:

• F Flcw. GPM 60 750 90 770 120 790 180 840 One hotwell pump plus one condensate booster punp will supply at least 3000 GEI at _550 psia, and one hotwell pump plus one booster pump _ plus one main 3.4-1 i

i feed pump will supply at least 3000 gpm at 1065 nsia. The emergency feed pump 3 will supply 1080 gpm at 1065 psia. In the event of complete loss of electrical power, feedwater is supplied by a turbine driven emergency feedwater pump which takes suction from the upper 3 ~ surge tanks and hotwell. Decay heat is removed from 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 Keowee through the con-denser for final heat rejection to the Keowee Hydro Plant tailrace. The minimum amount of water in the upper surge tank and condensate storage i tank is the amount needed for 11 hours of operating per unit. This is based on the conservative estimate of normal makeup being 0.5% of throttle 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 i 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 lbs/hr. The capacity of the four turbine. bypass valves is 2,817,000 lbs/hr. REFERENCE USAR, Section 10 t i -l l 3.4-2

3.5 INST 2DICTATICF 3YSTD.'S 3.5.1 Operational Saferv Instrumentation Applicabilitv Applies to unit instrumentation and control systems. Obiective To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety. Soecifications 3.5.1.1 The reactor shall not be in a startup mode or in a critical state unless the requ2renents of Table 3.5.1-1, Columns A and E are met. 3.5.1.2 In the event that the number of protective channels operable falls below the limit given under Table 3.5.1-1, Columns A and B; operation shall be limited as specified in Column C. 3.5.1.3 For on-line testing or in the event of a protective instrunent or channel failure, a key-operated' channel bypass switch associated with each rea::cr pro:2ctive channel =ay be used to leck the 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==y 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 come on scale, the overlap between the intermediate range and the source range instrunentation shall not be less than one decade. If the 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 fciled trip device shall be manually removed within 30 minutes. The condition will be corrected and the remaining trip devices shall be tested within eight hours. If the condition is not corrected and the remaining trip devices tested within the eight hour period, the reactor shall be placed in the hot shutdown condition within an additional four hours. 3.5-1

Bases '} Every reasonabic effort will be made 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: fcur reactor coolant te=perature 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 channels functioning correctly prior to a startup. Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column B (Table 3.5.1-1). This is in agreement with redundancy and single failure criteria of IEEE 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 channels is two out of three. Minimum trip logic on other channels is one out of two. The four reactor protective channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alarm and lights to indicate when that channel is bypassed. There will be one reactor protective system bypass switch key permitted in the control room. That key will be under the administrative control of the Shift 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 four 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 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 trip devices fails in the un-tripped 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 3.5-2 ~

) 1 IABLE 3.5.1-1 l INSTRUMENTS OPERATING CONDITIONS (A) (B) (C) ^ Minimum Minimum Operator Action If Conditions j Operable Degree Of Of Column A and B 1 Functional Unit Channels Redundancy Cannot Be Met i 1. Nuclear Instrumentation 1 0 Bring to hot shutdown within Intermediate Range 12 hours (b) Channels j 2. Nuclear Instrumentation 1 0 Bring to hot shutdown within l l Source Range Channels 12 hours (b)(c) 3. RF3 Manual Pushbutton 1 0 Bring to hot 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 hours Channels j 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 a. High Reactor Coolant 2 1 Bring to hot shutdown within Pressure Instrument 12 hours Channels - b.' Low. Reactor Coolant 2 1 Bring to hot shutdown within Pressure Channels 12 hours 9. RPS Power-Number of_ Pumps 2 1 Bring to hot shutdown within Instrument Channels 12 hours-10. RPS High Reactor Building 2 1 Bring to hot shutdown within Pressure Channels-12 hours I 4 3.5-3 o r-e <y. y- ,_-y.._-,-w._-- -.._,m. ..v-.. +,.,.,,. _ .,,,w. ,r-. .w--

4 TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (Cont'd) .] (A) Minimum (B) (C) Operable Minimum Operator Action if Conditions Analog Degree Of Of Column A and B Functional Unit Channels Redundancy Cannot Be Met 11. ESF High Pressure Injection System a. Reactor Coolant 2 1 Bring to hot shutdown within Pressure Instru-12 hours (e) ment 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 hot shutdown within 12 hours (e) 12. ESF Low Pressure In-jection System a. Reactor Coolant 2 1 Bring to hot shutdown within Pressure Instrument 12 hours (e) Channels b. Reactor Building 2 1 Bring to hot shutdown within l 4 PSIG Instrument 12 hours (e) Channels c. Manual Pushbutton 2 1 Bring to hot shutdown within 12 hours (e) 13. ESF Reactor Building Isolation & Reactor Building Cooling System a. Reactor Building 2 1 Bring to hot shutdown within 4 PSIG Instrument 12 hours (e) Channel b. Manual Pushbutton 2 1 Bring to hot shutdown within 12 hours (e) 14. ESF Reactor Building Spray System a. Reactor Building 2 1 Bring to hot shutdown within High Pressure 12 hours (e) Instrument Chernel 3.5-4 l'

e O ( TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (Cont'd) (A) Minimum (B) (C) Operable Minimum Operator Action If Conditions Analog Degree Of Of Column A and B Functional Unit Channels Redundancv Cannot Be Met b. Manual Pushbutton 2 1 Bring to hot shutdown within 12 hours (e) 15. Turbine Stop Valves 2 1 Bring to hot shutdown within Closure 12 hours (f) (a) For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours. (b) When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required. (c) When 1 of 2 intermediate range instrument channels is greater than 10-10 amps, hot shutdown is not required. (d) Single loop operation at power (af ter testing and approval by the AEC/ DOL) is not permitted unless the operating channels are the two receiving Reactor Coolant Temperature from operating loop. (e) If minimum conditions are not met within 48 hours after hot shutdown, the unit shall be in the cold shutdown condition within 24 hours. (f) One operable channel with zero minimum degree of redundancy is allowed for 24 hours before going to the hot shutdown condition. 3.5-5

l 1 3.5.2 control Rod Group and Power Distribution Limits Applicability This specification applies to power distribution and operation of control rods during power operation. Obiective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core suberiticality after a reactor trip. Soecification 3.5.2.1 The available shutdown =argin shall be not less than 1% ak/k with the highest worth control rod fully withdrawn. 3.5.2.2 Operation with inoperable rods: Operation with more than one inoperable rod as defined in a. Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted. b. If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated i= mediately to verify the existance of 1% ok/k hot shutdown margin. Beration may be initiated either to the worth of the inoperable red or until the regulating 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% ok/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is established. d. Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised within 24 hours and exercised weekly until the rod problem is solved. 3.5.2.3 The worth of a single inserted control rod shall not exceed 0.5% 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 thermal limits as defined in Specification 2.1. 3.5-6

~5 If the quadrant power tilt exceeds 10% except for physics tests, a. power shall be limited to 80% of the thermal power allowable for the reactor coolant pump combination. 4 b. If the quadrant power tilt exceeds 20% except for physics tests, power shall be limited to 60% of the thermal power allowable for the reactor coolant pump combination. If any control rod is declared inoperable per 4.7.1.2, power c. shall be reduced to 60% of the thermal power allowable for the reactor coolant pump combination. d. If the inoperable rod in paragraph (c) above is in Groups 5, 6, 4 7, or 8, the other rods in the group shall be trimmed to the same position. Normal operation may then continue provided that the rod that was declared inoperable is maintained within allowable bank average position limits. 3.5.2.5 Control Rod Positions: Control rod safety groups must be fully withdrawn before the a. reactor is critical or approaching criticality ~except for physics tests. Individual control rods may be exercised at power as re-quired by Tabie 4.1-2. b. Group 5 shall not be inserted during operation at the rated ~. except when exercising rods or physics tests. ' ) thermal power allowable for the reactor coolant pump combination c. Operating rod group overlap shall not exceed 30% between two sequential groups, except for physics tests. 3.5.2.6 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent. Bases The 30% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged in groups or banks defined as follows: 1 Croup Function 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Xenon transient overrice 8 APSR (Axial Power Shaping Bank) Control rod groups are withdrawn in sequence beginning with Group 1. Groups 3.5-7

5, 6 and 7 are overlapped 25%. The normal position at power is for Groups 6 and 7 to be partially inserted. The minimu= available rod worth provides for achieving hot shutdown by reactor trip at any time assuming the highest worth 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 analysis of the hypothetical red ejection accident.(2) A single inserted control rod worth of 1.0% ak/k at beginning of life, hot, zero power would result in the same transient peak thermal power and therefore the same environmental consequences as a 0.5% ak/k ejected rod worth at rated power. The quadrant power tilt li=1ts set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the definition of quadrant powar tilt given in Section 1.6. REFERENCES (1) FSAR, Section 3.2.2.1.2 (2) FSAR, Section 14.2.2.2 3.5-6

1 1 3.5.3 Engineered Safety Features Protective System Actuation Setpoints Aeolicability This specification applies t the engineered safety features protective system actuation setpoints. l Obiective To provide for automatic initiation of the engineered safety features protective system in the event of a breach of RCS integrity. I Soecification l The.e,ngineered safety features protective actuation setpoints and permissible bypasses shall be as follows: Functional Unit Action Setpoint 4 High Reactor Building Reactor Building Spray <30 psig i Pressure High-Pressure Injection <4 psig-Low-Pressure Injection g4 psig l Start Reactor Building Cooling & Reactor Building i Isolation g4 psig ^ I Penetration Room Ventilation <4 psig Low Reactor Coolant High Pressure Injection (l) g1500 psig ) System Pressure ) Low Pressure Injection (2) g500 psig

1 l

(1) May be bypassed below 1750 psig and is auto =atically reinstated above 1750 psig. (2) Sky be bypassed below 900 psig and is automatically reinstated 1 above 900 psig. Bases High Reactor Building Pressure l l The basis for the 30 psig and 4 psig setpoints for the high pressure signal is to establish a setting which would be reached i= mediately in the event of a DBA, cover the entire spectrum of break si=es and yet be far enough above normal operation maximum internal pressure to prevent spurious initiation. Lcw Reactor Coolant Svstem Pressure The basis for the 1500 psig icw reactor ecolant pressure setpoint for high presr.re injection initiation and 500 psig for low pressure injection is to 3.5-9

ided for establish a value which is high enough such that protection is prov l operating the entire spectrum of break sizes and is far enough below nor=a pressure to prevent spurious initiation.(1) REFERENCES (1) FSAR, Su tion 14.2.2.3. t i 3.5-10 ]

3.5.4 Incore Instrumentation Aeolicability Applies to the operability of the incore instrumentation system. Obiective To specify the. functional and operational requirements of the incere instru-mentation system. Specification Above 80% of operating power determined by the reactor coolant pump ce=b tion, reference tabic 2.3.1, at least in the periodic operable to check gross core power distribution and to assist i calibration of the out-of-core detectors in regard to the core imba lir.it s. basic arrangements. 3.5.4.1 Axial Imbalance Three detecters in each of 3 strings shall lie in the same axial a. plane with 1 plane in each axial core half. The axial planes in each core half shall be sym=etrical about the b. core mid-plane, The detector strings shall not have radial symmetry. c. 3.5.4.2 Rcdia Tilt Each set of Two sets of 4 detectors shall lie in each core half. The two sets in the same core a. 4 shall lie in the same axial plane. half may lie in the same axial plane. Detectors in the same plane shall have quarter core radial symmetry. b. Bases _ A system of 52 incore flux detector assemblies with 7 detectors per assembly The system includes data display and record functions has been provided. and is used primarily for out-of-core nuclear instrumentation calibration and fer core power distribution verification. The out-of-core nuclear instrumentation calibration includes: A. Calibration of the split detectors at initial reactor startup, during the power escalation program, and periodically thereafter. 1. A comparison check with the incore instrumentation in the event one of-the four out-of-core detector assemblies gives abnorcal 2 ~. . readings during operation. ) 3.5-11 i

3. Confirmation that the out-of-core axial power splits are ~N 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 power distribution is consistent with calculations. 3. Indication of power distribution in the event that abnormal situations occur during reactor operation. C. The safety of unit operation at or below 80% of operating power (l) for the reactor coolant pump combinations without the core imbalance trip system has been determined by extensive 3-D calculations. This will be verified during the physics startup testing program. D. The minimum requirement for 23 individual incore detectors is based on the following: 1. An adequate axial i= balance indication can be obtained with 9 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% (calculated) of the real core im-balance. The three detector strings are the center one, one from the inner ring of symmetrical strings and one from the outer ring of symmetrical strings. Both steady state and design transient data from the Oconee 1 maneuvering analysis were used for this comparison. 2. Figure 3.5.4-2 shows a detection scheme which will indicate the radial power distribution with 16 individual detectors. The readings from 2 detectors in a radial quandrant at either plane can be compared with readings from the other quadrants 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 j 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 least 23 specified incore detectors will be operable to check power distribution above 80% power determined by reactor coolant pump combina-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 3.5-12

M Lack radial symmetry ~ p ~ ' [O / pN\\ / N .J> ( \\,- / / Axial Plane s 5 N / ~ r N-5 7- / N [ h Top Axial Core Half 1 > w 5 E O \\ \\ j / ~ 0 N / N / 2 ~ Bottom Axial Core Halt. eM f ^ [# N [O 4 / \\ \\ \\ / N N / i ' ~~ i 533 ~~_ 's i / 's !/ N-1, i INCORE INSTRUWENiaT105 SPECIFtCATION a(tat IMBALANCE INDICail0N 'L h piinaa OCONEE NUCLEAR STATION Figure 3.5.4 1 3.5-13

I .~ y l Radial I - I I I /xl% h Radial Symmetry [ f M / in this plane \\ - - + / 5 N -l-neu o / w% x y b / / \\ \\ ( \\ / / m / 2 K ___~ ~- / y '\\ 5 - -s / g Radial Symmetry \\ in this plane g y -[-_ -+ # -~ ~ / N / N INCCRE INSTRUMENTATION SPECIFICATION l RADIAL FLUX TILT INDICATION \\ hQ>; OCONEE NUCLEAR STATION utraata -~ Figure 15.4 - 2 l 3.5-14

M 7EI /b, # W 'N A %*My / \\ g f N' M = E ~7 / .I3 -4m ( [ 4 h 7 \\ NB / / 2 / / ~, ~ E s 5 E VT Y N 4 y 3 I O \\ HS / 4 N 1 xec 1+r 23 ,-~s / / \\ / \\ ,st]:E !ssrauvENTATics SP~ .0:T!as 'ib\\ weau OCONEE NUCLEAR STATION

W Figure 3.5.4 3

3.5-13

3.6 REACTOR BUILDING Acolicability Applies to the containment whan thc r2_; tor is suberitical by less than 1% ak/k. Obiective To assure containment integrity during startup and operation. Snecification 3.6.1 Containment integrity shall be maintained whenever all three (3) of the following conditions exist: a. Reactor coolant pressure is 300 psig or greater. b. Reactor coolant temperature is 200'F or greater. c. Nuclear fuel is in the core. 3.6.2 Containment integrity shall be maintained when the reactor coolant system is open to the containment atmosphere and the requirements for a refueling shutdown are not met. 3.6.3 The containment integrity shall be intact whenever positive reactivity insertions which would result in the reactor being sub-criticsl by less than 1% ak/k are made by control rod motion or boron dilution. 3.6.4 The reactor building internal pressure shall not exceed 1.5 psig or five inches of Hg if the reactor is critical. 3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containnent 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 Recator Coolant System ruptures. The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in.iny occurrence. The reactor building is designed for an internal prersure of 59 psig and an external pressure 3.0 psi greater than the interna.'. pressure. The design external pressure of 3.0 psi corresponds to a margin of 0.5 ps1 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 ecoled to an internal temperature of 80*F with a concurrent rise in barometric pressure to 31.0 inches of Hg. The weather conditiens assumed here are conservative since an evaluation of 3.6-1

~ National Weather Service records for this area indicates that i highest is 30.85 inches of Hg. When containment integrity is established, the limits of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur. REFERENCES _ FSAR, Section 5 3.6-2

3.7 ~ AUXILIARY ELECTRICA< SYSTEMS Applicability Applies to the availability of off-site and on-site electrical power for station operation and for operation of station auxiliaries. Obiec:1ve To dafine those' conditions of electrical power availability necessary to provide for safe reactor operction and to provide for continuing avail-abili y cf engineered safety features systems in an unrestricted manner and ti prescribe safety evaluation and reporting requirements to be followed in the event that the auxiliary electric power systems become degraded. Speci'. cation 3.7.1 Under nornal 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 transmissica lines, on separate towers, shall be in servica. One of these two lines may be out for test or maintenance for pericds up to 24 hours once the reactor has been brought critical. b. Startup Transformers CTl or CT2'shall be operable and available to the Unit 1 4160V Main Feeder Bus No. 1 and No. 2 An operable Keowee Hydro unit shall be available to supply c. power through the Underground Feeder Bus, Transformer [ No. CT4, and the Station 4160 V Standby Buses Nos. 1 or 2 feeding the associated 4160 Main Feeder Bus. The other l Keowee Hydro unit shall be available to supply power through Start-up Transformer CTl or CT2 to Unit 1 4160V Main Feeder Buses Nos. 1 and 2, except that, one Keowee Hydro Unit may be inoperable for periods not exceeding 24 hours for test or maintenance once the reactor has been brought critical. d. Reactor Coolant Pump Buses 1TA or 1TS shall be energized. e. The two 4160 volt main feeder buses shall be energized except one may be de-energized for test or maintenance for periods t.at exceeding 24 hours after the reactor has been brought critical. f. The three 4160 V Engineered Safety Features Switchgear Buses 1TC, 1TD, and ITT shall be energized except one may be de-energized for test or maintenance for periods not exceeding 24 hours after che reactor has been brought critical. 3.7-1

The three.600 V load centers IX8, 1X9, and 1X10 plus the s, l g. three 600V-208Y Engineered Safety Features MCC Buses IXS1, 1XS2, and 1XS3 shall be energized, except that one string may be de-energized for test or maintenance for periods of not exceeding 24 hours after the reactor has been brought critical. i The Unit 1 125 VDC instrumentation and control batteries I h. with its respective chargers, buses, diode monitors, and diodes supplying vital instrumentation and the Unit 2125 VDC instrumentation and control batteries with its chargers and buses shall be operable except that one complete single string or any portion thereof of redundant chargers, batteries, buses and all associated isolating transfer diodes may be de-energized for test or maintenance for periods of not exceeding 24 hours. The 125 VDC switching station batteries with their respective i. chargers, buses, and isolating diodes shall be operable single strings of redundant buses, chargers, batteries, except and all of the related diode assemblies may be de-energized for test or maintenance for periods of not exceeding 24 hours. The Keowee batteries with their respective chargers, buses, 1 j. and isolating dicdes shall be operable except that redundant components may be de-energized for test or maintenance for periods not exceeding 24 hours. The level of the Keowee Reservoir shall be at least 775 feet k. above sea level. In the event that all conditions in Specification 3.7.1 are met except 3.7.2 that one of the two Xeowee Hydro units become unavailable for longer than the test or maintenance period of 24 hours, the reactor shall be permitted to remain critical or restarted provided the following restrictions are observed: The remaining hydro unit will be connected to the underground a. feeder and the Lee Station gas turbine will be energized to i the 4160 standby buses through the 100 KV transmission circuit which is completely separated from the system grid and non-safety-related loads. This shall be in effect prior to re-start or within 30 minutes if at power when a hydro unit is lost. b. The reactor coolant T shall be above 500*F. avg Within 24 hours after loss of a hydro unit, this fact shall be c. reported to the Directorate, Regulatory Operations Region II, to be followed by a written report of the circumstances of j the outage if it is expected to exceed ~24 hours and the es-timated time to eturn the hydro unit to operating condition. s 3.7-2 V

4 -3.7.3 In the event that all conditions in Specification 3.7.1 are met except that tha underground feeder circuit to the standby buses is lost, the reactor shall be permitted to remain critical or . restarted provided the following restrictions are observed: a. The Lee gas station shall be energized to the 4160 standby buses through the 100KV transmission circuit which is com-l pletely separate from the system grid and non-safety-related loads. b. The reactor coolant T shall be above 500*F. avg c. Within 24 hours after the loss of the underground feeder, this fact shall be reported to the Directorate, Regulatory Operations, Region II, to be followed by a written report of the circumstances of the outage if it is expected to exceed l 24 hours and the estimated time to return the underground j feeder circuit to operatica condition. 1 3.7.4 In the event that all conditi s of Specification 3.7.1 are =et i except that all'230 kV transe ssion lines are lost, the reactor shall be permitted to remain critical or restarted provided the ] following restrictions are observed: f a. A Lee 3tation combustion turbine shall be operating and i 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. It shall be avail- ~ able prior to a restart or within 30 minutes if this degradacion occurs during operation. 1 I b. Within 24 hours of loss of all 230 kV transmission lines, this-1egradation shall be reported to the Directorate, Regulatory Operations, Region II, in accordance with Specificatien 6.6. If the outage is expected to continue beyond 24 hours, the j circumstances of the outage and the estimated time to aurn at least two 230 kV circuits to service shall be reported f c. The reactor coolant T shall be above 500*F. avg 3.7.5 Any degradation beyond Specification 3.7.2 or 3.7.3 above shall be 4 l reported to the Directorate, Regulatory Operation, Region II, within } 24 hours. A safety evaluation shall be performed by Duke Power Company for the specific situation involved which justifies the i safest course of action to be taken. The results of this evaluatica together with plans for expediting the return to the unrestricted operating conditions of Specification 3.7.1 above shall be sub-mitted in a written report to the Directorate of Licensing with a i j copy to the Directorate of Regulatory Operations, Regica II, within 5 days. t h l 3.7-3

Bases i The auxiliary electrical power systems are so arranged that no single con-tingency can inactivate enough engineered safety features to jeopardize planc satecy. Giese e,s;;m;..r 2 :.aign:1 c net the following criteria: " Alternate power systems shall be provided and designed with I 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 shcIl each, independently, provide this capacity assuming a failure of a single active component in each power system." The auxiliary power system meets the above criteria and the intent of pro-posed AEC Criterion 17 (published February 20, 1971) with one exception. That exception is the Unit 1 startup transformer (CTl) whose loss would cause loss of all offsite power. This is acceptable because of the high i reliability of such transformers and the relatively short period of time i that will elapse before a redundant transformer, CT2, the Unit 2 startup transformer, is permanently installed as a redundant means of obtaining l Unit 1 4160 ESF bus power from the 230 kV switchyard. The adequaciee of the ac and de systems are discussed below as are-the bases for permitting degraded conditions for ac power. Capacity of AC Systems The auxiliaries activated by ESF signals (4.8 MVa) plus other safety related auxiliaries activated by the operator following a loss-of-coolant The accident (4.9 MVa) require a total ac power capacity of 9.7 MVa. minimum ac power capacity available from the onsite power systems (Keowee Hydro Units) is 12 MVa (limited by transformer CT4) if furnished by the underground circuit or 45 MVa (limited by CTl or CT2) if furnished through the 230 KV switchyard. The minimum capacity availoble from the multiple 230 KV offsite transmission lines is 45 MVa (limited by CT1 or CT2). Capacity available from the backup 100 KV offsite transmission line (Lee Station Gas Turbine Generator) is 12 HVa (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 Normally, ac power is rectified and supplies the de system buses as well 1 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-least I hour. One pendently supply.their required emergency loads tor at hour is considered to be conservative since there are redundant sources of The loss of all ac power providing energy to these de auxiliary systems. ac power to any de system is expected to occur very infrequently, and for i i 3.7-4

very short periods of time. The following tabulation demonstrates the margin of installed battery charger rating and battery capacity when ccm-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 systems installed at Oconee: A. 125 Vdc Instrumentation an-Control Power System a. 600 amps ea.- Charger ICA, 1CB or ICS Rating Battery ICA and 1CB Co=bined Capacity b. 698 ampere-hours Actual active loads on both 125 Vdc a. First min. - 1371 amps I & C buses IDCA and 1DCB next 59 min. -568.5 amps b. 581.9 ampere-hours during 1st hour of LOCA B. 125 Vdc Switching Station Power System Charger SY-1, SY-2 or SY-S Rating a. 50 ampc 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 of LCCA next 59 min. - 10 amps b. 12 ampere-hours C. 125 Vdc Keowee Station Power System Charger No. 1, No. 2 or Standby Rating a. 200 amps ea. Battery No. 1 or No. 2 Capacity b. 233 ampere-hours Actual active load per battery a. First min. - 1031 amps next 59 min.-179.4 amps during 1st hour of LOCA b. 193.6 ampere-hours Redundancy of A.C. Systems There are three 4160 engineered safety feature switchgear buses. Each bus can receive power from either of two 4160 main feeder buses. Each feeder bus in turn can receive power from the 230 KV switchyard through the common 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 from the onsite Keowee Hydro Station or from several offsite sources via transmission lines which connect the Oconee Station with the Duke Power System power distribution network. Redundancy of DC Svstems. A. 125Vdc Instrument and Control Power System All reactor protection.and engineered safety features loads on this system can be powered from either the Unit 1 or Unit 2 125Vdc Instrument 3.7-5

and Control Power Buses. The Unit 1 125Vdc Instrument and Centrol Power 's Buses can be powered from two battery banks and three battery chargers. As shown above, one battery Ce.g., ICA) can supply all loads for one hour. Also, one battery charger can supply all connected ESF and reactor protection loads. B. 125Vdc Switching Station Power System There are two essentially independent subsystems 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: continuously monitor operations of the protective relaying a. b. isolate Oconee (including Keowee) from all external 230 KV grid faults connect on-site power to Oconee from a Keovee hydro unit or c. d. restore off-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 There are essentially two independent physically separated Class I (Seismic) subsystems, 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 automatically or manually start, control and protect one of the hydro units. An open or short in any one battery, charger or de distributien center cannot cause loss of both hydro units. The 230 KV sources, while expected to have excellent availability, are not under the direct control of the Oconee 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 th2 cumulative need to dewater 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 when one hydro unit is out for maintenance, the other unit will be available for service. 3.7-6

In the event that only one hydro unit is available to backup the off-site power sources, and it is considered important for the Oconee Unit i reactor to remain critical or return to criticality from 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 important to continue to maintain the Oconee Unit 1 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 Oconee 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 1 reactor critical or return it to criticality from hot shutdown conditions for a specific period of time rather than require it to remain suberitical or be shutdown. A scheduled shutdown for inspection or a shatdown to perform minor maintenance would not constitute a compelling reason. Factors to consider in justification of such a rare, limited period of criticality l without the hydro station available would include number of offsite 230KV power sources available, availability of Unit 2 startup transformer, availability of the Lee Gas Turbine, weather conditions and all other f actors which could bear on potential for loss of these power sources. Also, the evaluation should show that reactor safety will not be compromised if during operation under such further degradation, an additional loss of ac power should be suffered. i i 1 3.7-7

J 3.8 FUEL LOADING AND REFUELING Aeolicability Applies to fuel loading and refueling operations. Obiective To assure that fuel loading and refueling operations are performed in a responsible manner. Specification 4 Radiation levels in the reactor' building refueling area shall be 3.8.1 monitored by RIA-48 and RIA-49. Radiation levels in the spent fuel storage area shall be monitored by RIA-41. If any of these instruments becomes inoperable, portable survey instrumentation, having the appropriate ranges and sensitivity to fully protect individuals involved in refueling operation, shall be used until the permanent instrumentation is returned to service. Core suberitical neutron flux shall be continuously monitored by 3.8.2 least two neutron flux monitors, each with continuous indication atavailable, whenever core geometry is being changed. When core I gecmetry is not being changed, at least one neutron flux monitor shall be in service. At least one low pressure injection pump and cooler shall be operable. 3.8.3 3.8.4 During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration shall be maintained at less than that required to shutdown the core to a k gg g0.99 e not if all control rods were removed. Direct communications between the control :com and the refueling 3.8.5 personnel in the reactor building shall exist whenever changes in core geometry are taking place. Durir.g the handling of irradiated fuel in the reactor building at 3.3.6 least one door on the personnel and emergency hatches shall be closed. The equipment hatch cover shall be in place with a minimum of four 'olts securing the cover to the sealing surfaces. Both isolation valves in lines containing automatic containment r 3.8.7 isolation valves shall be operable, or at least one shall be closed. When two irradiated fuel assemblies are being handled simultaneously 3.8.8 within the fuel transfer canal, a -imum of 10 feet separation shall 4 be maintained between the assemblic; wt all times. Irradiated fuel assemblies may be handled with the Auxiliary Hoist provided no other irradiated fuel assembly is being handled in the fuel transfer canal. 3.8-1 s

3.8.9 If any of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core ~3 shall cesse; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made-3.8.10 The reactor building purge system, including the radiation monitor, e RIA 45, which initiates purge isolation, shall be tested and verified to be operable immediately prior to refueling operations. f 3.8.11 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours. Bases Detailed written procedures will be available for use by refuell:g 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 i the refueling operations that would result in a hazard to public health and I safety. If no change is being made in core geometry, one flux monitor is f sufficient. This permits maintenance on the instrumentar'on. Continuous moni-toring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The low pressure injection pump is used to maintain a uniform boron concentration. (1) The shutdewn margin indicated in Specification 3.8.4 j will keep the core suberitical, even with all control rods withdrawn from the core. (2) The boron concentration will be maintained above. 1,800 ppm. Although this concentration is sufficient to maintain the core k,ff ~< 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 heff with all rods in the core and with refueling boron concentration is approximately 0.9. Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement. The specification requiring testing of the reactor bui'_ ding 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 3.8.11 is required,as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for 72 hours.(3) REFERENCES (1) FSAR, Section 9.7 (2) FSAR, Section 14.2.t.1 (3) FSAR, Section 14.2.2.1.2 i j 3.8-2

RELEASE OF LlQUID RADI0 ACTIVE WASTE 3.9 Aeplicabilit" Applies to the controlled release of all liquid waste discharged from <he station which may contain radioactive materials. Obiective To establish conditions for the release of liquid waste containing radio-acti"e materials and to assure that all such releases are within the concen-In addition, to assure that the tration limies specified in 10 CFR Part 20. releases of radioactive material in liquid wastes (above background) tothe following liquid unrestricted areas meet the low as practicable concept, release objectives shall apply: The annual total quantity of radioactive materials in liquid waste, shall be less than 5 curies per a. excluding tritium a d dissolved gases, unit; The annual average concentration of radioactive materials in liquid waste, upon release from the Restricted Area, excluding tritium and b. dissolved noble gases, shall not exceed 2 x 10-3 pCi/ml; The annual average concentraticn of tritium in liquid waste, , pCi/ml; shall not exceed 3 x 10-0 c. upon release frem the Restricted Area, The annual average concentration of dissolved gases in liquid waste, exceed 2 x 10-6 d. upon release from the Restricted Area, shall not pC1/ml; Snecifications If the experienced release of radioactive materials in liquid wastes, 3.9.1 these quantities when averaged over a calendar quarter, is such that the same release rate for a year would exceed twice if continued at the annual objectives the licensee will: Sbke an investigation to identify the causes for such release a. rates; Define and initiate a program of action to reduce such release b. rates to the design levels, and; Describe these actions in a report to AEC/ DOL within 30 days after c. incurring the reporting obligation. The release rate of radioactive liquid ef f,uents, excluding tritium 3.9.2 and dissolved noble gases, shall not exceed 10 curies during anySimilarly, calendar quarter without specific appreval of the Cc==issicn. concentratien of tritium released from the the quarterly averag2 Restricted Area shall not exceed 1 x 10-3 uCi/ml. 3.9-1

The rate of release of radioactive =aterials in liquid waste from the 3.9.3 station shall be controlled such that the instantaneous concentrations of radioactivity in liquid waste upon release frem the Restricted Area, does not exceed the values listed in 10CFR20, Appendix B, Table II, Column 2. The equipment installed in the liquid radioactive waste system shall be 3.9.4 maintained and operated for the purposes of keeping releases within the objectives of these specifications and shall process all liquids the activity, excluding prior to their discharge in order to limit tritium and dissolved noble gases, released during any calendar quarter to 1.25 curies or less. As f ar as practicable, the releases of liquid waste shall be coordinated 3.9.5 with the operation of the Keowee Hydro unit. Liquid waste discharged from the liquid waste disposal system shall 3.9.6 be continuously monitored during release. The liquid effluent monitor reading shall be compared with the expected reading of each discharge The monitor shall be tested daily or prior to releases and batch. calibrated at refueling intervals. The calibration procedure shall consist of exposing the detector to a referenced calibration source in The sources and geometry shall be a controlled, reproducible geemetry. referenced to the original monitor calibrction which provides the appli-cable calibration curves. The effluent control monitor shall be set to alarm and automatically 3.9.7 close the waste discharge valve such that the appropriate require-ments of the specification are met. In addition to the continuous monitoring requirements, liquid radioactive 3.9.8 waste sampling and activity analysis shall be perforced in accordance with Table 4.1.3. Records shall be maintained and reports of the sampling and analysis shall be submitted in accordance with Section 6.6 of these Technical Specifications. Bases It is expected that the releases of radioactive materials and liquid wastes will be kept within the design objective levels and will not exceed the concentration These levels provide the reasonable assurance that limits specified in 10CFR20. the resulting annual exposure to the whole body or any individual body organ will At the same time, the licensee is permitted the exceed 5 millirem per year. not flexibility of operation compatible with considerations of health and safety to the public is provided a dependable source of power under unusual assure thatconditions which may temporarily result in releases higher than design operating objective levels but still within the concentration limits specified in 10CFR20. It is expected that when using this operational flexibility under unusual oper-ating conditions, the licensee shall exert every effort to keep the levels of radioactive materials and liquid wastes as low as practicable and that annual releases will not exceed a small fraction of the annual average concentration limits specified in 10CFR20. 3.9-2

The anticipated annual releases from the three Oconee units have been developed taking into account a combination of variables including fuel failures, primary system leakage, primary-to-secondary leakage, and the performance of the various vaste treatment systems. The actual magnitude of these parameters are as follows: a. Maximum expected reactor coolant corrosion product concentrations. b. Reactor coolant fission product concentration corresponding to 0.25 percent fuel cladding defects. Steam generator pri=ary-to-secondary leakage rate of 20 gpd. c. d. 255,160 gallons per year processed by the vaste disposal system in a 30-day hold-up. c. 1,060,800 gallons per year processed by the reactor coolant bleed treat-mant system, f. A Creontamination factor of 104 for all radionuclides except tritium for the coolant bleed and waste evaporators and a decontamination factor of 10 for the demineralizers except for tritium which had an assumed de-contamination factor of 1 for evaporatien-demineralization. g. No re=cval by demineralization for Cs,n, and Y. A decontamination factor o of 103 was used for the evaporation of iodine. h. The decay time of the reactor coolant bleed system was 30 days. The application of the above estimates results in the radionuclide discharge concentrations and rates shown in Table III-12 of the "Fint ' Environmental State.ent Related to Operation of Otonee Nuclear Station Units 1, 2, and 3". Thase concentrations are based on an annual average flow in the Keowee River of 1,100 efs. Operating procedures will identify all equipment installed in the liquid waste handling and treatment systems and will specify detailed procedures for operating and maintaining this equipment. The icwest practicable liquid release objectives expressed in this specification are based on the guidelines contained in the proposed Appendix I of 10CFR50. Since these guidelines have not been adopted as yet, the release obj ectives of this specification will be reviewed at the time Appendix I becomes a regulation to assure that this specification is based upon the guidelines contained therein. 3.9-3

o RELEASE OF GASEOUS RADICACTIVE '4ASTE 3.10 Aoplicability Applies to the controlled release of all gaseous vaste discharged from the station which may contain radioactive materials. _ iective_ Ob To establish conditions in which gaseous vaste containing radioactive materials may be released and to assure that all such releases are within the concen-In addition, to assure that the tratien and dose limits specified in 10CFR20. releases of gaseous radioactive vastes (above background) to unrestricted areas the following objectives shall apply: meet the as low as practicable concept, Averaged over a yearly interval, the release rate of noble gases and other I-131 and particulate radio-isotopes with 1. radioactive isotopes, except the unit vent, shall be half lives greater than eight days, discharged at limited as fo11cws: i 3 4340 m /sec 01 , __(MPC)1 ~ where 01 is the annual controlled release rate (C1/sec) of radio-isotope is the permissible concentration for unrestricted areas in units i and (MPC) 3 (pCi/mi = Ci/m ) for any radionuclide given in Column 1, Table II 3 of C1/m of Appendix 3 to 10CFR20. Averaged over a yearly interval, the release rate of I-131 and other particulate radio-isotopes with half lives longer than eight days, 2. shall be limited as follows: discharged at the unit vent, 3 117 m /sec 01 7 (MPC)1 where Qi and (MPC)1 are as defined above. Specifications If the experienced rate of release of radioactive materials and gaseous wastes when averaged over a calendar quarter is such that 3.10.1 the same release rate for a year these quantities if continued at would exceed twice the annual objective, the licensee shall: Make an investigation to identify the causes for such releases; Define and initiate a program of action to reduce such release a. b. rates to the design levels; to the Ccesission within 30 Describe these actions in a report c. days after incurring the reporting obligation. rate of release of radioactive =aterials and gaseous If the experienced wastes, when averaged over a calendar quarter, is such that these quan-3.10.2 the same release rate for a year would exceed tities if continued at 3.10-1

e eight times the annual objectives, the licensee shall define and initiate a program of action to assure that such release rates are reduced and shall submit a report to the AEC within seven days af ter ) incurring the reporting requirement, describing the causes for such release rates and the course of action taken to reduce them. 3.10.3 The rate of release of radioactive sacerials and gaseous sastes frca the station (except I-131 and particulate radio-isotopes with half lives greater than eight days) shall be controlled such that the maximum release rate averaged over any one-hour period shall not exceed: 3 5 m /sec 3.0 x 10 01 = (MPC)1 3.10.4 During release of radioactive gaseous waste from the gaseous waste tanks to the unit vent, the following conditions shall be cet: The gaseous radioactivity monitor, the iodine monitor, and the a. 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 except, when j under unusual conditions the filter system is inoperable gaseous wastes shall be held up for the maximum period practicable prior to release. Every reasonable effort shall be made to return inoperable filters to the operable condition before releases to the environment are made. The gaseous waste tanks shall be maintained and operated for the 3.10.5 a. purposes of keeping releases within the objectives of these specifications and shall process all radioactive gases from the vent gas header prior to their release in order to limit the activity released during any calendar quarter to one fourth the annual release quantities or less as determined by the objectives of this specification, b. The maximum activity to be contained in one gaseous waste tank shall not ex.c_eed 17, 20Q'E curies. E will be assumed to be the same as the E of the noble gases in the reactor coolant syste= l as determined in accordance with Table 4.1-3 of Specification 4.1.2. As f ar as practicable, release of radioactive gas will be co-c. ordinated with favorable meteorological conditions. 3.10.6 During power operation, whenever the air ejector off-gas monitor is inoperable, grab samples shall be taken from the air ejector discharge and analyzed for gross radioactivity daily. 3.10.7 Gases discharged through the unit vent shall be continuously monitored for gross noble gas and particulate activity. Whenever either of these monitors is inoperable, appropriate grab sa=ples shall be taken and analyzed daily. j 3.10-2

3.10.8 The reactor building shall not be purged unless the followirg conditions are met: Reactor building purge shall be through the high efficiency parti-a. culate filters and charcoal filters until the activity concentration is below the occupational limit inside the reactor building, at which time bypass may be initiated, b. If reactor building is purged, the purge shall be through the high efficiency particulate filters whenever irradiated fuel is being handled or any objects are being handled over irradiated fuel in the reactor building. 3.10.9 In addition to the above continuous sampling and monitoring requirements, gaseous radioactive waste sampling and activity analysis shall be per-formed in accordance with Table 4.1-3. Records shall be maintained and reports of the sampling and analysis results shall be submitted in accordance with Section 6.6 of these specifications. Bases It is expected that the releases of radioactive materials and gaseous wastes will be kept within the design obje:tive levels and will not exceed on an instantaneous basis the dose rate limits specified in 10CFR20. These levels provide reasonable assurance that the resulting annual exposure from noble gases to the whole body or any organ of an individual will not exceed 10 mrem per year. At the same time, the licensee is permitted 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 the design objective levels but still within the concentration limits specified in 10CFR20. It is expected that using this operational flexibility under unusual operating conditions, the licensee shall exert every effort to keep levels of radioactive materials and gaseous wastes as low as practicable and that annual releases will not exceed a small fraction of the annual average concentration limits specified in 10CFR20. These efforts shall include considerativa of meteorological con-ditions during releases. The anticipated annual releases from the three Oconee reactor units have been developed taking into account a combination of system variables including fuel failure, primary system leakage, and the performance of radio-isotope removal mechanisms. The values assumed for these variables include the following: a. Reactor coolant fission product concentration corresponding to 0.25 percent fuel cladding defects; b. Steam generator primary-to-secondary leakage rate of 20 gpd; c. Reactor coolant leakage to the containment building of 120 gpd and 12 containment vents per year; d. Primary coolant stripped 12 times per year; 3.10-3

I Decey time of the waste gas processing system - 30 days; ~') e. f. Decontamination fact 7r of 1,000 for iodine in the evaporator; Charcoal filter decontamination factor of 10 for iodine removal in the g. purge exhaust system. The application of the above estimates results in the radio-gas discharge rates shown in Table III-13 of the " Final Environmental Statement Related to Operation of Oconee Nuclear Station Units 1, 2, and 3". The noble gas release rates stated in the objectives are based on a X/Q value from the annual meteorological data. The dispersion factor used, 4.61 x 10-6 sec/m, is conservative and the release rate is controlled to a small fraction 3 of 10CFR20 requtrements at the exclusion area boundary (.02 of 10CFR20 = less than 10 mrem per year). The I-131 and particulate release rates stated in the objectives limits the concentration at the exclusion area boundary to much less than 1 percent of the MPC listed in 10CFR20. The release rate also controls the expected iodine dose due to the milk pathway (using concentration factor in the milk pathway of 700) at the nearest cow and the nearest dairy (taken as five miles west, 3 X/Q = 1.22 x 10-7 sec/m ) to less than five millirem per year. This meets the intent of proposed Appendix I to 10CFR50. A survey will be conducted once per year to assure that no milk producing cows are within a five mile radius of the plant. -s -) The maximum one-hour release rate limits the dose rate at the exclusion area l boundary to less than 2 mrem per hour even during periods of unfavorable meteorology (using conservative meteorological conditions, i.e., two hour X/Q of 1.16 x 10- sec/m3 accident meteorology). l Themaximumactivityinagaseouswastetankisspecifiedas17,200 curies [E l based on a postulated tank rupture that allows all of the contents to escape I to the atmosphere. This specification limits the maximum off-site dose to l well below the limits of 10CFR100. ( l The lowest practicable gaseous release objectives expressed in this specifi-l cation are based on the guidelines contained in the proposed Appendix I of 1 10CFR50. Since these guidelines have not yet been adopted, the release objectives of this specification will be reviewed at the time Appendix I I becomes a regulation to assure that this specification is based upon the guidelines contained therein. l l l l l l l 3.10-4 i 1

. ~ -- 4 i r i 3.11 MAXIMUM POWFR RESTRICTION 4 i Applicability Applies to the nuclear steam supply system of Unit i reactor. Obj ective To restrict operations below significant power level until system perfor=ance with regard to reactor internals has been verified as acceptable. Specification Unit i power level may not be increased above 128 MWt (5.0% of full rated ' power) until the results of vibration monitoring during preoperational testing have been evaluated and subsequent approval is granted by the AEC/ COL staff. Bases Safety Evaluations done as part of the Final Safety Analysis Report were done for power levels of 2568 MWt. Since these evaluations,0conee Unit i suffered failures of internal compenents during preoperational t_ sting. The internal components which failed have been redesigned, installed in the reactor syste= and tested during additional preoperational tests. Since this is the first nucicar steam supply of this design to go into service, power level is being restricted until a complete evaluation of the preoperational vibration testing of vessel internals has been made. It is anticipated that the results of the vibration tests will be reported and evaluated prior to power escalation after cccpletion of low power physics tests. This pceer restriction permits low power physics testing but dees not permit pow'_r escalation. This temporary restriction will be lif ted pending acceptabla results from the preoperat10nal vibration tests. Subsequent power restrictions based upon evaluation of fuel densification or other safety considerations =ay be considered following the lifting of the above restriction. 3.11-1

3.12 REACTOR 3UILDING POLAR CLCE A"D AUXILIARY HOIST Acolicabilitr Applies to the use of the reactor building polar crane ever the stea:

enerator compartnents and the fuel transfer canal and the Auxiliary Hoist ever the fuel transfer canal.

Obiective To identify those conditions for which the operation of the reactor building polar crane and auxiliary hoist are restricted. Snecification 3.12.1 The reactor building polar crane shall not be operated over the fuel transfer canal when any fuel assembly is being moved. 3.12.2 The auxiliary hoist shall not be operated over the fuel transfer canal when any fuel assembly is being moved unless the hoist is being used to nove that assembly. 3.12.3 Dr. ring the period when the reactor vessel hecd is removed and irradiated fuel is in the reactor building and fuel is not being .r.c V e d, the reacter building polar crane and auxiliary hoist shall be operated over the fuel transfer canal only where necessary and in accordance with approved operating procedures stating the purpose of such use. 3.12.1 When the reactor vessel head is recoved 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 elevation of the fuel transfer canal. 3.12.5 During the period when the reactor coolant system is pressurized above 300 psig, and is cbove 200*F, and fuel is in the core, the reactor building polar crane shall not be operated over the steam generator compartments. Bases Restriction of use of the reactor building polar crane and auxiliary hoist over the fuel transfer canal when the reactor vessel head is recoved to those operations necessary for the fuel handling and core internals operations 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 nornally handle loads. Restriction of use of the reactor building polar crane over the steam generatcr cenpartments during the time when steam could be formed frca dropping a lead en the steam generator or reacter coolant piping resulting in rupture of the syster is required to protect against a loss of coolant accident. 3.12-1

i i 3.13 SECONDARY SYSTEM ACTIVITY Anelicabilttv Applies to the limiting conditions of secondary system activity for operation of the reactor. Obiective To limit the maximum 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 determining a maximum allowable secondary coolant activity, the activity contained in the mass released following a loss of load accident is censidered. 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 dose limit of 1.5 res is used. ihe wnola body dose is negligible since any noble gases entaring the seccndary coolant system are continuously vented to the atmosphere by the condenser air ej ector, thus, in the event of a loss of load incident there are only small quantities of these gases which would be released. I-131 is the significant isotope because of its icw MFC 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 en 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) = Ci V'B DCF *(0.1) X/Q C - Secondary coolant activity (2.0 uCi/cc I-131 equivalent) 3 =Secondarywatervolumereleasgdtoatmosphere(90m) C

= 3reathing rate (3.47 x 10-4 m /sec)

X/Q ' Ground level release dispersion f actor (1.16 x 10-4 sec/m ) 3 DCF = 1.43 x 106 res/Ci 0.1 = Fraction of activity released The resultant dose is 1.15 rem compared to the Radiation Protection Guide of 1.3 ren for an annual individual exposure in an unrestricted area. 3.13-1

i 4 SURVEILLANCE STANDMDS 3pecified intervals may be adjusted plus or minus 25% to accommodate normal test schedules. 1 4 4.1 OPERATIONAL SAFETY REVIE'J Applicability i Applies to items directly related to safety limits and limiting conditions f 2 for operation. i Obiective_ To specify the minimum frequency and type of surveillance to be applied to i j f unit equipment and conditions. f Specification The minimum frequency and type of surveillance required for reactor 4.1.1 l protective system and engineered safety feature protective system instrumentation when the reactor is critical shall be as stated in Table 4.1-1. and sampling test shall be performed as detailed in 4.1.2 Equipment Tables 4.1-2 and 4.1-3. 1 Bases Check Failures such as blown instrument fuses, defective indicators, faulted amplifiers which result in " upscale" or "downscale" indication can be d easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm i or annuniciator action. Ccmparison of output and/or state of independent j channels measuring the same variable supplements this type of built-in Based on experience in operation of both conventional and surveillance. nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation. Calibration } Calibration shall be performed to assure the presentation and acquisition of The nuclear flux (power range) channels amplifiers shall accurate information. 4 . be calibrated (during steady state operating conditions) when indicated neutron During non-steady l power and core thermal power differ by more than 2 percent. state operation, the nuclear flux channels amplifiers shall be calibrated daily to compensate for instrumentation drift and changing rod patterns and core physics parameters. 1 Channels subject only to " drift" errors induced within the instrumentation f itself can. tolerate longer intervals between calibrations. Process system instrumentation errors induced-by drift can be expected to remain within - l 4.1-1 l f i

acceptable tolerances if recalibration is performed at the intervals of each refueling period. 's, Substantial calibration shif ts within a channel (essen:ially a channel failure) will be revealed during routine checking and testing procedures. Thus, minimum calibration frequencies set forth are considered acceptable. Testing Oc-line testing of reactor protective channels is required once every four weeks on a rotational or perfectly staggered basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel. The rotation schedule for the reactor protective channels is.as follows: Channels A, B, C & D Before Startup Channel A One Week Afte'r Startup Channel B Two Weeks After Startup Channel C Three Weeks After Startup Channel D Four Weeks After Startup -) The reactor protective system instrumentation test cycle is continued with one channel's instrumentation tested each week. Upon detection of a failure that prevents trip action, all instrumentation associated with the protective channels will be tested after which the rotational test cycle is started again. If actuation of a safety channel occurs, assurance will be required that actuat'on was within the limiting safety system setting. The protective channels coincidence logic and control rod drive trip breakers are trip teste.d 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 'our protective channels shall be trip tested prior to startup and their individual channels trip tested on a cyclic basis. Discovery of an unsafe failure requires the testing of all channel logic and breakers, after which the trip test cycle is started again. The equipment testing and system sampling frequencies specified in Table 4.1-2 and Table 4.1-3 are considered adequate to maintain the equipment and systems in a safe operational status. REFERENCE FSAR, Section 7.1. 2.3.4 4.1-2

TABLE 4.1-1 INSTRUFIENT SURVEILLANCE REQUIRD1ENTS CHANNEL DESCRIPTION CHECK TEST _ CALIBRATE RDIARKS 1. Protective Chanael NA M NA Coincidence Logic 2. Control Rod Drive Ni M NA Trip Breaker 3, Power Range Amplifier D(1) NA (1) (1) lleat Balance Check daily. lleat balance calibration whenever indicated neutron power & core thermal power differ by more than 2%. 4. Power Range Channel S M M(1)(2) (1) Using incore instrumentation. (2) Axial offset upper & lower chambers 7 after each startup if not done pre-vious week. S. Intermediate Range Channel S(l) P NA (1) Uhen in service 6. Source Range Channel S(l) P NA (1) When in service S M R 7. Reactor Coolant Temperature Channel S M R 8. liigh Reactor Coolant Pressure Channel 9. l.ow Reactor Coolan: S M R Pressure Channel 10. Flux-Reactor Coolant S M R Flow Comparator 11. Reactor Coolant Pressure S M R Temperature Comparator 12. Pump Flux Comparator S M R

TABLE 4.1-1 Cont. CHANNEL DESCRIPTION CHECK TEST CALIBRATE REHARKS 13. High Reactor Building Pressure Channel D M R 14. High Pressure Injection Logic Channel NA M NA 15. High Pressure Injection Analog Channels a. Reactor Coolant Pressure Channel S R b. Reactor Building f-4 psig Channel S M R Y 8' 16. Low Pressure Injection Logic Channel NA M NA 17. 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 System Logic Channel NA M MA 19. Reactor Building Emergency Cooling and Isolation System Analog Channels a. Reactor Building 4 psig Channels S M R = 4 'i

TABLE 4.1-1 Cont. Cll!.h.,dL DESCR IPTION CllECK TEST CALTBRATE RDIARKS 20. Reactor Building Spray System Logic Channel NA M NA 21. R actor Building Spray e System Analog Channels a. Reactor Building liigh Pressure Channels NA M R 22. Pressurizer Temperature Channels S NA R 23. Control Rod Absolute Position S(1) NA R(2) (1) Check with Relative Position Indicator t-(2) Calibrate rod misalignment channel On 24. Control Rod Relative Position S(1) NA R(2) (1) Check with Absolute Position Indicator (2) Calibrate rod misalignment channel 25. Core Flooding Tanks a. Pressure Channels S NA R b. Level Channels S NA R 26. Pressurizer Level Channels S NA R 27. Letdown S w age Tank Levels Ch snels D NA R 28. Radiation Monitoring l l Sys ams W(1) M Q (1) Check functioning of self-checking feature on each detector. 29. liigh and Low Pressure Injection Systems: Flow Channels NA NA R

~_ - NP TABLE 4.1-1 Cont. ,CilANNEL DESCRIPTION CllECK TEST CALIBRATE REMARKS s (10. Borated Water Storage Tank Level Indicator W NA R

3.. Boric Acid Mix Tank a.

Level Channel NA NA R b. Temperature Channel M NA R 32. Concentrated Boric Acid Storage Tank a. Level Channel NA NA R f" b. Temperature Channel M NA R Y 33. Containment Temperature NA NA R 34. Incore Neutron Detectors M(1) NA NA (1) Check functioning; iacluding functioning of compu..er ceadout or recorder readout 35. Emergency Plant Radiation Instruments 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 R Sump Level 39. Steam Ceaerator Water Level W NA R - i 40. Turbine dverspeed Trip NA NA R 41. Engineered Safeguards NA R NA Channel 1 IIP Injection Manual Trip w

f----- TAbl.E 4.1-1 Cont. CilANNa'L DESCRIPTION CilECK TEST cal.IliRATE REMARKS 42. Engineered Safeguards MA R NA Channel 2 IIP Injection Manual Trip 43. Engineered Safeguards NA R NA Channel 3 LP Injection Manual Trip 44. Engineered Safeguards NA R NA Channel 4 I.P Injection flanual Trip 45. Engineered Safeguards NA R NA Channel 5 RB Isolation & Cooling !!anual Trip 46. Engineered Safeguards NA R NA L Channel 6 RB Isolation & Cooling Manual Trip 47. Engineered Safeguards NA R NA Channel 7 Spray Manual Trip 48. Engineered Safeguards NA R NA Channel 8 Spray llanual Trip 49. Reactor Manual Trip NA P NA S - Each Shif t R - Each Refueling Period D - Daily NA-Not Applicable W - Weekly Q - Quarterly t1 - tionthly P - Prior to ea:h startup if not donc previous week

Table 4.1-2 Minimum Equipment Test Prequency_ Frecuencv Test I t en_ s Rod Drop Times of all Each Refueling shutdown 1. Control Rods full length rods (1) Movement of each red Every two weeks 2. Control Rod Movement 50% each refueling period J. Pressurizer Safety Valves Setpoint 257. each refueling period 4. Main Steam Safety Valves Setpoint Each refueling period Functional 5. Refueling System Interlocks Turbine Steam Stop Valves (1) Movement of each stop Monthly 6. valve 7. Reactor Coolant System (2) Evaluate Daily Leakage Each refueling period 8. Charcoal and High DOP Test on HEPA Efficiency Filters for filters. Freon Test and at any time work Penetration Room, Control on Charcoal Filter on filters could alter their integrity. Room, and R3 Purge Filters Units Each refueling period 9. Ccndenser Cooling Water Functional System Gravity Flow Test Functional Monthly 10. High Pressure Service Water Pumps and Power Supplies Each refueling period Functional 11. Spent Fuel Cooling System prior to fuel handling (1) Applicable only when the reactor is critical is above 200*F and at a steady state (2) Applicable only when the reactor coolant temperature and pressure. 4.1-8

l 1 TABLE 4.1-3 t MINIMUM SAMPLING FREQUENCY Item Check Frequency i. l. Reactor Coolant a-. Gamma Isotopic Analysis a. Monthly l b. Radiochemical Analysis for b. Monthly Sr 89, 90 c. Tritium c. Monthly d. Gross Beta & Gamma Activity (1) d. S times / week e. 5 times / week Chemistry (C1, F and 0 ) e. 2 f. Boron Concentration f. 2 times / week n i b g. Gross Alpha Activity

g.. Monthly h.

E Determination (2) h. Semi-annually I 2. Borated Water Storage Boron Concentration Weekly and after i Tank Water. Sample each makeup -3. Core Flooding Tank Boron Concentration Monthly and after each makeup t 4. Spent Fuel Pool Water Boron Concentration Monthly and after each makeup Sample 5. Secondary Coolant a. Cross Beta & Gamma Activity a. Weekly b. Iodine Analysis (3) j 6. Concentrated Boric Boron Concentration Twice weekly j Acid Tank

TABLE 4.1-3 Cont. MINIMUM SAMPLTNd FREQUENCY Item Check Frequency Sensitivity of Waste-Analysts in Lab 7. Low Activity Waste a. Gross Beta & Camma Activity a. Prior to release

a. <10-7 pCi/ml Tank & Condensate of each batch Test Tank b.

Radiochemical Analysis b. Monthly

b. <10-8 pCi/ml Sr 89, 90 c.

Gamma Analysis including c. Monthly

c. Camma Nuclides <5x10-7 pCi/ml Dissolved Noble Cases Dissolved Gases <10-5 pC1/ml o

d. Tritium d. Monthly

d. <10-5 pCi/ul r

d e. ifonthly

e. <10-7 pCi/ul Cross Alpha Activity e.

f. Ba-La-140, I-131 f. Weekly Pro-

f. <5x10-7 pC1/mi portional 8.

Waste Gas Decay Tank a. Camma Isotopic Analysis a. Prior to release

a. <10-4 pCi/ae of each batch b.

Cross Camma Activity b. Prior to release

b. <10-ll pci/cc of each batch c.

Tritium c. Prior to release

c. <10-6 pCi/t c of each batch 3

Iodine Spectrum (4) a. Weekly

a. <10-10 pCi/cc 9.

Unit Vent Sampling a. b. Partic.ilates (4)

1) Gross Beta & Gamma Activity

1) Weekly

1) <10-11 i.Ci/cc

2) Gross Alpha Activity

2) Quarterly on a

2) <10-11 3.C1/cc sample of one week duration s

1

TABLE 4.1-3 Cont. MINIMUM S A M P 1. I N G FREOUENCY Sensitivity of Waste Analysis in lab Frequency Check Item

3) <10-10 pC1/cc

3) Monthly Composite

3) Gamma Isotopic Analysis

4) Quarterly Composite

4) <10-11 pC1/cc

4) Radiochemical Analysis Sr 89, 90

5) <10-10 pC1/cc

5) Weekly

5) Ba-l.a-140, 1-131 Annually Measure Leakage Flow Rate 10.

Keowee flydro Dam Dilution Flow _o One time if and when primary Measure Iodine Partition to secondary leaks develop 11. Condenser Air Ejector Factor in Condenser Partition Factor <10-4 pC1/cc a. a. Each Purge Gamma Isotopic Analysis 12. Recctor Building Purge a. b. <10-11 pC1/cc b. Each Purge b. Gross Gamma Activity <10-6 pC1/cc c. c. Each Purge c. Tritium limits of Specification 3.1.4, the sampling When rtdioactivity level is greater than 10 percent of the (1) increased to a minimum of once each day. fiequency shall be E determination will be started when gross beta-ga ma activity analysis ind! cates greater than 1 A radiochemical increase in gross beta-gamma activity analysis. (2) and will be redetermined each 10 pCf/mi % of radionuclides in reactor analysis for this purpose shall consist of a quantitative neasurement of 95of gamma isotopic analysis of This is expected to consist coolant with half lives of >30 minutes. radiochemical analysis for Sr 89, 90, and tritium primary coolant including dissolved gaseous activities, analysis.

TABLE 4.1-3 Cont. MINIMUM SAMPLTNC FREQUENCY (3) When gross activity increases by a factor of two above background, an lodlue analysis will be made and performed ther.after when the gross beta-gamma activity increases by 10 percent. (4) When activity level exceeds 10 percent of the limits of Specification 3.9, the sampling frequency shall be increased to a minimum of once each day. This can be done by RIA-44 (Unit Vent Iodine) monitor. When the gross activity release rate exceeds one percent of maximum release rate and the average gross activity release rate increases by 50 percent over the previous day, an analysis shall be performed for iodines and particulates. This can be done by RIA-44 (Unit Vent Iodine Monitor) and RIA-43 (Unit Vent Particulate Monitor). ? 7' C De l 2

4.2 REACTOR COOLANT SYSTEM SURVEILLANCE Applicability Applies to the surveillance of the reactor coolant system pressure bar:ndary. Obiective To assure the continued integrity of the reactor coolant system pressure boundary. Snecification 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 accor-dance with the methods and intervals indicated in IS-242 and 15-261 of Section XI of the ASME Boiler and Pressure Vessel Code, 1970, including 1970 Winter addenda, except as follows: IS-261 Item Connonent Excention 1.4 Primary Nozzle to Vessel 1 RC outlet nozzle to be Welds inspected after approx. 3 1/3 years operation. 2nd RC outlet nczzle to be inspected after approx. 6 2/3 yrs operation. 4 RC inlet nozzles and 2 core flood!ng nozzles to be in-spected at or near end l of interval 3.3 Primary No :le to Safe End Not Applicable Welds 4.3 Valve Pressure Retaining Not Applicable Bolting Larger than 2" 6.1 Valve Body Welds Not Applicable 6.3 Valve te 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 4.2-1

d result of the tests outlined in Table IS-261 of Section XI of the -s code, that defects have developed or grown, shall be investigated, j including evaluation of comparable areas of the reactor coolant system. 4.2.4 To assure the structural integrity of the reactor internals through-out the life of the unit, the two sets of main internals bolts i (connecting the core barrel to the core support shield and to the icwer grid cylinder) shall 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 shut-down. The core barrel to core support shield caps will be inspected each refueling shutdown. 4.2.5 Sufficient records of each inspection shall be kept to allow com-parison and evaluation of future inspections. 4.2.6 The inservice inspection program shall be reviewed at the end of five years to consider incorporation of new inspection techniques and equiptent which have been proved practical and the conclusions of this review and evaluation shall be discussed with the AEC/ DOL. 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 l intervals. A surface examination of exposed surfaces, and a com-plete volumetric examination at approximately 10 year intervals when disassembly and/or flywheel removal is required for main-tenance or repair. Disassembly or flywheel removal is not required to perform 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 withdrawn shall be tested in accordance with ASTM-E-185-70. A copy of the test report shall be forwarded to DOL within 90 days. 4.2.9 During the first two refueling periods, two reactor coolant system piping elbows 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 are numbers 7 and 9 in assemblies A57 and 357 respectively as identified in B&W Report 1364 dated December 1970. 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 Coolant Systems, 1970, including 1970 Winter addenda, edition. The program places major emphasis on the area cf highest stress concentrations and on areas / i i 4.2-2 -m r' ?s-"- a-mta--6 -usi y p-- ,nr-sr y+r m d up- -e-f=>T==, y-tv 7

==,W--- ~t'w f Q

ficient to change material properties. where fast neutron irradiation =ight be suf sure years The vessel specinen surveillance program is based on equivalent expoThe 1.7 life; and the 11.18 of 1.76, 11.18, 20.41, and 30.0. of NDTT; the 30-year figure is based on 3/4 vessel service between. and the 20.41 year figures are equally space' d d desir-Early inspection of reactor coolant system piping elbows is consi ere l base metal when able in order to reconfirm the integrity of the carbon steeIf no degradation is explosively clad with sensitized stainless steel. observed during i ents will Vessel Code. to Section XI of the ASME Boiler and Pressure revert 1 l l 4.2-3

4 4.3 TESTING FOLLOWING OPENING OF SYSTEM 7 Applicability Applies to test requirements for Reactor Coolant System integrity. Objective To assure Reactor Coolant System integrity prior to return to critic.111ty following normal opening, modification, or repair. Specification 4.3.1 When Reactor Coolant System repairs or modifications have been made, these repairs or modifications shall be inspected and tested to meet all applicable code requirements prior to the reactor being =ade 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. Bases Repairs or modifications made to the Reactor Coolant System are inspectable and testable under applicable codes, such as USAS B31.7, code for pressure piping, Nuclear Power Piping dated February, 1968, and as corrected for errata under date of June,1968, and ASME Boiler and Pressure Vessel Code, Section XI, IS-400, dated January,1970. REFERENCES FSAR, Section 4 i i i i 4.3-1

4.4 REACTOR BUILDING 4.4.1 Containment _ Leakage Tests Applicability Applies to containment leakage. Obiective To verify that leakage from the reactor building is maintainad within allow-able limits.

pecification 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 exceed 0.25 the 59 psig design pressure, P, shall not p

that pressure per 24 building at weight percent of the building atmosphere at hours. 4.4.1.1.2 Testing at Reduced Pressure a test The periodic integrated leak rate test may be performed at of not less than 29.5 psig provided the resultant pressure, Pt, leakage rate, Lt, does not exceed a pre-established fraction of La determined as follows: Prior to reactor operation the initial value of the integrated leakage rate of the reactor building shall be measured at a. design pressure and at the reduced pressure to be used in the The leakage rates thus periodic integrated leakage rate tests. measured shall be identified as L m and L m respectively. t p b. L shall not exceed La L f r values of Lem below 0.7. tm e Lm L mj p p above 0.7. c. Le shall not exceed L P for values of Lem a .P Lm ^\\ p p is less than 0.3, the initial integrated test If Lem/ Lpm d. results shall be subject to review by the AEC to establish an acceptable value of L. t 4.4-1 L

4.4.1.1.3 Conduct of Tests (a) The test duration shall be at least 24 hours, except that if both the following conditions are met, the test duration shall be at least 10 hours. (1) All test conditions, including the test procedure, shall be similar to the initial integrated leakage rate tests. (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 deconstrate 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 adjustment. 4.4.1.1.4 Frequency of Test After the initial preoperational leakage rate test, two integrated leakage rate tests shall be performed at approximately equal inter-vals between each major shutdown for inservice :nspection to be performed at 10 year intervals. In addition, an integrated test shall be performed at each 10 year interval, coinciding with the ~ inservice inspection shutdown. The test shall coincide with a shut-down for major fuel reloading. 4.4.1.1.5 Conditions for Return to Criticality a. If L is less than 50% of the value permitted in 4.4.1.1.2, e local leakage rate testing need not be completed prior to return to criticality following a periodic integrated leakage rate test. b. If L is between 50 and 100% of the value permitted in 4.4.1.1.2 e the return to criticality will be permitted conditioned upon demonstrating that local leakage rate into the penetration room measured at full design pressure, accounts for all leakage above 50% of that permitted by 4.4.1.1.2. If this cannot be demonstrated within 30 days of returning to criticality, the reactor shall be shutdown. 4.4.1.1.6 Corrective Action and Retest I If repairs are necesssry 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 demonstrate that the leakage rate reduction achieved by ' ) repairs reduces the overall measured integrated leak rate to an acceptable value. 4.4-2

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 Sufficient data used and a summary of local leak detection tests. and analysis shall be included to show that a stab'ilized 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 i instrumentation sensitivities and data scatter. 4 4.4.1.2. Local Leakage Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be meaeured for each of the following components: (a) Personnel hatch (b) Emergency hatch (c) Equipment hatch seals (d) Fuel transfer tube seals (e) Reactor building normal sump drain line (f) Reactor coolant cump seal outlet line (g) Reactor coolant pump seal inlet line (h) Quench Tank drain line (i) Quench Tank return line (j) Quench Tank vent line (k) Normal makeup to reactor coolant system (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 i 4.4 -_.

a J 4.4.1.2.2 Conduct of Tests ) (a) Local leak rate tests shall be performed at a pressure of not less than 59 psig. (b) Acceptable 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 j local leakage, the reactor shall be shutdown and depressu-rized until repairs are effected and the local leakage meets l the acceptance criterion as demonstrated by retest. 4.4.1.2.5 Test Frequency 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 four month intervals, except when the hatches are not opened during that interval. In no case shall the test interval be longer than 12 months. 1 4.A.l.3 Isolation Valve Functional Tests Every three months, remotely operated reactor building isolation valves shall 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. l 4.4.1.4 Annual Inspection I A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be s uncover any evidence of deterioration which may affect either the ,).. performed' annually and prior to any integrated leak test, to containment's structural integrity or leak-tightness. The dis-4.4-4

covery of any significant deterioration shall be acccmpanied by corrective actions in accord with acceptable procedu leak test. Such tical, prior to the conduct of any integrated repairs shall be reported as part of the test results. Reactor Building Modifications 4.4.1.5 h Any major modification or replacement of components affecting t e reactor building integrity shall be followed by either an inte-and grated leak rate test or a local leak test, as appropriate,th 23 shall meet respectively. Bases _(1) f 59 psig and a The reactor building is designed for an internal pressure oPrior to initial operation, the con-steam-air mixture temperature of 286*F. ure and leak rate tainment will be strength tested at 115% of design pressThe containment will also be These tests tested at the design pressure. design pressure. to initial operation at approximately 50% of thethe leakage rate from react i atis-will verify that fies the relationships given in the specification. i lant life The performance of a periodic integrated leakage rate test dur ng p h tafroent in provides a current assessment of potential leakage frcm t e con In f the cons.inment. case of an accident that would pressurize the interior o f the containment order to provide a realistic appraisal of the integrity o f rmed without pre-under accident conditions, this periodic test is to be per oli l ion valves are to be closed in the normal manner.for the periodic integ vide an accurate measurement of the leakage rate and it dupThe specification provide leakage rate test at 29.5 psig. 29.5 psig to the potential leakage at relating the measured leakage of air atThe minimum of 24 hours was specif racy and to better test to help stabilize conditions and thus improve accuThe frequency of the p 59 psig. tor, because th.se tests evaluate data scatter. test is keyed to the refueling schedule for the reac can best be performed during refueling shutdowns. ts is based on The specified frequency of periodic integrated leakage rate tesis the low proba First 0.25% leakage three major considerations. If ner, because of conformance of the complete containment to ace of any signifi-b im a at 59 psig during pre-operational testing and the a sen fre-Second is the more stresses in the liner during reactor operation. f the containment envelope q sent testing, at design pressure, of those portions olikely to deve cent that is specified that are most leaka33 and isolation valves) and the low value (0.125%) of Third is the tendon as acceptable from penetrations and isolation velves. stres important part of is maintained. the structural integrity of the containment 4.4-5

e Leakage to the penetration room, which is permitted to be up to 50% of the ~ total allowable containment leakage, is discharged through high efficiency ) particulate air (HEPA) and charcoal filters to the unit vent. The filters are conservatively said to be 90% ef ficient for iodine re= oval. i More frequent testing of various penetrations is specified as these locations are more susceptible to leakage than the reactor building liner due to the mechanical closure involved. Particular attention is given to testing those penetrations with resilient sealing materials, penetrations that vent directly to the reactor building atmosphere, and penetrations that connect to the j reactor coolant system pressure boundary. The basis for specifying a maxi-cum leakage rate of 0.125% from penetrations and isolation valves is that i one-half of the actual integrated leakage rate is expected from those sources. Valve operability tests are specified to assure proper closure or i; opening of the reactor building isolation valves to provide for isolation of functioning of Engineered Safe:y Features systems. Valves will be stroked to the position required to fulfill their safety function unless it is established that such testing 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. REFERENCES (1) FSAR, Sections 5 and 13. 4 4 i 1 4.4-6

1 4.4.2 Structural Integrity Acolicability a Applies to the structural integrity of the reactor building. Obiective To define the inservice surveillance program for the reactor building. Specification 4.4.2.1 Tendon Surveillance For the initial surveillance program, covering the first five years of operation, nine tendons shall be selected for periodic inspection 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 1D28 2D2S 3D28 Horizontal 13H9 SlH9 53H10 Vertical 23Vl4 45V16 61V16 4.4.2.1.1 Lift-Off 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 from each relaxed tendon shall be removed as a sample and visually inspected for corrosion or pitting. Tensile tests shall also be performed on a minimum of three specimens taken from the ends and middle of each of the three wires. The specimens shall be the maximum length acceptable for the test apraratus to be used and shall include areas representative of significant corrosion or pitting. After the wire removal, the tendons shall be retensioned to the stress level measured at the lift-off reading and then checked by a final lift-off reading. 4.4-7

\\ Should the inspection of one of the wires reveal any significant t corrosion (pitting or loss of area), further inspection of the other two sets in that directional group will be made to determine the extent of the corrosion and its significance to the load-l 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 tested. 4.4.2.2 Inspection Intervals and Reports The initial inspection shall be within 18 =onths of the initial Reactor Building Structural Integrity Test. The inspection intervals, measured from the date of the initial inspection, shall be two years, four years and every five years thereafter or as modified based on experience. Tendon surveillance may be conducted during reactor operation provided design conditions regarding 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 6.6.3.5) and shall especially address the following conditions, should they develop: (1) Broken wires. (2) The force-time trend line for any tendon, when extrepolated, that extends beyond either the upper or lower bounds of the predicted design band. (3) Unexpected changes in corrosion conditions or sheathing filler properties. 4.4.2.3 End Anchorage Concrete Surveillance a. The end anchorages of the surveillance tendons and adjacent concrete surface will be inspected. In addition, other locatiot.s for surveillance will be determined by information obtained from design calculations, prestressing record observations, and deformation measurements made during pres..ressing. b. The inspection interval will be one-half year and one year after the operation of Unit 1 and will occur during the warmest and coldest part of the year. c. The inspections made shall include: (1) Vis"sl 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. I 4.4-8 l

i i (3) Measurement of concrete temperatures at specific end anchorage concrete surfaces being inspected. (4) The mapping of the predominanc visible concrete crack patterns. (5) The measurement of the crack widths, by use of optical comparators or wire feeler gauges. (6) The measurement of movements, if any, by use of demount-able mechanical extensometers. The measurements and observations shall be compared with those d. to which prestressed s tructures have been subjected in normal and abnormal load conditions and with those of preceding measurements and observations at the same location on the reactor containment. The acceptance criteria shall be as follows: e. If the inspections determine that the conditions are favorable in comparison with experience and predictions, the close inspect-ions will be terminated by the last of the inspections stated in the schedule and a report will be prepared which documents the findings and recommends the schedule for future inspections, if If c>a inspections detect symptoms of greater than normal any. cracking or movements, an immediate investigation will be made to determine the cause. 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: Location of areas which have inward deformations. The magnitude a. of the inward 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 stiffeners which are on 15-inch centers. The measurements shall be accurate to i.01 inch. Temperature readings shall be obtained on both the liner plate and outside containment wall at the locations where inward deformations occur. Locations of areas having strain concentrations by visual exami-b. 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 he areas located in Section 4.4.2.4.1 shall be made. Measurer.ee ts of the inward deformations and observations of any strain concentrations shall be made. q, 4.4-9

N 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 corrective action. 4.4.2.4.4 The surveillance program shall be discontinued af ter the one year after initial start-up inspection if no corrective action was needed. If corrective action is required, the frequency of inspection for a continued surveillance program shall be determined. Bases Provisions have been made for an in-service surveillance program, covering the first five years of the life of the unit, intended to provide sufficient evi-dence to maintain confidence that the integrity of the reactor building is being preserved. This program consists of tendon, tendon anchorage and liner plate surveillance. To accomplish these programc, the following representative tendon groups have been selected for surveillance: Horizontal - Three 120* tendons comprising one complete hoop system below grade. Vertical - Three tendons spaced approximately 120* apart. Dome - Three tendons spaced approximately 120* apart. ) The inspection during this initial five year period of at least one wire from each of the nine surveillance tendons ( one wire per group per inspection) is considered sufficient representation to detect the presence of any wide spread tendon corrosion or pitting conditions in the structure. This program will be subject to review and revision as warranted based on studies and on results obtained for this and other prestressed concrete reactor buildings during this period of time. REFERENCE FSAR, Section 5.6.2.2 i i 4.4-10 1 ~

4.4.3 Hydrogen Purge System Aoplicability Applies to testing Reactor Building Hydrogen Purge System. Obiective To verify that this systen and components are operable. Specification 4.4.3.1 Operating Tests 1 An in-place system test shall be performed during each refueling These tests shall period using the written emergency procedures. consist of visual inspection, hook-up of system, a flow measurement using flow instruments in the portable purging station and pressure Flow shall be design drop measurements across the filter bank. flow or higher, and pressure drops across the filter bank shall not Fan motors shall be exceed two ti=es the pressure drop when new. operated continuously for at least one hour, and valves shall be This test shall demcastrate that under simulated proven operable. emergency conditions the system can be taken from storage and placed into operation within 48 hours. 4.4.3.2 Filter Tests During each refueling period, leakage tests using DOP on HEPA units and Freon-112 (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-ll2 (or equivalent) by each entire charcoal absorber unit shall constitute acceptable These tests must also be performed after any main-performance. tenance which may affect the structural integrity of either the filtration system units or of the housing. 4.4.3.3 H2 Detector Test Hydrogen concentration instruments shall be calibrated each refueling period with proper consideration to moisture effect. Bases The purge system is composed of a portable purging station and a portion of the 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 A suction =ay ventilation system penetrations and discharged to the unit vent. be taken on the Reactor Building via isolation valve PR-7 (Figure 6-5 of the FSAR) using the existing vent and pressurization connections. 4.4-11

i 'N The purge rate is controlled through the use of a portable purging station / (Insert, Figure 14A-5.1 of the FSAR). The station consists of a purge blower, dehumidifier, filter train, purge flow =eter, sample connection and flowmeter and associated piping and valves. The blower is a rotary positive type rated 60 scfm. The dehumidifier consists of two redundant heating elements inserted in a section of ventila* ton duct. The function of the dehumidifier is to sufficiently increase the temperature of the entering air to assure 70 percent relative humidity entering the filter train with 100 percent saturated air entering the dehumidifier. The purpose of the dehumidifier is to assure optimum charcoal filter efficiency. Heating element control is provided by a thermoswitch. Humidity indication is The provided downstream cf the heating elements by a humidity readout gage. filter train provides prefiltration, high efficiency particulate filtration and charcoal filtration. The filter train assembly is identical in design to the waste gas filter train assembly which is rated at 200 scfm, thus con-servatively capable of performing the assigned function. Face velocity to the charcoal filter is very low. The charcoal filter is cocposed 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 is metered using a 0-12 scfm rotometer. Both of these rotometers have an accuracy of i two percent of full scale, and each has remote readout capabil. ity. The purge discharge rate is controlled by a blower discharge throttling valve. The purge sample activities can be collected, counted and analyzed in the radio-chemistry laboratory. Makeup air to the Reactor Building is supplied by a compressed air system connection to one of the aforementioned existing vent and pressurization connections. That portion of the penetration room ventilation system piping and valves which is used as a part of the purge system is permanently installed and is 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, there is adequate time before purging is required to permit cheffo2t of the portable purging station and to optimize the system operation to mini-mize the total dose to the public. REFERENCES FSAR, Section 14A l ( 1 J i 4.4-12 -~ ~

4.5 EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emer[encyCoreCooling Systems Applicability Applies to periodic testing requirement for emergency core cooling systems. Obiective To verify that the emergency core cooling systems are operable. Soecification 4.5.1.1 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 oressure 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 Inj ection System (a) During each refueling period, a system test shall be conducted to demonstrate that the system is operable. The test snall be performed in accordance with the procedure summarized below: (1) A test signal will be applied to demonstrate actuation of the Irv pressure injection system for emergency core cooling operation. (2) Verification of the engineered safety features function of the low pressure service water system which supplies cooling water to the low pressure coolers shall be made to demonstrate operability of the coolers. (b) The test will be considered satisfactory if control boarl indication verifies that all components have responded to the actuation signal properly; all appropriate pump breakers shall have opened or closed, ata all valves shall have completed their travel. 4.5-1

i l 4.5.1.1.3 Core Flooding System s ] (a) During each refueling period, a system test shall be j conducted 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. 4.5.1.2 Component Tests 4.5.1.2.1 Pumps At intervals not to exceed 3 months, the high pressure and 4 low pressure injection pumps shall be started and operated to verify proper operation. Acceptable performance will be indicated if the pump starts, operates for fifteen minutes, 3 and the discharge pressure and flow are within 110% of a point on the pump head curve. 4.5.1.2.2 Valves - Power Operated (a) At intervals not to exceed three months each engineered ') safety features valve in the emergency core cooling systems and each engineered safety features valve associated with emergency core cooling in the low pres-sure service water system shall be tested to verify operability. (b) The acceptable performance of each power operated valve will be that motion is indicated upon actuation by appropriate signals. t Bases The emergency 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. The 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. The requirements of the low pressure service water system for cooling water are more severe during normal operation than under accident conditions. Rotation of the pump in operation on a monthly basis will verify that two pumps are operable. l .s/ 4.5-2

for operability by The low pressure injection pumps are. tested singularlynd the bypass valves l opening the borated water storage tank outlet va ves aThis allows water to be in the barated water storage tank fill line. h of the injection pumped from the borated water storage tank through eac lines and back to the tenk. ding line are checked 'Jith the reactor shutdown, the valves in es.h core floo til the for operability by reducing the reactor coolant system pressure un k and isolation indicated level in the core flood tanks verify that the chec valves have opened. REFERENCE FSAR, Section 6 4.5-3

Reactor Building Cooling Svstems 4.5.2 Applicability Applies to testing of the reactor building cooling systems. Obiective_ To verify that the reactor building cooling systems are operable. Soecification 4.5.2.1 System Tests 4.5.2.1.1 Reactor Building Spray System During each refueling period a system test shall be (a) conducted to demonstrate proper operation of the signal will be applied to demonstrate A test system. actuation of the reactor building spray system (except for reactor building inlet valves to prevent water entering noz:los). Water will be circulated from the borated water storage tank through the reactor building line to the spray pumps and returned through the test borated water storage tank. Station compressed air will be introduced into the spray (b) headers to verify the availability of the headers and spray nozzles at least every five years. The test will be considered satisfactory if visual (c) 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 During each refueling period, a system test shall be (a) conducted to demonstrate proper operation of the system. The test shall be performed in accordance with the procedure summarized below: A test signal will be applied to actuate the reactor (1) building cooling system for reactor building cooling operation. Verification of the engineered safety features function (2) of the low pressure service water system which supplies to the reactor building coolers shall be made coolant to demonstrate operability of the coolers. 4.5-4

(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 have completed their travel, fans are running at half speed, LPSW flow through each cooler exceeds 1400 GPM and cir flow through each fan exceeds 40,000 CFM. 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 15 minutes, and the discharge pressure and flow are within 110% 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 features valve associated with reacter building coeling in the low pressure service water system shall be tested to verify that it is operable. Bases 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 pressure and flow indication demonstrate per-formance. 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 ficw 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 units 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 out-side the reactor building are inspectable at all times. Operational tests and ) inspections will be performed prior to initial startup. ../ 4.5-5 (

_ _ = _ At least One low pressure service water pump will normally be operating. once per month, the operation will be rotated to another low pressure Testing will be, therefore, unnecessary. service water pump. The reactor bu!lding fans are normally operated periodically, constituting the test that these fans are operable. REFERENCE FSAR, Section 6 4.5-6

Each filter train is constructed with a prefilter, an absolute filter, and ) ) a charcoal filter in series. The design flow rate through each of these filters is 1000 scfm, which is significantly higher than the approximately 15 scfm maximum leakage rate from 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 deteriorntion, and unusual or excessive noise or vibration when the fan motor is running. 4 Less fregt cnt testing will verify the efficiency of the absolute and char-coal filters. 1 4 )) 4 4 i a 1 4.5-8

4.5.3 Penetration Room Ventilction System Applicability Applies to testing of the reactor building penetration room ventilation system. Obiective To verify that the penetration room ventilation system is operable. Soecification 4.5.3.1 System 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 hour, and the louvers and other mechanical system shall be proven operable and adjustable from their remote location. 4.5.3.1.2 The test will be considered satisfactory if control board indication verifies that all components have responded properly to the actuation signal, if flow rate through the system is design flow or higher, and if pressure drops across any filter bank do not exceed two times the pressure drop which existed when the filters were new. 4.5.3.2 Filter Tests No less frequently than each normal refue'ing period, "in-place" leakage tests using DOP and HEPA units and Freon-ll2 (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 absorber 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 penetration room ventilation system is designed to collect and process potential reactor building penetration room leakage to minimize environ-nental activity levels resulting from post accident recator building leaks. The system consists of a sealed penetration room, two redundant filter trains, and two redundant fans discharging 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. 4.5-7

e g Low Pressure Iniection Svstem Leakage 4.5.4 Applicability _ Applies to Low Pressure Injection System leakage. Obiective To maintain a preventative leakage rate for the Low Pressure Injection System which will prevent significant offsite exposures. Suecification_ 4.5.4.1 Acceptance Limit 1' The maximum allowable leakage from the Low Pressure Injection System ll components (which includes valve stems, flanges and pump seals) sha not exceed two gallons per hour. 4.5.4.2 Test During each refueling period the following tests of the Low Pressure J l Injection System shall be conducted to determine leakage: The portion of the Low Pressure Injection System, except as specified in (b), that is outside the containment shall be a. lly tested either by use in normal operation or by hydrestatica testing at 350 psig, Piping from the containment emergency sump to the b. at no less than 59 psig. Visual inspection shall be made for excessive leakage from com c. ponents of the system.by collection and weighing or by another equivale Bases _ for the Low Pressure Injection System is a judgmen, t ithout The leakage rate limit value based on assuring that the components can be expected to operate w Loss of mechanical failure for a period on the order of 200 days after aThe te i ver system operation or by hydrostatically testing, gives an adequate marg n o Coolant Accident. ident. the highest pressure within the system after a design basis accfor.the return Similarly, the pressure test to the design pressure Low Pressure Injection System (59 psig) is equivalentThe dose to the thyroid leakage is.76 rem for a 2 hr. exposure at the site boundary. (1) of the containment. REFERENCES FSAR, Section 14.2.2.4.4. 4.5-9

o s t a t i b. The specific gravity and voltage of each cell shall be S measured and recorded every month. ) Before initial operation and at each refueling outage, a one-c. hour discharge test at the required maximum safeguards load will be made. 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. 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. 4.6.9 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 components have operated properly. Bases The Keowee Hydro units, in addition to serving as the emergency power sources for the Cronee 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 from the Oconee Unit 1 Control Room. The frequent starting and loading of these units to meet Duke system 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, after a simulated requirement for engineered j safety features. To further assure the reliability of these units as emergency power sources, they will be, as specified, tested for automatic start on a monthly basis from the Oconee control room. These tests will include verification that each unit can be synchronized 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 i FSAR, Section 8 \\ 4.6-2

EMERGENCY POWER SYSTEM PERIODIC TESTING 4.6 Applicability Applies to the periodic testing and surveillance of the emergency power system. Obiective To verify that the emergency power scurces and equipment will respond promptly and properly when required. Specification intervals not to exceed one month, a test of the Keewee Hydro At 4.6.1 units shall be perforced to verify proper operation of these emer-This test shall gency power sources and associated equipment. assure that: Each hydro unit can be automatically started from the Oconee a. Control Room. Each hydro unit can be synchronized through the 230 kV over-b. head circuit to the startup transformers. Each hydro unit can energize the 13.6 kV underground feeder. c. During each calendar year at a refueling outage, the Keowee Hydro ch 4.6.2 Units will be started using the emergency start circuits in 6 control room to verify that each hydro unit and associated equip-ment is available to carry load within 25 seconds of a simulated requirement for engineered safety features. Annually a simulated emergency transfer to the 4160 volt main 4.6.3 feeder buses shall be made to transformers CT1, CT2, and CT3, and ) to the 4160 volt standby buses to verify proper operation. Quarterly the External Grid Trouble Protection System Logic shall 4.6.4 be tested to demonstrate its ability to provide an isolated power pith batween Keowee and Ocanee. Annually it shall be demonstrated that a Lee Station combustion i It shall 4.6.5 turbine can be started and connected to the 100 kV line. be demonstrated that the 100 kV line can be separated from the rest of the system and supply power to the 4160 volt main feeder buses. 3atteries in the Instrument and Control, Keowee Station, and 4.6.6 Switching Station 125 volt DC systems shall be tested as follows: The voltage and temperature of a pilot cell in each bank a. shall be measured and recorded daily, five days / week. 4.6-1

deviates from its group average position by core than nine (9) inches. Con-ditions for operation with an inoperable rod are specified in Technical Specification 3.5.2. (2) REFERENCES (1) FSAR, Section 14 (2) Technical Specification 3.5.2 4.7-2

t I 4.7 REACTOR CONTROL R0D SYSTEM TESTS 4.7.1 Control Rod Drive Svstem Functional Tests Applicability Applies to the surveillance of the control rod system. Obiective To assure operability of the control rod system. Specification 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions followir.g 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 Rods (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 1.40 seconds for no flow conditions. For the APSRs it shall be demonstrated that loss of power will not cause rod movement. If the trip insertion time above is not met, the rod shall be declared inoperable. 4.7.1.2 If a control rod is misaligned with its group average position by more than an indicated nine (9) inches, the rod shall be declared inoperable. However, if a safety rod absolui:e or relative position indication is inoperable and an energized out-limit light confirms the rod is fully withdrawn, the ' safety rod shall not be considered misaligned. 2 4.7.1.3 If a control rod cannot be exercised, or if neither absolute or relative position indication is operable, the rod shall be declared to be inoperable. Bases The control rod trip insertion time is the total elapsed time from power interruption at the control red drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The specified trip time is based upon the safety analysis in FSAR, Section 14. Each 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 red at reactor operating conditions. Exercising the drive mechanisms in this manner provides assuranca i of reliability of the mechanisms, i A rod is considered inoperable if it cannot be exercised, if the trip in-sertion time is greater than the specified allowable time, or if the rod 4.7-1 L

Control Rod Program Verification (Group vs. Core Positions) 4.7.2 Applicability Applies to surveillance of the control rod systems. Obiective the designated control rod (by core position 1 through 69) To verify that (Rod 1 is operating in its programmed functional position and group. through 12, Group 1-8) Soecification Whenever the control red drive patch panel is locked (af ter in-reprogramming, or maintenance) each control rod 4.7.2.1 spection, test, drive mechanism shall be selected from the control room and exercised by a =ovement of approximately two inches to verify that the proper rod has responded as shown on the unit computer print-out of that rod. l Whenever power er instru=entation cables to the control rod drive 4.7.2.2 assemblies atop the reactor or at the bulkhead are disconnected or renoved, an independent verification check of their reconnection shall be perfor=ed. Any rod found to be improperly progra=med shall be declared in-4.7.2.3 operable until properly progra=med. Bases Each control rod has a relative and an absolute position indicator system. One set of outputs goes to the plant computer identified by a unique number The other set of (1 through 69) associated with only one core position. outputs goes to a programmable bank of 69 edgewise meters in the control In the event that a patching error is made in the patch panel or connectors in the cables leading to the control red drive assemblies or room. to the control room meter bank are improperly transposed upon reconnection, these errors and transpositions will be discovered by a comparative check by (1) selecting a specific rod from one group (e.g. Rod 1 in Regulating (2) noting that the program-approved core position for this rod Group 6)

53) (3) exercise of the group (assume the approved core position is No.

the selected rod and (4) note that (a) the computer prints out both absolute and relative position response for the approved core position (assumed to be position No. 53) (b) the proper meter in the control room display bank (assumed to be Rod 1 in Group 6) in both absolute and relative i This type of comparative check will not assure detection meter positions. For these, of improperly connected cables inside the reactor building. (Spec. 4.7.2.2) it will be necessary for a responsible person, other than the one doing the work, to verify by appropriate means that each cable has been matched to the proper control red drive assembly. 4.7-3

4.8 MAIN STEAM STOP VALVES Aeplicability Applies to the main steam stop valves. Obiective To verify the ability of the main steam stop valves to close upon signal and to verify the leak tightness of the main steam stop valves. Soecification 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. 4.8.2 The leak 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. Bases The main steam stop valves limit the reactor coolant system cooldown rate and resultant reactivity insertion following a main steam line break acci-dent. Their 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 B 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 gene-rator 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 steam generators) shows that the reactor will remain subcritical after reactor trip following a double-ended steam line break. i The main stop valves would become isolation valves in the unlikely event 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 leakage from all penetrations and isolation valves.(2) REFERENCES (1) FSAR, Supplement 2, Page 2-7 (2) Technical Specifications 4.4.1 4.8-1

1 4.9 EMERGENCY FEEDWATER PUMP PERIODIC TESTING Applicability 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-conth 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 indication and visual observation of the equipment demonstrates that all components have operated properly. Bases 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 FSAR, Section 10.2.2 FSAR, Section 14.1.2.8.3 4.9-1 )

F 4.10 REACTIVITY ANOMALIES Apolicabilitv Applies to potential reactivity ancmalies. Obiectiva To require the evaluation of reactivity anomalies of a specified magnitude occurring during the operntion of the unit. Saecification 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 per-cent in reactivity, an evaluation as to the cause of discrepancy shall be made and reported to the Atomic Energy Commission. Bases Io eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup tnd the boron concentration, necessary to naintain adequate control charucteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted con-centration and the slope of the cur.a relating burnup and reactivity is ccmpared with that predicted. This process of normalization should be com-pleted after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1% wculd be unexpected, and its occurrence would be thoroughly investigated and evaluated. The value of 1% is considered a safe limit since a shutdown margin of at least 1% with the most reactive rod in the fully withdrawn position is always maintained. 1 4.10-1 L 1

TABLE 4.11-1 ENVIRONMENTAL SURVEILLANCE PROGRAM Type Samples Schedule Analysis Water (1) (3) Monthly Gross Alpha & Beta Activity Gamma Analysis *(4) Rain, Settled Dust Monthly Gamma Analysis (4) Air Particulate

Monthly, Gamma Analysis (4) 131 Charcoal Filter Weekly Gamma Analysis for 1 Vegetation, Terrestrial (2)

Quarterly Gamma Analysis Quarterly Aquatic Organisms Quarterly Gamma Analysis Quarterly f Bottom Sediment Quarterly Gamma Analysis Quarterly 5 l3 Radiation Dose & Dose Rate Quarterly mr & mr/hr Sr90 Csl37 Animals Quarterly Fish Quarterly Sr90 Cs137, g40, 7131 Milk Weekly 1131 Monthly Sr90, Cs137, g40

• Dependent on gross activity for individual samples.

(1) Lakes Keowee and llartwell will be sampled annually for tritium analysis. (2) Commercial Corps will be substituted when available. (3) The water samples from the Keowce River continuous sampler will be analyzed monthly for gross Alpha and Beta and undergo un isotopic analysis. (4) A gamma analysis on a quarterly composite-sample. gr 9

4.11 ENVIRONMENTAL SURVEILLANCE 1 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 j evaluating any significant effects of station operation and waste releases on the environment. ] Specification 1 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 of these specifications. l 4.11.2 Thermoluminescent dosimeters will be installed at various loca-i tions within the exclusion area including shoreline areas, and read quarterly. 4.11.3 The location of cows and/or goats used as a direct source of milk for individuals, and the location of any dairies within five miles of the plant will be determined annually by survey and reported with the results of the radiological environmental monitoring program. 1' Bases The program will be conducted in accordance with Section 2.7 of the FSAR. The sensitivity of counting used in the analyses (90% ccnfidence level when counted for 20 minutes) is based on nominal background levels of 0.5 cpm alpha and 1.0 cpm beta. Eavironmental monitoring results will be correlated with information on station radioactive vaste releases, site meteorological data and radiological controls, and with information obtained from the ine alled process radiation monitoring system. The Environmental Surveillance Program will provide means of detecting signi-ficant changes in levels of radioactivity. The results will damonstrr+ e the effectiveness of station control over radioactive waste disposal opera ; ions and of compliance with Federal and State regulations for disposal of these materials. The thermoluminescent desimeter 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 and as low as practicable for unrestricted areas. REFERENCES FSAR, Section 2.7 i 4.11-1 l l -=

1 CONTROL ROOM FILTERING SYSTDi 4.12 Applicability Applies to control room filtering system components. Obiective To verify that these systems and components will be able to perform their design functions. Specification 4.12.1 operatine Tests -{ System tests shall be performed at approximately quarterly inter-These tests shall consist of visual inspection, a flow installed at the outlet of vols. measruement using a flow instrument each unit and pressure drop measurements across each filter bank. Pressure drop across pre-filter shall not exceed 1" H O and pressure 2 Fan motors shall be drop across HEPA shall not exceed 2" H 0.one hour, and all louvers and 2 I least operstod continuously for at otner mechanical systems shall be proven operable. 4.12.2 Filter Tests Annually for the Unit 1 and 2 and the Unit 3 control room an "in-place" leakage test using DOP on HEPA units and Freon-ll2 (or equivalent) on charcoal units shall be performed at design flow on I Removal of 99.5% DOP by each entire HEPA filter each filter train. unit and removal of 99.0% Freon-112 (or equivalent) by each entire charcoal adsorber unit shall constitute acceptable performance. These tests must also be performed after any maintenance which may the structural integrity of either the filtration system affect l units or of the housing. Bases The purpose of the control room filtering system is to limit the particulate and gaseous fission products to which the control area would be subjected l during an accidental radioactive rf ease in or near the Auxiliary Building. percent capacity filter trains each of The system is designed with two J-which consists of a prefilter, hiah efficiency l 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 Quarterly testing of this system will insure their operability when needed. During this test the the system is available for its safety action.for such things as water, oil, show that system will be inspected. d material; gasket deterioration, adhesive deterioration in the HEPA units; an unusual or excessive noise or vibration when the f an motor is running. Annual testing will verify the efficiency of the charcoal and absolute filters. I I I f 4.12-1

] l l 4.13 FUEL SURVEILLANCE Apolicability Applies to the fuel surveillance program for fuel rods of Unit 1. l J i Obiective 1 I 5 ] To specify the fuel surveillance program for fuel rods. Specification 4.13.1 Visual Inspection Two (2) Oconee Unit 1 fuel assemblies will be designated for visual l; inspection. These same assemblies will be inspected during each of the first three refuelings of Unit 1. Underwater viewing devices will be used to determine that the fuel rods have maintained their structural integrity. l 4.13.2 Dimensional Examination Measurements of the length and outside diameter will be made on selected peripheral rods of the following fuel assemblies of the first core of Unit 1 both prior to operation and at the times l specified: a. One assembly after the first cycle. 1 b. Four assemblies after the second cycle. c. Two assemblies after the third cycle. 7 'l Bases l This fuel surveillance program provides substantiating information for the first core in the present generation of BhW reactors. It provides for examination of fuel rods at the end of the first, second, and third cycles j of Unit 1 to determine if fuel rods have maintained their integrity and to determine the exten:, if any, of dimensional changes in diameter and length. 4 4.13-1 A

4.14 REACTOR BUILDING P'7GE SYSTDI Applicability Applies to testing Reactor Building Purge Filters. Obiective To verify that the Reactor Building Purge Filters will perform their design function. Specification Annually, leakage tests using DOP on the EEPA filter and Freon-ll2 (or equivalent) on the charcoal unit shall be performed. Removal of 99.5% DOP by the HEPA filter unit and removal of 99.0% Freon-ll2 (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 prefilter, an acsolute filter and a charcoal filter in series. This test will verify the efficiency of the absolute and charcoal filters. 4.14-1 r i

I i IODINE RADIATION MONITORING FILTERS i 4.15 Applicability Applies to the Iodine Radiation Monitoring Filters. I Obiective_ j i To assure that the Iodine Radiation Monitoring Filters perform the r i intended function. f Soecification The Iodine Radiation Monitoring Charcoal Filters will l 44. f 4.15.1 l All spare Iodine Radiation Monitoring Charcoal Filters will be t I 4.15.12 discarded after 2 years of shelf-life. i Bases _ } dine The purpose of this specification is to assure the reliability of the IoThis s t Radiation Monitoring Charcoal Filters. compliance to a request by the AEC. l 1 t l i I t i 4 { i j l l i I l 4.15-1 l.

l 1 5 DESIGN FEATURES 1 5.1 SITE 4 4 j 5.1.1 The Oconee Nuclear Station is approximate 1: ight miles northeast of Seneca, South Carolina. Figure 2-3 of tau Oconee FSAR shows the plan of the site. The minimum distance from the reactor center line to the boundary of the exclusion area and to the outer boundary of the low population zone as defined in 10 CFR 100.3, l 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 release purposes only, is the same as the exclusion area as defined above except that the temporary con-struction quarters (2) in the east southeast section of the exclusion area shall not, when occupied, be deemed to be within j the restricted area. REFERENCE (1) FSAR, Section 2.2 (2) Technical Specification 3.9 i a 1 J l e t l 5.1-1 l .t

~S The princi . design basis for the structure is that it be capable .) l 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 into the reactor building through a break in the reactor coolant piping. Subsequent pressure behavior is datermined 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 component can result in loss-of-isolation or intolerable leakage. The installed double barriers take the form or closed piping systems, both inside and outside the reacter building and various types of isolation valves. (2) s.2.3 Penetration Room Ventilation System This system is designed to collect, control, and minimize the release of radioactive naterials from the reactor building to the environment in post-accident conditions. It may also operate intermittently during nor=al conditions as required to maintain ') satisfactory temperature in the penetrations rooms. WN a the system is in operation, a slight negative pressure will be main-tained in the penetration room to assure inleakage. (3) REFERENCES (1) FSAk cetion 5.1 (2) FSAR Section 5.2 (3) FSAR Section 5.3 \\ l 5.2-2

i .i ) 5.2 CONTAUCiENT Specification The containment for this unit censists of three systems which are the reactor building, reactor building isolation system, and penetration room ventilation j system. 5.2.1 Reactor Building i The reactor building completely encloses the reactor and its asssalated reactor coolant system. It is a fully continuous re-inforced concrete structure in the shape of a cylinder with a 3 shallow domed roof and flat foundation slab. The cylindrical j pcrtion 1a 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 loss of 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 reinforced with high strength reinforcing steel. The entire structure is lined with 1/4" welded steel plate to provide vapor tightness. The internal volume of the reacter building is approximately 1.91 x 106 cu. ft. The approximate inside dimensions are: diameter-- i 116'; height--208 1/2'. The approximate thickness of the concrete j 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. i The reactor building is designed for an external atmospheric pressure of 3.0 psi greater than the internal pressure. This is greater than the differential pressure of 2.5 psig 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 cooled to an internal temperature of .i i 80*F with a concurrent rise in barometric pressure to 31.0 inches of Hg. Since the building is designed for this pressure differ-ential, vacuum breakers are not required. Penetration assemblies are seal welded to the reactor building liner. Access openings, electrical penetrations, and fuel trans-fer 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) 5.2-1 i b

5.3 REACTOR S'pecification I 5.3.1 Reactor Core 5.3.1.1 The reactor core contains approximately 93 metric tons of slightly enrich,a uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 177 fuel assemblies, all of which l are prepressurized with Helium. 1 5.3.1.2 The fuel assemblies shall form an essentially cylindrical lattice with an active height of 144 in. and an equivalent diameter of 128.9 in. (2) 5.3.1.3 There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSR) distributed in the reactor core as shown in FSAR Figure 3-46. The full-length CRA contain a 134 inch length of silver-indium-cadmium alloy clad with stainless steel. The APSR contain a 36 inch length of silver-indium-cadnium alloy. (3) 5.3.1.4 Initial core and reload fuel assemblies and rods shall conform to design and evaluation described in FSAR and shall not exceed an enrichment of 3.5 percent of U-235. 5.3.2 Reactor Coolant System 5.3.2.1 The design of the pressure components in the reactor coolant system shall be in accordance with the code requirements.(4) 5.3.2.2 The reactor coolant syste= 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*F. The pressurizer and pressurizer surge line shall be designed for a temperature of 670*F.(5) 5.3.2.3 The maximum reactor coolant system volume shall be 12,200 ft3, REFERENCES (1) FSAR Section 3.2.1 (2) FSAR Section 3.2.2 (3) FSAR Section 3.2.4 (4) PSAR.Section 4.1.3 (5) FSAR Section 4.1.2 5.3-1

5.4 NEW AND SPENT FUEL STORAGE FACILITIES Specification 5.4.1 New Fuel Storage 5.4.1.1 New fuel will nor= ally be stored in the spent fuel pool serving the respective unit. The fuel assemblies are stored in racks in parallel rows, having a ncminal center to center 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 fuel assemblies are stored in five racks in a row having a nominal center to center distance of 2' 1 3/4". One rack is oversized to receive a failed fuel assembly container. The other four racks are normal size and are capable of receiving new fuel assemblies. 5.4.1.3 New fuel may also be stored in shipping containers. 5.4.2 Spent Fuel Stors e 5.4.2.1 Irradiated fuel assemblies will be stored, prior to off site shipment, in the stainless steel lined spent fuel pool, which is located in its respective auxiliary building. Each pool is sized to acec=modate 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 r_anagement program. 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. 5.4.2.3 Spent fuel may also be stored in storage racks in the fuel transfer canal when the canal is at refueling level. 5.4.2.4 The spent fuel pool and fuel transfer canal racks are designed for an earthquake force of 0.lg ground motion. REFERENCES FSAR, Section 9.7 5.4-1

b. At least two licensed reactor operators shall be at the station, _s ) one of whom shall be in the control room, at all times when there is fuel in the reactor vessel. One of these operators shall hold a Senior Reactor Operator license. Two licensed reactor operators shall be in the control room during c. startup and scheduled shutdown of the reactor, and during recovery from reactor trips. d. At least one licensed reactor operator shall be in the reactor building when fuel handling operations are in ; ogress in the reactor building. An operator holding a Sen;ur Reactor Operators license and assigned no other concurrent operational duties shall be in direct charge of refueling operations. e. At least one operator per shift will have sufficient training to perform routine health physics requirements. J 6.1.2 Review and Audit In matters of nuclear safety and radiation exposure, review and audit of station operation, maintenance and technical matters shall be provided by two co=mittees as follows (Reference Figure 6.1.2): d 6.1.2.1 Station Review Co=mittee a. Membership Assistant Superintendent--Chairman 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 each month and as required on call by the chairman. c. Quorum The chairman plus two members shall constitute a quorum, d. Responsibilities The committee shall have the following responsibilities: (1) Review all new procedures or proposed changes to existing procedures as determined by the Station Superintendent to 3 affect operational safety. / 6.1-2 ,-m---

6 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION, REVIEW AND AUDIT Introduction Administrative controls relate to the organization and management procedures, 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 limits or safety limits are exceeded. Specification These administrative controls in regard to operations, review and audit become effective at the time of issuance of the facility license by the AEC, and after normal design and construction activities have been essentially completed. 6.1.1 Organization 6.1.1.1 The Superintendent is directly responsible for the safe operation of the facility. 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 Incorporated in the staff of the station shall be supervisory and' professional personnel coeting the minimum requirements specified in FSAR Section 12.2 ene rpassing the training and experience des-cribed in Section 4 of t.te ANSI 18.1, " Selection and Training of Nuclear Power Plant Personnel." 6.1.1.5 Retraining and replacement of station personnel shall be in accord-ance with Section 5.5 of the ANSI 18.1, " Selection and Training of Nuclear Power Plant Personnel." 6.1.1.6 AEC/ DOL licensed operators shall be provided as follows: a. Unit 1 minimum shift staffing for other than cold shutdown will be according to Table 6.1-1. This minimum operating staff shall be permitted to assist in the pre-licensing activity of Units 2 and 3 only to the extent that it does not affect their full availability for Unit 1 operation. 6.1-1

j l Provisions for assuring that the committee is kept informed of ~} f matters within its purview. c. The committee shall be composed of: Chairman At least two members from the Steam Production Department. (May include Superintendent or Assistant Superintendent but not other Oconee Nuclear Station personnel) At least two members from the Engineering Department. Others deemed advisable. (May include consultant from outside the company) The committee shall elect a vice chairman. i Qualified alternates shall be appointed or other provisions shall be made for covering the absence of fulltime members of the group. The use of alternates shall be restricted to legitimate and unavoidable absences of principals. d. Qualifications: At least one-half of the members of the committee (and/or ] alternates attending a specific meeting) shall have extensive _,/ nuclear experience and all members and alternates shall be engineering or science graduates. No more than a minority of the members or alternates shall have direct line responsibility for station operation. All members shall have a minimum of three years' professional level experience in nuclear services, nuclear plant operation, or nuclear engineering and the neces-sary overall nuclear background to detect when to call con-sultants and contractors for dealing with complex problems beyond the scope of their own organization. e. Members of the committee shall collectively have the capability required to review the areas of: (1) Nuclear power plant operations (2) Nuclear Engineering 4 (3) Chemistry and Radiochemistry (4) Metallurgy (5) Instrumentation and Control (6) Radiological Safety (7) Mechanical and Electrical Engineering i (8) Other appropriate fields associated with the unique characteristics of the Oconee Nuclear Station. s 6.1-4

r j l l l (2) Review station operation and safety considerations. l (3) Review abnormal occurrences and violations of Technical Specifications and make recommendations to prevent recurrence. l (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 Committee shall make recommendations to the Superintendent regarding Specification 6.1.2.1-d. f. Records Minutes shall be kept at the Station of all r stings of the Committee and copies sent to the Superintencent, Assistant Vice President Steam Production, and the chairman of the Nuclear Safety Review Committee. 6.1.2.2 Nuclear Safety Review Cecmittee a. The Executive Vice President and General Manager shall appoint a Nuclear Safety Review Ccamittee having responsibility for verify-ing that operation of the station is consistent with company policy and rules, approved operating procedures, and license pro-visions; to review important proposed plant changes, and tests; to verify that abnormal occurrences and unusual events are promptly investigated and corrected in a manner which reduces the probability of recurrence of such events; and to detect trends which may not be apparent to a day-to-day observer. b. The activities of the Nuclear Safety Review Co=mittee shall be guided by a written charter that contains the following: Subjects within the purview of the committee. Responsibility and authority. Mechanisms for convening meetings. Provisions for use of sub-groups. Authority for access to station records. Reporting requirements. Identification of management position to which the group reports. 6.1-3 i

(6) Abnormal occurrences as defined in 1.8 of these Specifi- ) cations. (7) Station operating records, logs, reports, and tests on a periodic basis. (8) Station Review Committee Minutes (9) Conduct special reviews or investigations as required by the Assistant Vice President, Steam Production or the Station Superintendent. (10) Non-Routine Reports to the AEC and AEC Responses The commictee shall make recommendations relating to the review of the above items to the appropriate members of management to prevent or reduce the probability of re-currence. Copies of these recommendations shall be included in the meeting minutes. (11) Activities of the Steam Production Qaality Assurance Organization. j. It is the intent of this specification to fulfill the reauirements and reco=mendations of Proposed Standard ANS-3.2, Section 4, entitled, " Standard for Administrative Controls for Nuclear Power ]' Plants," - Draft 7, March 29, 1972. ../ : I f 6.1-6

When the nature of a particular situation dictates, special consultants shall be utilized to provide expert ' advice to the committee. f. Meeting Frequency: The committee shall meet at least three times per year at inter-vals not to exceed five months and as required on call by the chairman. During the period of initial operation, this committee shall meet at least once per calendar quarter. g. Quorum The chairman or vice-chairman plus three members, or appointed alternatives, shall constitute a quorum. No more than a minority I of the quorum shall have direct line responsibility for station operation. f h. Meeting Minutes: Minutes of all scheduled meetings of the committee shall be pre-pared and shall identify all documentary materials reviewed. These minutes shall be formally approved, retained, and also promptly distributed to the Executive Vice President and General Manager; Senior Vice President, Engineering and Construction; i Senior Vice President, Production and Transmission; Vice President, Design Engineering; Assistant Vice President, Steam Production; and Station Superintendent. A copy of these minutes shall be kept on file at the station. 1. As a safety review and audit backup to the normal operating organization, the committee shall review the following: (1) Proposed tests and experiments, and results thereof, when these constitute an unresolved safety question defined in 10CFR50.59. (2) Proposed changes in equipment or systems which may con-stitute an unresolved safety question defined in 10CFR50.59, or which are referred by the operating organization. (3) All requests to the AEC/ DOL for changes in Technical Specifications or license that involve unresolved safety questions as defined in 10CFR50.59. (4) Violations of Statutes, Regulations, Orders, Technical Specifications, License Requirements, or Internal Proce-dures, or Instructions having Safety Significance. l (5) Significant operating abnormalities or deviations from j normal performance of unit equipment. i a 6.1-5

t Executive Vice President (____________________, and General Manager g i l i i a i i (_______ _q Senior Vice President Senior Vice President Production and Transmission l Engineering and Construction A i i I e i I Assistant Vice President 8 Vice President l -y (_________1_ _____q g Steam Production Design Engineering g g i A I i r----- l----- Superintendent 3 3 3 (_______q L_-{ Nuclear Safety Review Committet-__J Oconee Nuclear Station t I g L_______ ______J I j c____L____, I I I L ___ _ ___J Station Review Committee i I I l______ \\ OCONEE NUCLEAR STATION l MANAGEMENT ORGANIZATION CHART FIGURE 6.1-2 'Y

Superintendent isitors Center Clerical Staff Asst. Supt. Staff Operating SRO Technical Support Maintenance Engineer Engineer Supervisor Asst. Maintenance Asst. Operating m Supervisor Engineer SR0_ Performance Instrument Chemistry-Health Engineer Supervisor Physics Supv. Maintenance Shift SRO Personnel Supervisor Instrument Chemistry-liealth Personnel Physics Personnel Control RO SRO - Senior Reactor Operator License Operator RO - Reactor Operator License Asst. Control Operator RO OCONEE NUCLEAR STATION STATION ORGANIZATION CilART Utility Operator FIGURE 6.1-1

TABLE 6.1-1 OCONEE NUCLEAR STATION MINIMUM CPERATING SHIFT REQUIREMENTS UNT.T 1 Minimum Number / Responsibility Oualifications Shift Shift Supervisor SRO 1 Control Operator R0 1 Asst. Control Operator R0 1 Utility Operator 2, Total / Shift 5 SRO - AEC Senior Reactor Operator License R0 - AEC Reactor Operator License 6.1-9

6.2 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL OCCURRENCE OR UNUSUAL EVENT Specification 6.2.1 Any abnormal occurrence or unusual event shall be investigated promptly by the Superintendent. 6.2.2 The Superintendent shall promptly notify the Assistant Vice President, Steam Preduction, of any abnormal occurrence or unusual event and shall cause the Station Review Cc=mittee to perform a review and prepare a written report which shall describe the circumstances leading up to and resulting from the occurrence and shall reccamend appropriate action to prevent or minimize the probability of a recurrence. 6.2.3 The Station Review Co=mittee report shall be submitted to the Nuclear Safety Review Committee for review and approval of any recommendations. Copies shall also be sent to the Superintendent and the Assistant Vice President, Steam Production. 6.2.4 The Senior Vice President, Production-Transmission, shall report the circumstances of any abnormal occurrence or unusual event to the AEC as specified in Section 6.6, Station Reporting Requirements. l ) i 6.2-1

1 6.3 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED Specification If a safety limit is exceeded: 6.3.1 The reactor shall be shut down i=sediately and maintained in a safe shutdown condition until otherwise authorized by the AEC. 6.3.2 The Superintendent shall make an immediate report to the Assistant Vice President, Steam Production; the Senior Vice President, Production and Transmission; and the Chairman of the Nuclear Safety Review Committee. 6.3.3 The circumstances shall be promptly reported to the AEC by the Senior Vice President, Production-Transmission as indicated in l Section 6.6, Station Reporting Requirements. I 6.3.4 The Superintendent shall direct the Station Review Committee to perform an analysis of the circumstances leading up to and l resulting from the situation together with reco=mendations to prevent a recurrence. The report covering this analysis shall be sent to the Nuclear Safety Review Com=ittee for review and approval. Copies of this report shall also be subuitted to l the Superintendent; Assistant Vica President, Steam Production; the Senior Vice President, Production and Transmission; the Chairman of the Nuclear Safety 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 AEC by the Senior Vice President, Production-Transmission as indicated in Section 6.6, Station Reporting Requirements. l 6.3-1

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. l 6.4.6 If the Superintendent concludes that a proposed major change in the facility or operating procedures, tests, or experiments involving systems that are nuclear safety or radiation exposure related does not involve a change in the Tcchnical Specifications or is not an unreviewed safety question, he may order the change, test, or experiment to be made, but shall enter a description thereof in the operating records of the facility and shall send a copy of the pertinent instructions to the Chairman of the l Nuclear Safety Review Committee. If the Chair =an 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 Committee, 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 experiment 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 Nuclear Safety Review Committee for review and report. If that Committee is of the opinion that the proposed change, test, or experiment does not require approval by the Atomic Energy Commission 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 experiment. If, on the other hand, the Co==ittee is of the opinion that approval of the Atomic Energy Commission is required, the Co=mittee 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, and Senior Vice President, Production and Transmission, who shall thereupon forward the report and request to the Atomic Energy Commission for approval unless, after review, the Vice Presidents either (a) disagree with the opinion of the Committee that approval of the Atomic Energy Commission is required, or (b) decide that the proposed change, test or experiment is not necessary. i 6.4-2 4 1

o 6.4 STATION OPERATING PROCEDURES Spectftcation 6.4.1 The station shall be operated and maintained in accordance with approved procedures. Detailed written procedures with appropriate check-off 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 oparations. c. Actions taken to correct specific and foreseen potential malfunctions of systems or components involving nuclear j safety 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. e. Preventive or corrective maintenance which could affect nuclear safety or radiation exposure to personnel. f. Station survey following an earthquake, j g. Radiation control procedures. h. Operation of radioactive waste management systems.

i. Control of pH in recirculated coolant after loss of coolant j

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 to adjust the pH te a minimum of 7. j. Nuclear safety related periodic test procedures. 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 by the Superintendent for use except as provided in Specification 6.4.3 below. 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. 6.4-1

f 6.5 STATION OPERATING RECORDS Specification 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 =anner convenient for review. Records of normal station operation including power levels and periods a. 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. Reports of abnormal occurrences and safety limits exceeded. c. 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 correct-ive action taken shall be recorded.

• e.

Records of changes made in the station as described in the FSAR.

• f.

Special nuclear material inventory records.

• g.

Routine station radiation surveys and monitoring records.

• h.

Records of Environ = ental Off-Site Monitoring Surveys.

• i. Records of radiation exposure for all station personnel, contractors and visitors to the station as required by 10 CFR 20.
• j. Records of radioactive releases and waste disposal.

k. Records of reactor physics tests and other special tests pertaining to nuclear safety. 1. Changes to Operating Procedures, m. Shift Supervisors log. By-product material inventory records and source leak test results. n. Station Review Committee and Nuclear Safety Review Co=mittee minutes. o.

• These items will be permanently retained.

6.5-1

1 (b) A summary al results of surveillance tests and inspections. (c) The results of any periodic containment leak rate tests per-formed during the reporting period. (d) A brief summary of those changes, tests, and experiments requiring authorization from the Commission pursuant to 10CFR50.59a. (e) Any changes in plant operating organization which involve positions for which minimum qualifications are specified in the Technical Specifications. (2) Power Generation A summary of the nuclear and electrical output of the unit during the reporting period, and the cumulative total outputs since initial criticality, including: (a) Gross thermal power generated (in MWH). (b) Gross electrical power generated (in MWH). (c) Net electrical power generated (in MWH). (d) Number of hours the reactor was critical. (e) Number of hours the generator was on line. (f) Histogram of thermal power versus time. (3) Shutdowns Descriptive material covering all outages occurring during the report-ing period. The following information shall be provided for each outage: (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. (d) Unit status during the outage; e.g., cold shutdown, hot shutdown, or hot standby. l (e) Corrective and preventive action taken to preclude recurrence j of each unplanned outage. .) 6.6-2

6.6 STATION REPORTING REQUIRDiENTS 6.6.1 Routine Reports 6.6.1.1 Operating Reports The following reports shall be submitted to the Directorate of Licensing, USAEC, Washington, D. C., 20545 A. Startup Report 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 commencement of commercial power, (i.e. following synchronization of the turbo-generator to produce ccm=ercial power). B. First Year Operation Reeort A report submitted within 60 days of completion of one year of commercial operation covering: (1) An evaluatica cf unit perfcrmance to date in comparisen with design specifications. (2) A reassessment of the validity of prior accident analysts in light of measured operating characteristics, which may affect conse-quences; and system, component, and personnel perfor=ance which may affect accident probabilities. (3) A progress and status report on all ite=s identified in the oper-ating license review as requiring further effort. (4) An assessment of the performance of structures, systems, and com-ponen'.s important to safety. C. Semi-Annual Ooerating Report A Semi-Annual Station Operations Report shall be prepared and submitted within 60 days after the end of each reporting period. The report shall provide the following information (summarized on a conthly 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. (1) Operations Summary (a) A narrative su= mary of operating experience and of changes in facility design that relate to safe operation, performance characteristics (including fuel performance) and operating procedures related to safety occurring during the reporting period. 6.6-1

4 (3) Percentage of limit. '} (c). rticulate Releases (1) Gross radioactivity (S y) released (in curies) excluding background radioactivity. (2) Gross alpha radioactivity released (in curies) excluding background radioactivity. (3) Total radioactivity released (in curies) of nuclides with half-lives greater than eight daye. (4) Percentage of limit. (d) Liquid Releases (1) Gross radioactivity (S y) released (in curies) and average concentration released to the unrestricted area at the j Keowee Hydro unit. (2) Total tritium and alpha radioactivity (in curies) released and average concentration released to the unrestricted area at the Keewee Hydro unit. (3) Total dissolved gas radioactivity (in curies) and average ~ concentration released to the unrestricted area at the ) Keowee Hydro unit. (4) Total volume (in liters) of Keowee Hydro liquid waste released. (5) Total volume (in liters) of dilution water used prior to release from the restricted area. (6) The maximum' concentration of gross radioactivity (S y) released to the unrestricted area (averaged over the period of a single release). (7) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed. (8) Percentage of limit for total activity released. (e) Solid Waste (1) The total amount of solid waste packaged (in cubic feet). (2) The dates of shipment and disposition (if shipped off-site). (3) Estimated total radioactivity (in curies): ) 6.6-4

4 (4) Maintenance A discussion of saf ety-related maintenance (excluding preventive maintenance performed during the reporting period on systems and components that are designed to prevent or mitigate the consequences of postulated accidents or to prevent the release of significant amounts of radioactive material). For any malfunction for which corrective maintenance was required, infor=ation should be pro-vided on: ) (a) The system or component involved. (b) The cause of the malfunction. ] (c) The results and effect on safe operation. (d) Corrective action taken to prevent repetition. (e) Precautions taken to provide for reactor safety during repair. 1 (5) Chances. Tests, and Experiments A summary of all changes in the facility design and procedures that relate to the safe operation of the facility should be included in the Operations Summary section of this semi-annual report. This section should include a brief description and the su= mary of the safety evaluction for those changes, tests, and j experiments carried out without prior Commission approval. (6) Reporting of Radioactive Effluent Releases Data shall be reported to the Commission in the form shown in Table 6.6-1 and shall include the following: (a) Gaseous Releases (1) Total radioactivity (in curies) releases of noble and activation gases. (2) Maximum noble gas release rate during any one-hour period. (3) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed. (4) Percentage applicable limits released. (b) Iodine Releases (1) Total (I-131, I-133, I-135) radioactivity (in curies) 4 released. (2) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed. 6.6-3

~ B. A report shall be submitted to the employee and the Directorate of 'S i Licensing, USAEC, Washington, D. C., within 30 days af ter the exposure determination or 90 days from the termination of employment, whichever comes first, on the total exposure to radiation and radioactive material received during the period of employment. 6.6.1.3 Material Statts A. The licensee will file Form AEC-742 within 30 days of December 31 and June 30 to report the status of all special nuclear materials. B. The licensee will file Form AEC-741 within 10 days of shipping or receiving special nuclear =aterial. 6.6.2 NON-ROUTINE REPORTS 6.6.2.1 Reporting of Abnormal Occurrences & Unusual Events A. Events requiring notification within 24 hours (by telephone or telegraph to the Director of Region II Regulatory Operations Office followed by a written report within 10 days to the Directorate of Licensing, USAEC, Washington, D. C. 20545. (1) Abnormal occurrences specified in Section 1.8 of the Technical Specifications. (2) Any significant variation of measured values in a non-conservative r ~# direction from 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 (1) and (2) above. B. Unusual Events as defined in Section 1.9 of the Technical Specifications shall be reported within 30 days to the Directorate of Licensing and to Directorate of Regulatory Operations, Region II, Atlanta, Georgia. the 6.6.2.2 Radiation Exposure and Monitoring The licensee will report any over exposure, excessive radiation level or concentration to the Directorate of Regulatory Operations, USAEC, Washington, D. C. 20545, and Regulatory Operations, Region 7L L:lanta, Georgia, per 10CFR20. 6.6.2.3 Loss of Licensed Material A. The licensee will report immediately of telephena or telegraph the theft or loss of any licensed material in such quan(ities and under such circumstances that a substantial hazard may result to persons in an unrestricted area. 6.6-6

a a. Packaged b. Shipped (f) Environmental Monitoring (1) For each medium sampled during the six-month period, the following information shall be provided. a. Number of sampling locations. b. Total number ot' samples c. Number of locations at which levels are found to be significantly greater than local backgrounds. d. Highest, lowest, and the average concentrations or levels of radiation for the sampling point with the highest average and description of the location of that point with respect to the site. (2) If levels of station contributed radioactive materfsis in environmental media indicate the likelihood of public intakes in excess of 3 percent of those that could result 3 from continuous exposure to the concentration values listed in Appendix 3, Iable II, Part 20, estimates of the likely resultant exposure to individuals and to population groups, and assumptions upon 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 station releases, correlation of these results with effluent releases shall be provided. 6.6.1.2 Personnel Exposure and Monitoring Reports A. This report shall be submitted to the Directorate of Licensing, USAEC, Washington, D. C., 20545 within the first quarter of each calendar year. (1) A report of the total number of individuals for whom personnel monitoring was provided during the calendar year. (2) A report of individuals, 18 years of age or older whose annual radiation dose exceeded the applicable quarterly numerical values, and for each individual under 18 years of age whose annual dose exceeded 10 percent of the applicable quarterly numerical values will be submitted, i 6.6-5

6.6.3.3 Single Loop Operation Report i A report covering single loop operation, permitted by Specification 3.1.8, within 90 days after completion of 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 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 flow 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). 6.6.3.4 Reactor Building Structural Tests The results of the initial reactor building structural tests (as specified in FSAR Section 5.6.1.2 including the Structural Instrumentation Report contained in Amendment No. 25, dated Dece=ber 30, 1970) shall be reported ~ within 90 days following completion of the test. 6.6.3.5 Reactor Building Structural Integrity Report A reactor building structural integrity 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 6.6.5.7b 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 6.6.3.6 Inservice Inspection Program Report The Inservice Inspection program shall be performed as specified in Technical Specification 4.2. 6.6.3.7 Fuel Surveillance Program Report Report to be submitted upon completion of program on fuel surveillance per Specification 4.13. I J 6.6-8

o l B. Within 30 days af ter the loss of a quantity licensed material, the licensee will file a written report with the Directorate of Licensing, USAEC, Washington, D. C., and the Regulatory Operations, Region II Office, Atlants, Georgia, containing the following information: (1) A description of the licensed material involved, including kind, quantity, chemical and physical form; (2) A description of the circumstances under which the loss or theft occurred; (3) A statement of disposition or probable disposition of the licensed material involved; (4) Radiation exposures to individuals, circumstances under which the exposures occurred, and the extent of possible hazard to persons in unrestricted areas; (5) Actions which have been taken, or will be taken, to recover the material; and (6) Procedures or measures which have been or will be adopted to prevent a recurrence of the loss or theft of licensed material. 6.6.2.4 Accidental Criticality The licensee will report prospely any accidental criticality to Regulatory Operations, Region II, Atlanta, Georgia. 6.6.2.5 Incidents Involving Licensed Material In the event of an incident involving licensed material, the licensee will immediately notify Regulatory Operations, Region II, Atlanta, Georgia. 6.6.3 Special Reports The following reports shall be prepared and submitted to the Directorate of Licensing, USAEC, Washington, D. C., 20545. 6.6.3.1 Authorization of Changes, Test, and Experiments The licensee will file a request with the Directorate of Licensing, USAEC, Warhington, D. C., for authorization of a change in technical specifications or any change, test or experiment which requires authorization by the Commission. 6.6.3.2 Reactor Building Integrated Leak Rate Test The initial reactor building integrated leak rate test shall be the subject of a summary technical report and shall be submitted within 90 days and shall include analyses and interpretations of the results which demonstrate compliance in meeting the leak rate limits specified in the Technical Specifications. Other containment leak rate tests that fail to meet the acceptance criteria shall be the subject of a special summary report. 6.6-7

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w o (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 may 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: ) (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. \\' 6.7-2 l

i o . 6.7 RADIOLOGICAL CONTROLS Specification 6.7.1 The radiation protection program shall be organize 4 with the following exceptions, to meet the requirements of 10 CFR 20. Pursuant to 10 CFR 20.103(c)(1) and (3), allowance may be made a. 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 in Appendix B, Table I, Column 1, of 10 CFR 20, subject to the following conditions and limitations: 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 materials to levels belew 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-active material, the licensee maf make allowance for such use in estimating exposures of Individuals to such naterials provided: (a) Intake of radioactive material by any individual within any period of seven consecutive days will not exceed that which would result frca 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. If 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 3 for H, 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 upcn 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 procedures 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 may be appropriate to the circumstances of the occurrence with proper consideration of critical organs and limiting doses, s 6.7-1 J

TAllLE 6.7-1 PROTECTION FACTORS FOR RESPIRATORS I PROTECTION FACTORS 2/ GUIDES TO SELECTION OF EQUIPMENT PARTICULATES BUREAU OF MINES APPROVAL SCllEDULES* AND VAPORS AND FOR EQUIPMENT CAPABLE OF PROVIDING AT CASES EXCEPT LEAST EQUIVALENT PROTECTION FACTORS TRIT 1UM OXIDg3/

• or schedule superseding for equipment DESCRIPTION MODES 1/

of type listed I. AIR-PURIFYING RESPIRATORS Faceplece, half-mask 4/ 7/ NP 5 21B 30 CFR 14.4 (b) (4) Faceplece, full 7/ NP 100 21B 30 CFR 14.4(b)(5); 14F 30 CFR 13 II. ATMOSPilERE-SUPPLYING RESPIRATOR i 1. Airline respirator Facepiece, half-mask CF 100 19B 30 CFR 12.2(c)(2) Type C(i) Facepiece, full CF 1,000 19B 30 CFR 12.2(c)(2) Type C(i) Faceplece, full ]/ D 500 19B 30 CFR 12.2(c)(2) Type C(ii) P Facepiece, full PD 1,000 19B 30 CFR 12.2(c)(2) Type C(iii) y liood CF 5/ See note 6/ Suit CF 5/ See note 6/ 2. Self-contained breathing apparatus (SCBA) Facepiece, full J/ D 500 13E 30 CFR ll. 4 (b) (2) (i) Facepiece, full PD 1,000 13E 30 CFR ll.4 (b) (2) (ii) Facepiece, full R 1,000 13E 30 CFR 11.4 (b) (1) III. COMBINATION RESPIRATOR Any combination of air-Protection factor for 19 B CFR 12.2(e) or applicable purifying and atmosphere-type and mode of opera-schedules as listed above supplying respirator tion as listed above 1/, 2_/,.3_/,,4_/, 5/, 6/, 7_/, [These notes are on the following pages] i e M

(6) Bioassays and/or whole body counts of individuals, and other surveys, as appropriate, to evaluate in-dividual exposures and to assess protection actually provided. (7) Records sufficient to permit periodic evaluation of the adequacy of the respiratory protective program. (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 evaluated the equipment and can demonstrate by testing, or on the basis of reliable test informatica, that the material and performance characteristics of the equipment are at least equal to those afforded by U. S. Bureau of : lines 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 these 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-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). 6.7-3 sm

x. 5/ Appropriate protection factors must be determined taking account of the design of the suit or hood and its permeability to the contaminant under l conditions of use. No protection factor greater than 1,000 shall be used except as authorized by the Commission. 6/ No approval schedules currently available for this equipment. Equipment must be evaluated by testing or on basis of available test information. 2/ Only for clean-shaven faces. NOTE 1: Protection factors for respirators, as may be approved by the U. S. Bureau of Mines according to approval schedules for respirators to protect against airborne radionuclides, may be used to the extent that they do not exceed the protection factors listed in this Table. The protection factors in this Table may not be appropriate to circumstances where chemical or other respiratory hazards exist in addition to radioactive hazards. The selection and use of respirators for such circumstances should take into account approvals of the U. S. Bureau of Mines in accordance with its applicable schedules. NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table I of 10 CFR Part 20 are based on internal dose due to in-halation may, in addition, present external exposure hazards at higher concentrations. Under such circumstances, limitations on occupancy may have to be governed by external dose limits. T 6.7-6 J /}}