ML19322A757
| ML19322A757 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/01/1966 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7911210781 | |
| Download: ML19322A757 (41) | |
Text
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TABLE OF COUTE:ITS Section M 11 RADIOACTIVE WASTES AND RADIATION PROTECTION 11-1 11.1 RADIOACTIVE WASTES 11-1 11.1.1 DESIGN BASES 11-1 11.1.1.1 Performance Objective 11-1 11.1.1.2 Radioactive Waste Quantities 11-1 11.1.1 3 Waste Activity H-1 11.1.1.4 Disposal Methods 11-5 11.1.1 5 Shielding H-5 11.1.2 SYSTEM DESIGN AND EVALUATION H-6 11.1.2.1 Solid Waste Disposal System 11-6 H.1.2.2 Liquid Waste Disposal System 11-6 11.1.2 3 Gaseous Waste Disposal System n-10 11.1.2.4 Process Radiation Monitoring 11-10 11.1.2 5 Design Evaluation n-ll 11.1 3 TESTS AND HISPECTIONS H-18 11.2 RADIATION PROTECTION 11-19 11.2.1 PRIMARY, SECONDARY, AND REACTOR BUILDHIG Sn m aniG 11-19
- 11.2.1.1 Design Bases H-19 11.2.1.2 Description 11-19 11.2.1 3 Evaluation 11-20 11.2.2 AREA RADIATION MONITORING SYSTEM 11-22 H.2.2.1 Design Bases 11-22 11.2.2.2 Description 11-22 n.2.2 3 Evaluation H-23
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H-1 269
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< - COITIENTS (Cont'd)
Section Pg 11.2 3 HEALTH PHYSICS 11-23
- 11.2 3 1 Personnel Monitoring Systems 11-24 11.2 3 2 Personnel Protective Equipment 11-24 11.2 3 3 Change Room Facilities 11-24 11.2 3 4 Health Physics Laboratory Facilities 11-23 11.2 3 5 Health Physics Instrumentation 11-25 11.2 3 6 Medical Programs 11-25 11 3 REFERECES H-26 O
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LIST OF TABLES Table No. Title Page 11-1 Radioactive Waste Quantities From Two Units 11-2 i 11-2 Escape Rate Coefficients for Fission rroduct Release 11-4 11-3 Reactor Coolant Activities From One Per Cent Defective Fuel 11-4 11-4 Waste Disposal System Component Data 11-7 11-5 Maximum Activity Concentrations in the Station Effluent for Both Units Operating With One Per Cent Failed Fuel 11-12 11-6 Waste Disposal System Failure Analysis 11-17 i
LIST OF FIGLEES (At Rear of Section)
Figure No. Title 11-1 Waste Disposal System i
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11 PADI0 ACTIVE WASTES AND RADIATION PROTECTION 11.1 RADIOACTIVE WASTES 11.1.1 DESIGN BASES 11.1.1.1 Perfor:ance Objective The Waste Disposal System is designed to provide for controlled handling and disposal of liquid, gaseous, and solid wastes from Units 1 and 2 of the Oconee Nuclear Station. The principal desi6 n criterion is to insure that station personnel and the General public are protected against excessive exposure to radioactive caterial in accordance with the regulations of 10 CFR 20.
11.1.1.2 Radioactive Waste Quantities The estimated volumes of the various forms of radioactive wastes generated in the station are listed in Table 11-1.
11.1.1 3 Waste Activity Activity acetmulation in the reactor coolant system and associated waste handling equipment has been determined on the basis of fission product leakage through clad defects in 1'per cent of the fuel. The activity levels were computed as-suming full power operation of 2,568 mwt for one core cycle with no defective fuel followed by operation over the second core cycle with 1 per cent defective p) t fuel. Continuous reactor coolant purification at a rate of one reactor system V volume per day was used with a 50 per cent removal efficiency for Cs, a zero removal efficiency for Kr and Xe, and a 99 per cent removal efficiency for an other nuclides except Te, which was assumed to deposit on the system surfaces.
Activity levels are relatively insensitive to small changes in demineralizer efficiencies, e.g., use of 90 per cent instead of 99 per cent would result in only about a 10 per cent increase in the coolant activity.
The quantity of fission products released to the reactor coolant during steady state operation is based on the use of " escape rate coefficients" (sec-1) as determined from experiments involving purposely defected fuel elements.(1,2,3,4 Values of the escape rate coefficients used in the calculations are shown in Table 11-2.
Calculations of O e activity released from the fuel were performed with a digi-tal enmputer code wL!ch solves the differential equations for a five-member radioactive chain for buildup in the fuel, release to the coolant, removal from the coolant by purification and leakage, and collection on a resin or in a hold-up tank. The activity levels in the reactor coolant during full power operation at the end of the second core cycle are shown in Table 11-3 m
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V Table 11-1 Radioactive Maste Quantities From Two Units Quantity / Year, Waste Source ft3 Assumptions and Consnents Liquid Waste Reactor Coolant System Startup Expansion 26,200 Startup once per quarter per unit from cold condition.
Startup Dilution 8,200 One startup from cold condition at 190 and 266 full power days, respectively, per unit.
Lifetime Shim Bleed 44,000 Dilution 2,270 to 175 ppm in each unit.
f ' System Drain 12,200 Drain of each unit to level of outlet nozzles N during refueling.
Sampling and Iaboratory Drains 5,900 24 samples per day at 5 gal per sample.
Purification Demin. Sluice 320 160 ft 3/ year replacement r.t a ft3 /rt3 resin sluice.
Spent Fuel Pool Demin. Sluice 84 42 ft 3/ year replacement at 2 ft3 /ft3 resin sluice.
Deborating Demin. Regen, and Rinse 40o0 -@ 2 regenerations per year at 20 ft3/ft3 resin re-generation.
[ Misc. System Leakage 11,700 10 gph leakage.
g laundry 14,600 300 gpd.
N$ Showers 29,200 20 showers per day at 30 gal per shower.
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Table 11-1 (Cont'd)
Quantity / Year, Waste Source ft3 Assumptions and Comments Gaseous Waste (a)
Off-Gas From Reactor Coolant System 2,700 Degas at 25cc Hf per liter concentration.
Liquid Effluent and Misc. Leakage Off-Gas From Liquid Sampling 148 Degas at 25cc H2 per liter concentration.
Letdown Storage Tank 900 Vent once per year.
Pressurizer 60 Vent once per year.
Solid Waste b
t'a Purification Resin 160 Resin replacement twice per year.
Spent Fuel Pool Demin. Resin 42 Resin replacenent twice per year.
Evap. Cond. Demin. Resin 4 Resin replacement twice per year.
Evaporator Bottoms 1, 32 0 Evaporation to 25 per cent solids for reactor coolant system liquid effluent and deborating demineralizer regenerant and rinse.
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M (a) Excluding reactor building and station ventilation.
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O Table 11-2 Escape Rate Coefficients for Fission Product Release Escape Rate Coefficient, Element sec-1 xe 1.0 x 10-7 Kr 1.0 x 10-I Br 2.0x10]0 2.0 x 10 0 Cs Rb 2.0 2.0 xx 10 10-0 Mo 4.0 x 10-9 Te 4.0 x 10-9 sr 2.0 x 10-10 Ba 2.0 x 10-10 Zr 1.0 x 10-11 Ce and other rare earths 1.0 x 10-11 O
Table 11-3 Reactor Coolant Activities From one Per Cent Defective Fuel Isotope Activity, nc/ml Isotope Activity, pc/mi Kr 85m 2.0 I 131 33 Kr 85 15 5 I 132 49 Kr 87 1.1 I 133 45 Kr 88 37 I 134 0 55 l Rb 88 37 I 135 2.1 sr 89 0.057 Cs 137 0 39 sr 90 0.0m8 Cs 138 o.68 sr 91 0.057 Mo 99 1.2 sr ge 0.018 Ba 139 0.088 xe 131m 2.1 Ba 140 0.076 xe 133m 32 Ia 1h0 0.026 xe 133 290.0 Y 90 0.0007 xe 135m 1.0 Y 91 0.0043 xe 135 94 Ce 144 0.0027 Xe 138 o,5 i
O rists n.2, u -3 N 11 4 275
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The liquid vaste generated by leakage, samplin6, and de=ineralizer sluice or rinse is assu=ed to have an activity concentration equal to the concentration in the reactor coolant. Reactor coolant bleed is taken from the downstream side of the purification demineralizer. It is assu=ed to have the sa=e activ-ity concentration as the reactor coolant reduced by the deconta=ination factor of the purification decineralizer. Laundry and shower vastes are assumed to contain negligible a=ounts of radioactivity.
Gaseous activity is generated by the evolution of radioactive gases from )
liquids stored in tanks throughout the station. Therefore, the activity of I the gases is dependent upon the liquid activity. The assumptions for liquid activity are described above. The resulting Easeous activities are described in 11.1.2 5, Design Evaluation.
11.1.1.4 Disposal Methods Liquid wastes from the station are handled in one of three ways:
- a. Collected, monitored, and discharged directly to the Keowee Hydro ta11 race,
- b. Collected, monitored, held up for decay, and then discharged to the Keowee Hydro.ta11 race.
- c. Collected, monitored, concentrated, packaged, and shipped offsite.
Gaseous wastes are disposed of using one of two methods:
- a. Continuous dilution and discharge through vaste gas filters to the station vent with sweep gas being drawn through tank voids.
- b. Diversion to waste gas holdup tanks with sampling and controlled sube sequent release through waste gas filters to the station vent.
Solid radioactive wastes are ace-hted and packed in containers suitable for ICC-approved shipment offsite to a licensed waste disposal facility.
11.1.1 5 Shielding Shielding for the co=ponents of the vaste disposal system will be designed on the basis of system activity levels with 1 per cent failed fuel. With the exception of the quench tank and the rersetor building sumps, all components are located in the auxiliary building. The shield design criteria for the auxiliary building is a dose rate of 2.0 crem/hr in nor= ally accessible areas and15crem/hrinareasrequiringlimitedaccess. Tt3 m.'ponenus of the vaste disposal system will be shielded by concrete valls and floors of vary-ing thickness depending on the =a6nitudes of the sources in each component and on the access requirements in a particular area. In soce areas local shielding in the form of lead or removable concrete blocks vill be utilized to facilitate maintenance or repair operations.
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O Q 11.1.2 SYSTEM DESIGN AND EVAIUATION 11.1.2.1 Solid Waste Disposal System Solid wastes are placed in ICC-approved containers appropriate for the vaste material. Icaded containers are monitored for surface radiation levels and stored in a special area prior to shipment to an offsite disposal facility.
Evaporator concentrate from the evaporator that does not contain reusable beric acid is pumped into a shipping cask for offsite disposal. Spent resins from the demineralizers are sluiced to a spent resin storage tank, and the sluice water is transferred from the tank to the miseenaneous waste holdup tank.
The spent resin storage tank can hold one complete charge of resins from the reactor auxiliary systems. Spent resin is transferred from the storage tank to a chipping cask for disposal. In the cask they are mixed with cement and vermiculite, if necessary, to form a fixed matrix.
Other miscellaneous solid wastes are disposed of using a baler and light metal shipping containers.
11.1.2.2 Liquid Waste Disposal System Liquid vastes cre accumulated in storage tanks and, depending on their composi-tion as determined by local sampling and laboratory analysis, are either dis-char;ed throu@ a disposal line to the tailrace cf Keovee Hydro, or are pro-cessed by evaporation to concentrate impurities for ultimate aisposal and to O provide for return of purified water for reuse as makeup.
All piping and equipment in contact with reactor coolant are constructed of corrosion-resistant material. This equipment is arranged and located to per-mit detection and collection of system loss'es and to prevent escape of any un-monitored radioactive liquid to the environment. A flow diagram of this sys-ten with the necessary instrumentation and controls for operation is shown in Figure 11-1. Component data is shown in Table 11-4. Control of equipment is ,
from localfy mounted control panels. Shielding of equipment insures operator I protection.
Reactor coolant is received from the high pressure injection and purification system and is the largest single source of operational liquid waste to be han-died. The liquid is received as a result of reactor coolant expansion and operational requirements for reduction of the reactor coolant boric acid con-tent. It is either conveyed to reactor coolant bleed holdup tankc for storage, or passed through deborating demineralizers for boric acid removal and returned as unborated makeup to the high pressure injection and purification system. The liquid received from reactor coolant system expansion is normally stored and reused as unprocessed borated makeup during a subsequent reactor coolant system cooldown. Liquid having a boric acid concentration above 1,000 ppm is normally conveyed to bleed storage and subsequently evaporated for re-moval of boron and impurities. Liquid having a boric acid concentration below 1,000 ppm is passed through the deborating demineralizers and returned to the high pressure injection andtpurification system as unborated makeup.
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Table 11 4 Waste Disposal System Component Data (Component quantities for two units)
Quench Tank Number 2
, Volume Esca, cu ft 1,000 Material Carbon Steel, Corrosion-Resistant Lining Deborating Demineralizer Number 3 Type Semiautomatic Regeneration Material Carbon Steel, Corrosion-Resistant Lining Reactor Coolant Bleed Holdup Tank Number 6 Volume Each, eu ft 11,000 Material Aluminum Miscellaneous Waste Holdup Tank Number 1 Volume, cu ft 2,700 d Material Carbon Steel, Corrosion-Resistant Lining Waste Neutralization Tank Number 1 Volume, cu ft 400 Material Carbon Steel, Corrosion-Resistant Lining Spent Resin Storage Tank Number 1 Volume, eu ft 450
- terial Carbon Steel, Corrosion-Resistant Lining Evaporator Condr'. sate Test Tank Number 2 Volume Each, cu ft 400 l Material Aluminum Waste Evaporator Number 1 ProcessRate,lb/hr 1,000 Material Stainless Steel 1
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Table 11 4 (Cont'd )
Evaporator Condenser Number 1 Heat Transferred, Btu /hr x 10 4 1 Material,shell/ tube Stainless Steel / Stainless Steel Reactor Coolant Water Flow, gpm 100 Condensate Demineralizers Number 2 Material Stainless Steel er a .
Number 2 Capacity Each, gpm 600 Material Stainless Steel Wste Transfer Pu=p '
Number 2 Capacity Each, gpm 100 Material Stainless Steel Auxiliary Building Sump Tank Pump Number 2 Capacity Each, gpm 50 Material Stainless Steel ,
Evaporator Feed Pump Number 2 Capacity Each, gpm 5 Material Stainless Steel Evaporator Condensate Pump Number 2 Capacity Each, gpm 20 Material Stainless Steel Evaporator Concentrate Pump Number 2 Capacity Each, gpm 5 Material Stainless Steel j Evaporator Vacuum Pump l Number 1 l
Capacity, cfm 6 Material Carbon Steel Waste Gas Compressor Number 2 Capacity Each, cfm 20 Material Carbon Steel e TABLE ll-b (cont'd) 11-8 279
Table 11 4 (Cont'd)
Waste Cas Decay Tank Number 2 Volume Each, cu ft 1,350 Material Carbon Steel Waste Gas Filter Hu=ber 1 Type Pre, Absolute, and Charcoal Filter Combination The bleed storage capacity consists of six reactor coolant bleed holdup tanks, each sized to contain one reactor coolant system volu=e, and a miscellaneous vaste holdup tank sized to contain one-fourth of a reactor coolant system vol-ume.
Reactor building drainage and minor miscellaneous leaka6e from equipment lo-cated in the reactor building is collected in the reactor building sump. Liquid is pmped from the sump to the miscellaneous vaste holdup tank located outside of the reactor cuilding.
A quench tank, located inside the reactor building, condenses and contains any
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effluent from the pressurizer safety valve:.. The quench tank is sized to con-dense one nor=al pressurizer steam volume (700 ft3 at 2185 psig) without reliev-ing to the reactor buildin6 atmosphere.
The miscellaneous vaste holdup tank collects resin sluice water from the spent resin storage tank, liquids pumped from the reactor building sump, spent fuel storage pool overflow, and condensate drainage from the vaste gas disposal equip-ment. Liquids from this tank are pumped to the required disposal process follow-ins sampling and laboratory analysis.
A' divided auxiliary building sump tank is provided. One tank section collects lov activity vastes, such as laundry and shower vastes, which are normally dis-charged directly to the tailrace of Keovee Hydro. The other tank collects liquids from auxiliary equipnent and floor drains. The disposal method is decided by sampling and laboratory analysis.
Accumulated liquid vastes requiring processing are evaporated. They are first pumped to a vaste neutralization tank, where the pH is adjusted-if required; and then fed to the vaste evaporator. Distillate from the evaporator is con-densed and collected in an evaporator condenstate test tank, where the vcter quality and activity level are checked. Condensate is then pumped through a condencate demineralizer to either the station demineralized water stora6e tank for reuse, or to the Keovee Hydro tailrace for disposal. The evaporator train is designed to provide a decontamination factor of 104 , and is capable of processing eight reactor coolant system volumes in 180 days of continual oper-ation.
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rhe ficy rcte and activity of all liquid discharged from this system to the _
Keovee FJd ro tailrace are indicated and integrated te maintain release within safo limits. An alarm notifies the operator of discharge of activity above preset values. The process radiation monitoring equipment is discussed in 11.1.2.4.
11.1.2 3 Gaseous Waste Disposal System All components that can potentially contain radiocetive off-;ases are vented throu;;h collection lines to a vent header. The. vent ;ases are subsequently drawn from this vent header by one of two vaste gas co= pressors, which dischar e through vaste 3as moisture separators to either the station vent or to vaste cas decay tanks for holdup prior to relecse to the station vent. A filter bank con-tcinine a prefilter, an absolute filter, and a charcoal filter is installed in the dischar;c line for filtering all vaste cas prior to release to the station vent.
The vent header noted above is normally separated into two sections by a re-motely operated valve. One of these sections receives gases from the pressur-izer, the letdown storage tank, the gas sampling lines, and the quench tank.
Gases from this section of the header are nor= ally passed through one of the two vaste gas decay tanks prior to release to the station vent since their po-tential radioactivity level is higher than levels from other vents. All other vents are connected to the second section of the header and are normally dis-
- harged through the waste gas filters directly to the station vent. In the event of above-tolerance activity release as determined by monitoring of the discharge line to the Station vent, the air purge inlets are closed, and these vents will be diverted to the vaste gas decay tanks for decay prior to release. .
A flow diagram of this system with the instrumentation and controls for opera-tion is given in Figure 11-1. Component data is contained in Table 114. Con-trol of equipment is from locally mounted control panels. Shielding and ade-quate ventilation insure operator protect en. The gases in the vaste gas de-cay tanks are sampled and analyzed prior to discharge to evaluate the isotopes present and to establish proper dilution rates. The gaseous effluent is fil-tered, diluted, and discharged to the station vent. Radiation =cnitoring (see 11.1.2.4) of the effluent insures gaseous activity discharge rates within 10 CFR 20 limits.
11.1.2.h Process Radiation Monitoring Radiation monitoring of station effluent vill include alarms and indications designed to provide early warning of possible equipment calfunctions or poten-tial biological hazards. Station effluents will be monitored to instu e that prescribed limits of radiation release are not ' exceeded. The release of gaseous and liquid effluents vill be controlled within the limits of 10 CFR 20.
All gaseous radioactivity vill be discharged to the atmosphere through the sta-tion vent. The source of this activity is the gaseous vaste disposal system.
After a delay in the holdup tanks, the gas will be analyzed prior to release.
Flov in the discharge line from the vaste disposal system and flov in the sta-tion vent will be monitored to insure that the system is operating correctly and that the releases are within the limits of 10 CFR 20.
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D A monitor located in the gaseous vaste discharge line to the Station vent vill be equipped with an indicator and an alam to annunciate a high activity level.
A high-level interlock will stop the discharge of gaseous effluents from the vaste disposal system and direct the effluent to the vaste gas decay tank.
The condenser air ejector discharge vill be cocitored for gaseous activity.
Should leakage of reactor coolant to the steam occur and result in high activity levels, the monitor on the air ejector vill initiate an alarm in the control room.
All liquid vaste vill be collected, stored, and analyzed for radioactive con-centration prior to disposal by evaporation er discharge. If discharge to the envircnment is permissible, a flow indicator and appropriate valving vill permit controlled release. Activity concentration vill be determined by sampling of the stored liquid vaste prior to release. The flow rate and activity of all liquids discharged from the vaste disposal system vill be indicated and record-ed. An alarm notifies the operator of dischsrge of activity above preset values such that 10 CFR 20 vill not be exceeded.
The cooling water systems that remove heat frca y.tentially radioactive sources
) are monitored to detect accidental releases. ::enitors will be provided on the low pressure service water outlets from the reactor coolant pu=p seal return coolers spent fuel coolers, low pressure injection and decay heat removal coolers deactor building ccaponent coolers, and reactor building spray coolers.
Radiation monitors vill be located on the reactor building co.ponent cooling water lines to detect leakage of reactor coolent into the reactor building com-ponent cooling system. An alarm vill alert ths operator, and the heat ex-changer can be isolated. In addition, a monitor vill be located on the lov pressure service water header upstream of itc pcint of discharge. This monitor vill serve as a backup to the preceding monit: n. Alares vill alert the oper-ator to isolate the source of release.
Reactor coolant letdown flow vill be monitorea to detect a gross fuel assembly failure. A smaller fuel asse=bly leak vill be detected by regular laboratory analycis of reacter coolant samples.
Air s:.=ples frc.; the recetor buildings and th: s:aticn vent will be conitored for nir particulate, scsccus, cc.d icdine actisity.
These radiation monitors are co=mercially available equipment. The required charneteristics vill be established during detailed station design. The mini-mum se nsitivity of det ;ctors, when Ocebined with appropri2te dilution factors, vill insure safe limits of release.
11.1.2 5 Design Evaluation r
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(V All ualyses on liquid and gaseous vaste disposal were performed on the basis of bc:h units operatin; with 1 per cent faile; mel. Although it is not ex-pectel that the number of clad defects vill e fer sp;rcach 1 per cent of the 7- 'e" 11-11 ...
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total fuel, the objective is to demonstrate the capability of safe station operation within the limits of 10 CFR 20 with quantities of radioactive fission products in the system. Thus, the adequacy of the waste disposal system design is demonstrated.
A summary of the various operations considered in the analyses, and the total concentrations resulting in the station effluent from operation of two units with failed fuel, are given in Table 11-5 The activity concentrations result-ing are given as fractions of the MPC for unrestricted areas, i.e., the concen-tration of each radioactive nuclide has been divided by its respective Faximum Permissible Concentration for discharge into unrestricted areas as set forth in 10 CFR 20.
Table 11-5 Maximum Activity Concentrations in the Station Effluent for Both Units Operating With One Per Cent Failed Fuel Liquid Waste Yearly Average Concentration in Ta11 race Discharge, Operation Fraction of MPC Lifetime Shim Bleed Including Startup Expansion and Dilution 0.05 Discharge of Miscellaneous Wastes 0.06 Gaseous Wastes -
Yearly Average Concentration at Site Boundary, Operation Fraction of MPC Lifetime Shim Bleed 0.04 Startup P.xpansion and Dilution 0.03 Venting of letdown Storage Tank O.004 Venting of Pressurizer 0.002 Reactor Building Purge 0.006 Steam Generator Tube Ieakage of 1 gpm 0.07 11.1.251 Liquid Wastes The average coolant bleed rate per unit over a core cycle is a. bout 30 gph, in- '
cluding startup expansion and letdown to storage for boric acid reduction.
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n The activity level in the station effluent was determined by assuming that re-actor coolant system liquil was discharged throu6hout the entire chemical shim bleed life of 266 full power days. Iatdown throudh the purification deminer-alizer was assumed to give a decontamination factor vf 100 for iodine, zero for cesium, and 10 for all other elements for the purposes of this calculation.
Activity levels in the reactor coolant system were those at the end of the second core cycle, i.e., the maximum levels. Discharge to the Keowce Hydro tailrace at the rate of 60 gph allows for operation of both Oconee units. No holdup or decay was assumed. Dilution through the ta11 race was the average stream flow of 1,100 cfs. The results were a concentration level of 0 3 times the !EO for discharge into unrestricted areas. If the ecolant bleed is held up for a period of 30 days, the concentration level is reduced to 0.05 times the MFC.
Six reactor coolant bleed holdup tanks, each with a capacity of 11,000 ft3, are provided for a total storage capacity of 66,000 ft3 The mv4- quantity of coolant 8,000 ftgetdown for bothforunits.
chemical shin, Thus, onlyduring any 30 two tanks, or day period .in one-third, of life, the is about available stora6e capacity, are required to provide a 30 day holdup period for all of the coo 3 ant which is bled down for lifetime chemical shim.
The maximum volume of coolant removed from one unit during heatup and dilution to startup from a cold shutdown is 7,000 ft3 . This occurs at the end of the chemical shim period. Earlier in life the quantity removed would be less than this, e.g., prior to 190 full power days no dil..non is necessary, and the vol-(' use of coolant removed from the system for startup from a cold condition is i about 3,700 ft 3. Even if both units were started simultaneously, or if one unit was started up twice in a short period of time, the maximum storage ca-pacity required would be 14,000 ft3. Thus, a 30 day holdup period can be pro-vided with about 22 per cent of the available storage capa:ity.
The remaining coolant renaved from the reactor system is the partial drain which occurs once.per year during refuelin6 The coolant is removed in a batch of 6,100 ft3 per unit and returned to the reactor coolant system upon comple-tion of refueling. Thus, it occupies storage capacity only during the period of refueling. The required storage volume for refuelins operations of 6,100 ft3 is less than 10 per cent of the available capacity.
It is extremely unlikely that operating conditions could occur which would re-quire simultaneous storage for all of the above liquid wastes. However, even if simultaneous storage were required, it could be accommodated by only two-thiras of the available storage capacity. This demonstrates that the six tanks provide adequate capacity to accommodate radicactive wastes as well as provid-ing extra capacity for liquid storage when dc41 red.
The storage facilities for miscellaneous wastes in the dual-unit system include the miscellaneous waste holdup tank - 2,700 ft3, the auxiliary building sump tank 4h0 ft3, and the two reactor building sumps - estimated at 1,000 ft3 each. Activity levels in the waste holdup tank were determined by assuming that all liquid collected in the tank was reactor coolant leakaSe containing the mv4== fission product activity (at the end of the second core cycle).
Collection was assumed to take place at the rate of 240 gpd from both units A for 60 days. At the end of this time the contents were discharged, without
() holdup, to the Keovee Hydro tailrace with a dilution flow of 1,100 cfs. The 11-13 2 4
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concentration at the point of discharge averaged over 60 days was 0.8 of the s MFC for unrestricted areas. Essentially all of this value is due to I-131 which has an eight day half-life. Thus, by using the auxiliary building or reactor building su=ps for collection, the activity collected in the miscellan-eous vaste holdup tank could be held up for decay prior to discharge to give lower effluent concentrations. For example, by use of these sumps and delay of discharge from the vaste holdup tank for 30 days, the average effluent concen-tration could be reduced to 0.06 times the MPC. If even longer holdup times are desired, one of the reactor coolant ble i holdup tanks can be employed for an additional 11,000 ft) of storage capacity for liquid wastes.
The abe- analysis demonatrates the station capability for handling large quan-tities af liquid wastes within allowable limits. For the reactor coolant bleed system, the purification demineralizers and the large system storage capacity provide ample means of collection and disposal for liquid wastes even in the remote case of 1 per cent fuel failure. Similarly, the miscellaneous vastes are shown to present no problem when analyzed on this conservative basis. It is concluded that the capacity of the liquid waste disposal system is large enough to permit vide flexibility in station operations while providing a means for safe disposal of vastes with activity well below the acceptable limits.
11.1.2 5 2 Gaseous Wastes In determining the activity concentra , ions in the gaseous effluent, the atmos-pherie dilution was computed using the Gifford model for wake release as de-scribed in 14.2.2 3 6. Concentrations were calculated at the one mile exclusion distance under the long term release conditions.
J The collection of gaseous activity was determine.d for those components repre-senting the maximum potential radiation hazard, including the reactor building, letdown storage tank, pressurizer, and reacto" coolant bleed holdup tanks.
The discharge of activity to the atmosphere as a result of reactor coolant bleed was determined for two situations: (1) contituous bleed over life, and (2) di-lution and expansion following shutdown and startup. For the case of continuous bleed all of the Kr, Xe, and I in the coolant letdown was assumed to co=e out in the void space of the reactor coolant bleed holdup tanks. The coolant activ-ity levels were those computed at the end of the second core cycle with 1 per cent failed fuel. Before reaching the reactor coolant blecd holdup tanks, the letdown flow was taken throu6h the demineralizert, assumin6 a 99 per cent re-moval efficiency for iodine. The activity was released to the atmosphere, without holdup, at a rate equal to the average shim bleed rate over life of 30 gph per unit. With both units releasing activity at this rate, the total frac-tion of the MFC (for unrestricted areas) at the exclusion distance is about 0.0h.
In the case of unit shutdown and startup, it was pestulated that a cold shut-down occurred at a time in lifetime just prior to beginning the use of the de-borating demineralizer for boric acid removal. This results in the maximum quantity of coolant bleed during startup. No coolant activity decay was as-sumed during the shutdown. As a result of this cperation a bleed quantity equivalent to 0.8 reactor system volu=es occurs from one ' unit. Ietdown through the demineralizers with a removal efficiency of 99 per cent for iodine was as-su=ed. As the coolant is let down to the bleed holdup tanks, all of the Kr, Xe, and I is assumed to come out of the water and go into the waste gas decay a?'
11-14 285
tank. With a design pressure of 150 psi and a volume of 1,350 ft), one vaste
( gas decay tank can hold the totd1 gas volume displaced by this quantity of coolant bleed. The approximately 7,000 ft3 of gas displaced from the bleed holdup tank would only pressurize the vaste gas decay tank to 76 psig. The gaseous activity could then be discharged over a period of one week to allow dispersion in accordance with the long term atmospheric diffusion model. The average yearly concentration at the exclusion distance, for one such operation per quarter, would be about 0.015 of the MFC for one unit. The average concen-tration for two uaits operating in this same manner would be 0.03 MPC.
The gaseous concentrations in the letdown storage tank void were determined from Henry's Law assuming the tank gas space is in equilibrium with the reac-tor coolant. The fraction of activity in the reactor coolant system which col-1ected in the letdown storage tank was approximately 45 per cent for Kr, 30 per cent for Xe, and 0.14 per cent for I. The activity levels used for sources in the letdown storage tank corresponded to the reactor coolant system activity at the end of the second core cycle. It is assumed that the tank in each Unit is vented, once a year, to the vaste The volume of gas in the letdown storage tank is about 300 ft) gas decay at 45 psia.tank.
This gas would only increase the vaste gas decay tank pressure 10 psi. This gas can be discharged to the atmosphere over a period of one week to ensure dispersion in accordance with the long-term atmospheric diffusion model. The average yearly concentration of activity at the exclusion distance is computed to be 0.004 of the MPC for both units.
w Calculations similar to those used for the letdown storage tank were performed to determine the activity in the pressurizer. It was found that the activity in the pressurizer was approximately one third the activity in the letdown storage tank. Venting of the pressurizer results in only about 60 ft3 of gas, which can be released from the waste gas decay tank over a period of one week to give a yearly average concentration of less than 0.002 of the MFC at the ex-clusion distance.
The activity level in the reactor building atmosphere was computed assuming a reactor coolant system leakage to the reactor building air of 10 gpd per unit.
All of the Kr and Xe, and 50 per cent of the I and Cs that leaked from the re-actor coolant system, were dispersed throughout the reactor building at=osphere.
Activity buildup in the reactor building was computed over the last 30 days of fuel leakage, i.e., it was assumed that no purge had been made for 30 days.
This quantity of activity was then discharged to the atmosphere, without decay, j by way of the reactor building purge system. The concentration at the exclusion !
distance averaged over 30 days was computed to be 0.003 of the MPC. Venting both reactor buildings once each 30 days would give an average yearly dose of 0.006 of the MPC.
This evaluation demonstrates that the total yearly average concentration of activity at the exclusion distance from all modes of release, including pres-surizer vent, reactor building purge, venting of the letdown storage tank, startup expansion and dilution, and chemical shim bleed,is a maxi =um of about 0.04 of the MFC for one unit. Even in the remote instance of 1 per cent fuel failure in both units concurrently, the average yearly concentration at the site boundary would be about 0.08 of the MPC. The evaluation also demonstrated that equipment capacities are adequate to accommodate and store radioactive x
iE' 11-15 --- -
m .
gases as necessary. Thus, the system design is adequate to insure safe dis-posal af gaseous wastes.
A preliminary analysis has been cade to examine the consequences of reactor operation with steam generator tube leakage end 1 per cent failed fuel rods.
The analysis considered the direct dose at various locations in the steam and condensate systems and also the activity release to the environment. The limit-ing concentration was established by the activity carried with the air ejector exhaust to_ the station vent to re=ain within the allowable dischar6e limit 3 of 10 CFR 20. At this limiting concentration, the direct dose rate from the con-denser is below the permissible value for continued access.
In the air ejector exhaust, the controlling isotope is xenon-133 The analysis assumed that the xenon passed directly from the reactor coolant system leak to the condenser with all the activity ultimately released to the station vent with no radioactive decay. With this conservative assumption, a reactor cool-ant leak rate of 1 gpm results in a concentration of 0.07 of the MPC at the ex-clusion distance. The analysis was based on 1 gpm tube leakage continuously over a year. Therefore, much h1 5 her tube leakages could be permitted for shorter periods of time, 11.1.253 Radioactive Waste Disposal System Failures The possibility of a significant activity release off the site from accidents in either the solid or the liquid waste disposal equipment is extremely re-mote. Both of these systems are located in shielded, controlled-access areas with provisions for =aintaining contaminntion control in the event of spills J or leakage. Solid wastes are disposed by licensed contractors in accordance with ICC regulations. Liquid wastes are sampled prior to discharge and are monitored during discharge to insure compliance with 10 CFR 20. A tabulation of potential waste disposal system failures and their consequences is presented in Table 11-6.
Radioactive gaser, are sa= pled and discharged in ecmpliance with the require-ments of 10 CFR 20. In the event of waste gas decay tank failure, these gases would be released to the decay tank compartment, and then released to the sta-tion vent via the nor=al ventilation system.
The maxi =um activity in a vaste gas decay tank will occur following a boron dilution cycle during reactor startup just prior to switching to deborating domineralizer for boron re= oval. The reactc.,r coolant water activity used for the analysis assumes prior operation for an extended period with failed fuel rods, equivalent to exposure of 1 per cent of the fuel. Approximately 0.8 of an equivalent reactor nolant system volume would be let down at this time.
It is assumed that the pur fication demineralizers have a removal factor of only 100 for iodine, although factors of 103 to 10 4have been reported in the literature. No removal of noble gases is assumed.
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Table 11-6 Waste Disposal System Failure Analysis Component Failure Coruments and Consequences Nonnal Reactor Building Sump Fails to close. Backup isolation is provided on opposite side of Drain Valve (inside or outside) reactor building.
Component Drain Pump Drain Valve Fails to open. Continuous drainage is not required; the valve (inside or outside) is . located for maintenance during operation.
Reactor Building Sump Pump Fails to operate. Continuous operation is not required; located for maintenance during operation.
Component Drcin Pump Fails to operate. Continuous operation is not required; located for maintenance during operation.
Quench Tank Vent Valve Fails to open. Continuous venting is not required; relief pro-tection is provided for tank.
Fails to close. Vent gas is conveyed to vaste gas decay tank and discharged through filters to station vent.
h n Waste Gas Vent Filters Rupture or lose High activity level monitored and alarmed if in-Q efficiency. sufficient Station vent dilution is available.
Q Waste gas is diverted to vaste gas decay tanks.
E
$. Waste Gas Decay Tanks Leak or Rupture. Building purged to station vent threugh filters.
e Tanks are protected by relief valves.
' ~.
& Reactor Coolant Bleed Holdup Tanks Leak. Leakage is collected in auxiliary building drain U sump for process or disposal; building is con-tinuously purged to statio vent.
{ Evaporator Train Fails to operate. Continuous operation is not required; waste gas yg decay tanks provide for vaste collection during CO g maintenance.
CO fj g Deborating Demineralizers Exhausted resin. Spare ur.it placed in service while original unit 87 is regenerated. Startup time is increased near a end-oI-life depending on balance between rod worth and boric acid required.
Q The remaining gaseous activity is carried with the water to the reactor cool-ant bleed holdup tanks, where it is assumed that the gases are immediately re-leased from the water and carried with~the purge gas to the waste gas decay tank. This assumption is quite conservative since the gas release rate will occur due to diffusion from the surface in accordance with Henry's Law and oc-cur over a considerable time period. Similarly, it is conservatively assumed that the gases do not undergo radioactive decay after leaving the reactor cool-ant system. With these assumptions, the following activity is calculated to exist in one of the waste gas decay tanks:
Isotope Total Curies Kr 85m 475 Kr 85 3,890 Kr 87 280 Kr 88 820 I 131 8.2 I 132 11 9 I 133 11.0 I 134 13 I 135 50 Xe 131m 540 Xe 133m 780 Xe 133 69,200 Q
V Xe 135m Xe 135 260 1,510 Xe 138 110 The area surrounding the waste gas decay tanks is ventilated and discharges to i
the station, vent. The discharge from the station vent is conservatively as-sumed to mix in the wake of the building structures rather than remain at its 4
elevated release point. This assumption produces less favorable dilution and
, therefore higher ground concentrations at the exclusion distance. Also, with this assumption, the doses at the exclusion distance are essentially the same whether or not the ventilation system is operating.
The activity from a vaste gas tank failure is assumed to be released as a puff fro = the station vent. Atmospheric dilution is calculated using the two hour
- meteorological model discussed in 14.2.236. The total integrated dose to the whole body at the one mile exclusion distance is 0 3 rem, and the thyroid dose at the same distance is 0.4 rem. These doses are well below the Guideline values of 10 CFR 100.
11.1 3 TESTS AND INSPECTIONS Functional operation 1 tests and inspections of the Waste Disposal System vill be cade as required to insure performance consistent with the requirements of 10 CFR 20.
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11.2 PADIATION PROTECTION 11.2.1 PRIMARY, SEONDARY, AND REACTOR BUILDING SHIEIDING 11.2.1.1 Design Bases The shielding is designed to perform two primary functions: (1) to insure that during nor:::a1 operation the radiation dose to operating personnel and to the general public is within the limits set forth in 10 CFR 20, and (2) to insure that operating personnel are adequately protected in the event of a reactor accident so that the accident can be terminated without undue hazard to the
~
general public. The shielding design is based on the two Oconee reactors op-erating at the maximum expected power level of 2,568 mwt each with system activ-ity levels equivalent to 1 per cent failed fuel, and is governed by the follow-ing criteria for radiation levels.
Location Dose Rate, mrem /hr Site Boundary 0.05 Office, Control Room, and Turbine-Generator 05 Normal Accessible Areas in Reactor Building during Operation at Ful.1 Power 25 Inside Control Room following MHA 3 rem integ ated whole body .J dose for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts over 90 days after accident.
11.2.1.2 Description 11.2.1.2.1 Primary Shield
- The primary shield consists of reinforced concrete which surrounds the reactor vessel and extends upward from the reactor building floor to form the walls of the fuel transfer canal. The preliminary shield thickness is 5 ft up to the height of the reactor vessel flange, where the thickness is reduced to 4.5 ft.
The primary shield is designed to meet the following objectives:
- a. To attenuate the neutron flux to limit the activation of component and structural materials.
- b. To limit the radiation level after ehutdown to permit access to the reactor coolant system equipment. j
- c. To reduce, in conjunction with the secondary shield, the radiation l 1evel from sources within the reactor vessel to allow access to the l reactor buiMing during nonnal full power operation.
Oi 11-19 290
m 11.2.1.2.2 Secondary Shield The secondary shield is composed of reinforced concrete which surrounds end eevers the reactor coolant equipment, including the piping, pumps, and steam generators. The preliminary thickness of structure which surrounds the cool-ant system is 4.pe ft, Oad t" W im4"a .
/ thic!=;;; cf cover ever4ha-system 14-G-ft. The shielding is designed to reduce radiation levels from activity in the reactor coolant, and to supplement the primary shield in the attenua-tion of neutrons and secondary gamma rays so as to permit limited access to the reactor building during full power operation.
11.2.1.2 3 Reactor Building Shield The re.actcr buildin6 shield is a reinforced,,ft-thick dome.-ft-thick with.31*j3 cylindrical walls and a4 In conjunction with the primary and secondary shields, it will limit the radiation level outside the reactor building from all sources inside the reactor building to no more than 0 5 mrem /hr at full power operation. The shielding is also de-signed to protect station personnel from radiation sources inside the reactor building followind the Mwim" Hypothetical Accident. Additional shielding is provided around the control room to insure,that exposure to operating personnel in the control room is within the design limits following the MHA.
I 11.2.1.2.4 Materials p The material used for the primary, secondary, and reactor building shields is '
() ordinary concrete with a density of approximately 140 lbs/ft 3 . Since the pri-mary and secondary shielding walls serve as the refueling structure, give sup-port for the reactor coolant components under pipe rupture conditions, and provide missile shielding, they are reinforced and designed to be self-support-ing. The concrete primary shielding immediately surrounding the reactor vessel will require supplemental cooling to remove internal gamma heat. This is ac-comp 11shed by water-cooled coils embedded in the concrete in this area.
11.2.1 3 Evaluation 11.2.131 Padiation Sources Gamma-ray yield and spectral distributions from prompt fission and gross fis-sion product activity are based on the information in Volume III, Part B, of the Reactor Handbook. The yield and spectral data for capture gammas are taken from ANL-5800, Reactor Physics Constants, and the Reactor Handbook.
Data on activation product gamma rays are derived primarily from the Review of Modern Physics, Vol 30, No. 2, April 1958. The production of N-16 in the reactor coolant is calculated with a code by The Babcock & Wilcox Company which computes the integral of the 0-16 (n,p) N-16 cross section over the neu-tron flux in a water-cooled reactor, subject to variables in coolant flow and density and in neutron flux spectra and magnitude. The 0-16(n,p) N-16 cross section used is that reported in WAPD-BT-25 Activities of individual fission products in the core, reactor, coolant, and reactor auxiliary systems are de-termined by a B&W computer code designed to predict activities from a five-member radioactive chain at any point in the core history. Fission product p)
(O leakage from the core to the coolant, and removal from the coolant by purifi-cation and leakage, are calculated.
s
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11.2.132 Calculation Methods s Neutron penetration in shield regions is calcu3ated using the B&W LIFEX code as a coefficient generator to provide input data into either the TOPIC code or FJ 3T code. TOPIC (ID0-16968) and MIST (IDO-16856) are programs which solve the transport equation using the Carlson Sn method in cylindrical and slab geometries respectively, and are used to generate 4-group fluxes in the radial and axial directions from the core.
Gamma ray attenuation is calculated using the Taylor exponential form of build-up with the ga=ca source strengths divided into 1 mev energy intervals between 1 and 10 mev. The equations for the flux from the simpler geometric sources (line, disc, truncated cone, and cylinder) are solved by the B&W Basic Geometry Code. Equations and data for the generation and attenuation of se.condary gan:ma rays in a inminated, semi-infinite shield array are contained in the B&W Secondary Ga=ma Program. For the more complex source-shield configurations where nonuniform source distributions may exist, BONGO, a B&W kernel integration code, is used. The program utilizes a point kernel attenuation along a line-of-sight from the source point to the dose point, and computes the flux by suming over the source distribution. A description of the afore-mentioned B&W codes and techniques can be found in ILO-24467 11.2.133 IEA Dose Calculation The thickness of the reactor building shielding, in accordance with the design dose rate criteria, is based upon radiation levels due to fission product re-lease following a reactor accident. For the calculations it was assumed that ,
100 per cent of the gases, 50 per cent of the halogens, and 1 per cent of the J solid fission products were instantaneously released to the reactor building following a buildup period in the core of 600 full power (2,568 cut) days.
The fission product activity was assumed to be uniformly dispersed throughout the reactor building volume, and the reactor building was represented by a cy:.indrical source for the dose calculations. The integrated dose over vari-ous time intervals was computed as a function of distance from the reactor building. The results are given in 14.2.2.4.
31.2.134 Operating Limits All parameters governing the shield design, including heating and dose rate profiles, temperature distributions, and coolant flow requirements, will be performed during the detailed design of the station.
11.2.135- Tcdiation Surveys neutron and ga==a radiation surveys will be performed in all accessible areas of the station as required to determine shielding integrity. Plans and pro-cedures for radiation surveys during operation and following shutdown will be formulated during the detailed station design. -
a c' l 11-21 M2
s, 11.2.2 AIEA 1%DIATION MONITORING SYSTH4 11.2.2.1 Design Bases The area radiation monitoring system will be designed to indicate and alarm high radiation levels inside the station. Indication from the beta-gamma de-tectors located in selected areas of the station will be used in conjunction with operating procedures to assure that personnel exposure does not exceed 10 CFR 20 limits.- The control room and Auxiliary Building ventilation sys-tems will be monitored.
11.2.2.2 Description Beta-gamma detectors are located as follows:
- a. One detector on each of the fuel handling bridges inside the Reactor Building,
- b. Inside the Reactor Building near the personnel access hatch.
- c. Fesr incore instrument space inside the Reactor Building.
- d. Fuel handling bridge in Auxiliary Building.
- e. Auxiliary Building pump area.
- f. Auxiliary Building near sample sink.
- g. Auxiliary Building cask decontamination and loading area.
- h. Auxiliary Building in shutdown cooler area.
- i. Auxiliary Building near Reactor Building component cooling water coolers.
J. Chemistry laboratory.
- k. Cable and computer room.
- 1. Centaminated machine shop.
- m. Control room.
Readout for each detector will be provided in the control room. High radia-tion alarm signals for each detector will be furnished to the control room and to each remote detector location. Sources will be available to allow the overall system performance to be verified at regular intervals.
Detector ranges will be determined depending upon the normal back6 round at the detector locations and the expe'cted radiation levels for abnormal conditions.
f%
E s 11-22 -
11 2.2 3 _Ev_aluation h Area radiation monitor detectors vill be located on each of the fuel handling bridges to warn personnel if a high radiation level is approached during re-fueling operations.
A vide range detector vill be mounted near the access hatch of the Reactor Building to indicate radiation levels inside the hatch before it is opened.
The upper range of the detector will be sufficiently high to indicate the ac-cessibility of the Reactor Building followin6 a serious accident inside.
The incore instru::ent area vill be monitored, and a local alam vill be pro-vided to varn if a high radiation level exists or is created while incore as-se=blies are being manipulated.
The sample sink area in the Auxiliary Building win be equipped with a detec-tor to alam an abnor=al condition in connection with system sampling.
Alams vill be actuated in the control room and at the detectors if an abnor-mal change in radiation background occurs.
11.2 3 HEALTH PHYSICS The station superintendent is responsible for radiation protection and contami-nation control for Oconee. This responsibility is, in turn, shared by all supervisors. All personnel assigned to the station and all visitors vill be required to fonov rules and procedures established by administrative control ;
for protection against radiation and contamination. .)
The administration of the radiation protection program vill be the responsibil-ity of the station Health Physicist. It vill be the responsibility of the Health Physics section to train station personnel in radiation safety; to lo-cate, measure and evaluate radiological proble=s; and to make recommendations for control or elimination of radiation hazards. The Health Physics section vill function in an advisory capacity to assist all personnel in carrying out their radiation safety responsibilities and to audit all aspects of station operation and maintenance to assure safe conditions and compliance with AEC and other federal and state regulations concerning radiation protection.
Administrative controls vill be established to assure that all procedures and requirements relating to radiation protection are followed by all station per-sonnel. The procedures that control radiation exposure vill be subject to the same review and approval as those that Sovern all other station procedures (see w etion 12 5, Ad=inistrative Control). These procedures win include a Radiation Work Pemit system. A n work on systems or locations where exposure to radiation or radioactive materials is or may be involved vill require an appropriate Radiation Wor'k Permit initiated by Health Physics and approved by cognizant supervisors before work can begin. The radiological hazards associ-ated with the job vill be determined and evaluated prior to issuing the permit.
The work permit vill list the precautions to be taken, the protective clothing to be vorn and any other radiation control and safety precautions that =ay be required.
O 294 n-23 -
c 11.2 3 1 Personnel l'onitoring Syste=s Personnel monitoring equip =ent consisting of film badges or their equivalent vill be assigned by the Health Physics section and vorn by all personnal at Oconee. In addition, those persons who ordinarily work in restricted areas or whose job requires frequent access to these areas vill have pocket cha=bers, self-reading dosi=eters, pocket high radiation alar =s, vrist badges and finger tabs readily available for use, when required by station conditions. 'This per-sonnel monitoring equip =ent will also be available on a day-to-day basis for those persons, e=ployees, or visitors not assigned to the station who have oc-casion to enter. Restricted Areas or to perform work involving possible exposure to radiation. Recorqs of radiation exposure history and current occupational ,
exposure vill be maintained by the Health Physics section for each individual for whoc personnel conitoring is required. The external radiation dose to per-sonnel vill be determined on a daily and/or weekly basis, as required, by means of the pocket cha=ber and dos 1=eter. Film bad 2es will be processed conthly or more frequently when conditions indicate it is necessary.
11.2 3 2 Personnel Protective Equipment Special " protective" or "anticontamination" clothing vill be furnished and worn as necessary to protect personnel against contact with radioactive contamina-tion. Change Roems will be conveniently located for proper utilization of this protective clothing. Respiratory protective equipment vill also be available for the protection of personnel against airbo n e radioactive contamination and vill consist of full face filter masks, self-contained air-breathing units, or O,, air-supplied casks and hoods. The first line of defense against airborne con-tamination in the work area is the ventilation syste=. However, respiratory protective equip =ent will be provided should its use become necessary.
Maintenance of the above equip =ent will be in accordance with the manufacturer's recommendations and rules of good practice such as those published by the Amer-ican Industrial Hygiene Association in its " Respiratory Protective Devices Manual". The use and maintenance of this equip =ent will be under the direct control of the Health Physics section, and personnel vill be trained in the use of this equip =ent before using it in the perfor=ance of work.
11.2 3 3 Change Room Facilities Change room facilities will be provided where personnel may obtain clean pro-tective clothing required for station work. These facilities will be divided into "c ean" and " contaminated" sections. The " contaminated" section of the change rooms will be used for the re= oval and handling of contaminated protec- ,
tive clothing after use. Showers, sinks, and necessary monitoring equipment '
also will be provided in the change areas to aid in the decontamination of per-sonnel.
Equipment decontamination facilities vill also be provided at the station for large and s=all ite=s of plant equipment and components.;
i Provision vill also be cade for deconta=ination of work areas throughout the i station.
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295 11-24 i
Appropriate written procedures vin govern theproper use of protective clothing, 3 where and how it is to be vorn and removed, and how the change room and decon-tamination facilities for personnel, equipment, and station areas are to be used.
In order to protect personnel from access to high radiation areas that =ay exist temporarily or semiper=anently as a result of station operations and caintenance, varning signs, audible and visual indicators, barricades, and locked doors will be used as necessary. Administrative procedures vill also be written to control access to high radiation areas. The Radiation Work Permit System vill also be utilized to control access to high radiation areas, 11.2 3 4 Health Physics Laboratory Facilities The station vill include a Health Physics Laboratory with facilities and equip-ment for detecting, analyzin6, and measuring all types of radiation and for eval-uating any radiological problem which =ay be anticipated. Counting equip =ent (such as G-M, scintillation, and proportional counters) vill be provided in an appropriate shielded counting roo= for detecting and ceasuring all types of radi-ation as well as equipment (such as a multichannel analyzer) for the identifica-tion of specific radionuclides. Equip =ent and facilities for analyzing en'viron-
= ental survey and bicassay samples vill also be included in the Health Physics Laboratory. Maintenance and use of the Health Physics Laboratory facilities and equip =ent will be the responsibility of the Health Physics section.
11.2 3 5 Health Physics Instru=entation Portable radiation survey instruments vill be provided for use by the Health Physics section as voll as for operating and maintenance personnel. A variety -
of instru=ents vill be selected to cover the entire spectrum of radiation =ea-sure=ent proble=s anticipated at Oconee. Sufficient quantities vill be ob-tained to allow for use, calibration, =aintenance, and repair. This vill in-clude instruments for detecting and measuring alpha, beta, ga==a, and neutron radiation. In addition to the portable radiation =onitoring instruments,, fixed monitoring instruments, i.e., count rate meters, vill be located at exits frcm restricted areas. These instru=ents are intended to prevent any contamination on personnel, material, or equipment from being spread into unrestricted areas.
Appropriate monitoring instruments vill also be available at various locations within the restricted areas for contamination control purposes. Portal monitors vill also be utilized, as appropriate, to control personnel egress from restrict-ed areas or from the station.
The station vill have a per=r.;ntly installed re=ote radiation and radioactivity monitoring system for locations where significant levels can be expected. This system vill monitor airborne particulate and gaseous radioactivity as well as' external radiation levels. This system vill present an audible alarm and radia-tion level indication in the area of concern in addition to the control room.
11.2 3 6 Medical Programs Facilities and counting equipment for screening personnel for internal exposure l vill be available on site with outside services utilized as backup and support for this program.
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A comprehensive medical examination program appropriate for radiation workers will be conducted to establish and maintain records of the physical status of each employee at Oconee. Subsequent medical examinations will be held as de-termined necessary for radiation worke:s. Medical doctors, preferably in the local area, vill be used for this program. The Health Physics section will be responsible for the program and vill assist the physicians in =aintaining medical control of personnel. This program will be designed to preserve the health of the employees concerned and to confirm the radiation control methods employed at the station.
11 3 REFERENCES
' (1) Frank, P. W., et al., Radiochemistry of Third PWR Fuel Material Test -
X-1 Icop NRX Reactor, WAPD-TM-29, February 1957 (2) Eichenberg, J. D., et al., Effects of Irradiation on Bulk 00 '
WAPD-183, October 1957 2 (3) Allison, G. M. and Robertson, R. F. S., The Behavior of Fission Products in Pressurized-Water Systems. A Review of Defect Tests on U0 2 Fuel Elements at Chalk River, AECL-1338, 1961.
(4) Allison, G. M. and Roe, H. K., The Release of Fission Gases & Iodines From Defected 00 2 Fuel Elements of Different Lengths, AECL-2206, June 1965 O
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.._.___________________________._____.._-.__---___.... OCONEE NUCLEAR STATION FIGURE 11-1 299 ,
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