ML19319D688
| ML19319D688 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, 05000303  |
| Issue date: | 08/10/1967 |
| From: | FLORIDA POWER CORP. |
| To: | |
| References | |
| NUDOCS 8003240669 | |
| Download: ML19319D688 (37) | |
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TABLE OF CONTENTS Section Pye,e 11 RADIOACTIVE WASTE AND RADIATION PROTECTION 11-1 11.1 RADIOACTIVE WASTES 11-1 11.1.1 DESIGN BASES 11-1 11.1.1.1 Performance Objectives 11-1 11.1.1.2 Radioactive Waste cuantities 11-1 11.1.1.3 Waste Activity 11-1 11.1.1.h Disposal Methods 11-2 11.1.1.5 Shielding 11-3 11.1.2 SYSTEM DESIGN 11-3 11.1.2.1 Liould Waste Disposal System 11-3 11.1.2.2 Solids Waste Disposal System 11-5 11.1.2.3 Gaseous Waste Disposal System 11-5 11.1.2.h Process System Rad (i(tion Monitoring 11-5 11.1.2.5 Design Evaluation 11-6 11.1.2.5.1 Liquid Waste $ 11-6 11.1.2.5.2 Gaseous Wastes l'l-8 11.1.2.5.3 Radioactive Waste Disposal System Ftilures 11-10 11.1.3 TESTS AND INSPECTIONS 11-12 11.2 RADIATION SHIELDING 11-12 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING,.JD AUXILIARY SHIELDING 11-12 11.2.1.1 Design Criteria 11-12 11.2.1.2 Descrintion of Shielding 11-13 l
11-1 I
TABLE OF CONTENTS
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Section Page, 11.2.1.2.1 Primary Shield 11-13 11.2.1.2.2 Secondary Shield 11-13 11.2.1.2.3 Reactor Building Shield 11-13 11.2.1.2.h Control Room Shield 11-14 11.2.1.2.5 Auxiliary Shield 11-lh 11.2.1.2.6 Spent Fuel Shielding 11-lh 11.2.1.2.7 Materials and Structural Requirements 11-15 11.2.1.3 Evaluation 11-15 11.2.1.3.1 Radiation Sources 11-15 11.2.1.3.2 Calculation Methods 11-16 11.2.1.3.3 MHA Dose Calculation 11-16 11.2.1.3.h operating Limits 11-16 g-)s,
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11.2.1.3.5 Radiation Surveys 11-16 11.2.2 AREA RADIATION MONITCHING SYSTEM 11-17 11.2.2.1 Design Bases 11-17 11.2.2.2 Description 11-17 11.2.2.3 Evaluation 11-18 11.2.3 HEALTH PHYSICS 11-18 11.2.3.1 Personnel Monitoring System 11-20 11.2.3.2 Personnel Protective Equipment 11-20 11.2.3.3 Change Room Facilities 11-20 11.2.3.h Health Physics Laboratory Facilities 11-21 11.2.3.5 Health Physics Instrumentation 11-21 11.2.3.6 Medical Programs 11-22
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TABLE OF CONTENIS Section Page 11.3 BEFERENCES 11-22 1
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O 0307 11-111
LIST OF TABLES 1o . . . v. . .- . . .
(At Rear of Section)
Table No. Title Pajyt 11-1 Radioactive Waste Quantities 11-23 11-2 Escape Rate Coefficients for Fission Product Release 11-25 11-3 Reactor Coolant Activities For a Unit Containing One l
Per Cent Defective Fuel 11-25
! ll-k Waste Disposal System Component Data 11-26
- 11-5 Maximum Activity Concentrations in the Plant Effluent With One Per Cent Failed Fuel 11-29 11 Waste Disposal System Failure Analysis 11-30 4
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LIST OF FIGURES (At Rear of Section)
Figure No. Tith 11-1 Waste Disposal System O
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11-v 0309
n 11 RADIOACTIVE WASTE AND RADIATION PROTECTION
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11.1 RADIOACTIVE WASTES 11.1.1 DESIGN BASES 11.1.1.1 Performance Objectives The vaste disposal system vill be designed to provide controlled handling and disposal of liquid, gaseous, and solid vastes which will be generated during plant operation. The design criteria are to insure that plant personnel and the general public are protected against excessive exposure to radiation from vastes in accordance with limits defined in 10 CFR 20.
11.1.1.2 Radioactive Waste Quantities The estimated volumes of radioactive vastes generated during plant opera-tion are listed in Table 11-1.
11.1.1.3 Waste Activity Activity accumulation in the reactor coolant system and associated vaste 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 assuming full power op. ration of 2,5hh MWt for one core cycle with no defective fuel followed by operation over the second core cycle with
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1 per cent defective fuel. Continuous reactor coolant purification at a rate of one reactor system volume per day was used with a zero removal efficiency for Kr, Cs, and Xe. and a 99 per cent removal efficiency for all other nuclides. Activity levels are relatively insensitive to small changes in demineralizer efficiencies, e.g. , use of 90 per cent instead of 99 per cent vould 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 best i on the use of " escape rate coefficients" (sec -1) as determined from experiments involving purposely defected fuel elements (References 1, 2, 3, 4). Values of the escape rate coefficients used in the calculations are shown in Table 11-2.
Calculations of the activity released from the fuel were performed with a digital computer code which 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 holdup tank. The activity levels in the reactor coolant for a unit containing one per cent defective fuel during full power operation at the end of the second core cycle are shown in Table 11-3.
The liquid vaste generated by leakage, sampling, and demineralizer sluice or rinse is assumed to have an activity concentration equal to the con-centration in the reacter coolant. Reactor coolant bleed vill be taken n
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from the downstream side of the purification demineralizer. It is assumed to have the same activity concentration as the reactor coolant reduced by the decontamination factor of the purification demineralizer. Laundry and shover vastes are assumed to contain negligible amounts of radioactivity.
Gaseous activity will be generated by the evolution of radioactive gases from liquids stored in tanks throughout the plant. These include such items as reactor coolant bleed tanks, miscellaneous vaste tanks, and the makeup tank which are vented to the vaste gas disposal system. The activity of the gases ic dependent upon the liquid activity. The assumptions for liquid activity are described above. The resulting gaseous activities are described in Section 11.1.2.5, " Design Evaluation."
11.1.1.4 Disposal Methods Liquid vastes from the plant will be handled in two separate streams using two evaporator chains. Reactor coolant bleed will be fed through one chain; and miscellaneous vastes, which include reactor building sump drains, reactor coolant drains, equipment drains, and floor drains, vill be processed through the other chain. The treatment of the liquid vastes vill be in one of the following ways:
- a. Reactor Coolant Bleed
- 1. Collected, demineralized, monitored, and returned to makeup tank.
- 2. Collected, monitored, concentrated, and discharged to vaste dru= ming area for packaging and off-site disposal.
- 3. Condensate resulting from the concentration operation vill either be reclaimed as demineralized water or discharged with the condenser circulating water.
- b. Miscellaneous Wastes
- 1. Collected, monitored, and discharged with the condenser circulating water.
- 2. Collected, monitored, held up for decay, then discharged with the condenser circulating water.
- 3. Collected, monitored, concentrated, packaged, and shipped off-site.
- h. Condensate resulting from the concentration operation vill be either reclaimed as demineralized water or discharged with the condenser circulating water.
Gaseous vastes are disposed of using one of two methods:
- a. Continuous dilution and discharge through vaste gas filters to the plant vent when activity levels permit.
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- b. Diversion to vaste gas holdup tanks with sampling and controlled
,- subsequent release through vaste gas filters to the plant vent.
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Solid radioactive vastes will be accumulated and packaged in special drums suitable for Interstate Com=erce Commission (ICC)-approved shipment off-site to a licensed waste disposal facility.
11.1.1 5 Shieldine Shielding for the components 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 reactor coolant drain tank and the reactor building sumps, all components will be located in the auxiliary building. The shield design criteria for the auxiliary building is a dose rate from 1 to 5 mrem /hr in controlled areas and 15 mrem /hr in areas requiring limited access. The components of the vaste disposal system will be shielded by concrete walls and floors of varying thicknesses depending on the mssnitudes of the sources in each component and on the access requirements in a particular area.
In some areas local shielding in the form of lead or removable concrete blocks will be utilized to facilitate maintenance or repair operations.
11.1.2 SYSTEM DESIGN 11.1.2.1 Liould Waste Disposal System Liquid waste handling will be divided into two separate vaste processing chains. One chain will process the reactor coolant bleed stream and reactor
,, coolant drains, and the other will handle all miscellaneous liquid wastes.
(
) The conceptual system flow diagram is shown in Figure 11-1. Waste disposal system component data is given in Table 11-h.
Reactor coolant will be received from the make'.g and purification system and will be the largest single source of operational liquid waste to be handled. This liquid will be received as a result of reactor coolant expansion and operational requirements for reduction of reactor coolant boric acid content. It vill be either conveyed to reactor coolant bleed holdup tanks for storage or passed through deborating demineralizers for boric acid removal and returned as unborated enkeup to the makeup and purifi-cation system. The deborating demineralizers will be used only for boric acid concentrations below 1000 ppm to limit the rate at which resins are used up.
The reactor coolant bleed holdup tanks will be sized to contain one reactor coolant system volume each. The contents of each tank will be periodically campled to determine their radioactive content. These tanks will feed the vaste batch tank which in turn supplies the waste evaporator or concen-trator. The contents of the waste batch tank will be pumped continuously through the evaporator using the evapor stor feed pumps and returned to the batch tank in a closed loop so that the wastes in the loop become pro-gressively concentrated. When vastes are sufficiently concentrated, the 11-3 0si2
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evaporator feed pu=ps vill be shut down, and the concentrated vastes in the evaporator vill automatically drain to the vaste batch tank. Residual vastes remaining in the evaporator vill be flushed out by returning small amounts of condensate through the evaporator body. This backwashing vill ensure a relatively clean evaporator shell and tube bundle with a minimum radiation hazard following operation of the unit. The concentrated vastes in the batch tank vill be sa= pled to determine their content. The vastes may then be disposed of by pumping to the vaste dru= ming area. The evapor-ator condensates vill be collected in a condensate test tank where they are sampled to determine quality and activity level. Condensates vill then be pumped through a condensate demineralizer to the evaporator conden-sate storage tanks for re-use or mixed with the condenser circulating water discharge. Gaseous vastes will be removed continuously from both the vaste batch tank and evaporster using vacuum pumps in the vaste gas line and passed through moisture separators into the vaste gas holdup tank. In the vaste gas holdup tank, gases are monitored for activity, held-up for decay as required, and then released at a controlled rate through the plant vent. A monitor located in the gaseous discharce line to the plant vent will be equipped with an indicator, and alarm to an- 1 nunciate a high activity level. The high level alans actuates an inter-lock to stop the discharge of gaseous effluents from the vaste gas system.
The second evaporator chain vill process liquid vastes collected by the miscellaneous vaste tank and auxiliary building sump tank. The liq; 3 wastes will be pumped into a neutralizing tank where the pH of the solu-tion is adjusted as necessary to prevent fosming and samples are taken to determine activity. Vastes will then be transferred to a vaste batch tank for cycling through an evaporator. When vastes are sufficiently concentrated, the concentrates vill be collected in the vaste batch tank and subsequently pumped to the vaste dru= ming area for packaging and disposal. Condensate vill be collected in a condensate test tank, sam-pled for activity, and subsequently either re-used as demineralized water or discharged at a controlled rate to the condenser circulating water. Gaseous vastes vill be removed by vacuum pumps from the evaporator and passed to the vaste gas decay tank for ultimate release through the plant vent.
Both evaporato higher than 10{ chains and willwill bebe designed sized to givevastes to process decontamination at e rate wellfactors in excess of the expected vaste accumulation rate.
As indicated above, all liquid va ste vill be sampled and analyzed for radioactive concentration prior to disposal. If discharge directly to the environment is permissible, a flow indicator and appropriate valving vill permit controlled release. The flow rate and activity of all liquids discharged from the vaste disposal system vill be indicated and alarmed.
The high activity alarm will actuate an interlock to stop the discharge in the event of excessive activity release.
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11.1.2.2 Solids Waste Disposal System 7-s . ,, . . , , , .w . .
Solid wastes will be placed in ICC-approved containers for the vaste material. Loaded containers will be monitored for surface radiation levels and stored in a shielding area prict to shipment to an off-site disposal facility.
Evaporator concentrate from the evaporator will be pumped into a shipping cask for off-site disposal. Spent resins from the demineralizers will be sluiced to a spent resin storage tank, and the sluice water vill be transferred fro = the tank to the miscellaneous vaste holdup tank. The spent resin storage tank vill hold one complete charge of res' ins from the reactor auxiliary systems. Spent resin vill be transferred from the storage tank to special drums for disposal. The activity of the spent resins and the shielding capability of the drum and shipping cask will determine the mixture propor-tions in the drum. Othe. miscellanecus solid vastes such as filters, clothing, laboratory equipment, pieces of metal, and paper vill be disposed of using a baler and light =etal shipping containers, 11.1.2 3 Gaseous Waste Disposal System Gaseous vastes i ll be removed from both evaporator chains during the con-centration operation. Vacuum pumps vill maintain a constant vacuum in the batch tanks and evaporators and will draw off all vaste gases to the gas decay tanks via the vaste gas compressors and moisture separators.
Gaseous vastes in the decay tank vill be monitored for activity and held up for decay and then released at a predetermined calculated rate through HEPA and charcoal filters to the plant vent. A high activity alarm and 1 indicator, will be located in the discharge line and provide automatic shut off of releases at a preselected level.
11.1.2.4 Process System Radiation Monitoring The cooling water systems that remove heat from potentially radioactive 1 sources will be monitored to detect accidental releases. A radiation monitor vill be located in the intermediate cooling leop, the nuclear services closed cooling loop, the nuclear servicea discharge header, the spent fuel cooling loop, and in the liquid waste discharge header.
Reactor coolant letdown flow will be monitored to detect a gross fuel assembly failure. A s= aller fuel assemMy leak will be detected by regular laboratory analysis of reactor coolant samples.
- Air nample from the reactor buildings and the plant vent vill be monitored for air particulate, gaseous , and iodine activity.
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,11-5 (Revised 1-15-68)
These radiation monitors are commercially available equipment. The required characteristics will be established daring detailed plant design. The minimum sensitivity of detectors, when combined with appropriate dilution ll factors , vill insure safe limits of release.
11.1.2.5 Design Evaluation All analyses on liquid and gaseous vaste disposal vill be performed on the basis of operation with 1 per cent failed fuel. Although it is not expected that the number of clad defects vill ever approach 1 per cent of the total fuel, the objective is to demonstrate the capability of safe plant operm. tion within the limits of 10 CFR 20 with quantities of radio-active fission products in the system.
A summary of the various operations considered in the analysis , and the total concentrations resulting in the plant effluent from operation of the unit with failtl fuel, are given in Table 1.~-5 The activity concen-trations resulting are given as fracticns of the Maximum Permissible Con-centrations (MPC) for unrestricted areas, i.e. , the concentration of each radioactivt nuclide has been divided by its respective MPC for discharge into unrestricted areas as set forth in 10 CFR 20.
11.1.2.5.1 Liquid Wastes The first operation considered in the analysis was the release of liquid vastes corresponding to continuous chemical shir. bleed. The average coolant bleed rate over a core cycle vill be approximately 25 gph, which l5 corresponds to letdown to storage for boric acid reduction. The activity level in the plant effluent was determined by assu=ing that reactor coolant system liquid was processed through the evaporator chain, and the condencates vere discharged continuously for a period of 278 full power days. Letdove chrouch the purification demineralizer was assumed to give a de;ontamination factor of zero for cesium, and 100 for all other nuclides. Activity levels in the reactor coolant system were those at the end of the second core cycle. Evaporator condensates were discharged to the condenser circulating sea water at 25 gph, and no holdup or decay was assumed. The dilution flow 5 was the condenser circulating water flow of about 2000 cfs. The results for a yearly average concentration were several orders of magnitude less than the MPC for discha26e into unrestricted areas as shown in Appendix 2D.
In the above analysis, no credit was taken for the further dilution of the liquid vastes by the environmental sea waters. Tnerefora, it is concluded that a large margin of safety will exist in the concentration of the plant l'iquid vaste discharges even on a continuous release basis.
Three reactor coolant bleed holdup tanks, each with a capacity of 82,500 l5 gul. will be provided for a total storage capacity of 247,500 gal. The maximum quantity of coolant lecdown for chemical shin, during any 30 day period in life, will be approximately h5,000 gal. Thus, only one tank, or one-third of the available storage capacity is required to provide a 30 day holdup period for the coolant which will be bled down over 30 days. 1 The maximum quantity of coolant removed during heatup and dilution to 11-6 (Revised k-8-68) f)_
e startup from a cold shutdown will be 105,000 gal. This occurs at the end d . .of the. chemical shi:operiod . Two cold .startups atsthise stage 'vould generate.. w a -
210,000 gal. of vaste. Earlier in life the quantity removed would be less than this due to the smaller amount of dilution required. Two cold startups, early in life, vill contribute about 23,500 gal of liquid vastes.
The remaining coolant removed from the reactor system is the partial drain which occurs once per year during refueling. The coolant is removed in a batch of 45,000 gal. and returned to the reactor coolant system upon completion of refueling. Thus, it occupies about 20 per cent of available storage capacity only during the period of refueling.
It is extremely unlikely that operating conditions could occur which would require storage for excessive amounts of liquid vastes. However, even in the event of two cold startups toward the end of core life, the available 5 storage capacity would accommodate the liquid vastes. This demonstrates that the three tanks will provide adequate capacity to accommodate all anticipated radioactive vastes as well as providing extra capacity for liquid storage when desired.
The storage facilities for miscellaneous vastes will include the miscellaneous vaste holdup tank (20,000 gal ) and the auxiliary building sump tank (3,500 l5 gal). Activity levels in the vaste holdup tank were determined by assuming that all liquid collected in the tank was reactor coolant leakage. Collec-tion was assumed to take place at the rate of 120 gpd for 60 days. At the l5 end of this time, the contents were processed through the evaporator chain, and the condensates were discharged to the condenser circulating water with a dilution flow of 2000 cfs. The concentration at the point of discharge, b
U averaged over the year, was significantly less than the MPC for unrestricted areas. This concentration vill normally be far lover, since it is intended to reuse the evaporator condensates as a demineralized water supply.
In addition to the above analyses, the effect of an inadvertent liquid vaste slug release of 450 gal. from a vaste condensate storage tank was examined. This release into the plant discharge was ar.umed to occur over a period of one hour. The release produces a concentration of only 0.003 of the MPC in drinking water in the condenser circulating water discharge.
The reactor coolant and miscellaneous vaste hndling systems described above vill adequately process the anticipated quantities of liquid vastes.
In the reactor coolant bleed system, the purification demineralizers and the large system storage capacity will provide ample means of collection and disposal for liquid vastes, 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 vaste disposal system vill be large enough to permit vide flexibility in plant operations, while providing a means for safe disposal of vastes with activity well below the acceptable limits.
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i0316 11-7 (Revised k-8-68)
11.1.2.5.2 Gaseous Wastes In determining the activity concentrations in the gaseous effluent, the atmospheric dilution was co=puted using the model for release as described in Section 2.3. Concentrations were calculated at the site boundary under the long term release conditions and also on-site concentrations around the adjacent fc'sil units were determined. l5 The collection of gaseous activity was determined for those components l5 representing the major sources of gaseous release--including the reactor building, makeup tank, pressurizer, and reactor coolant bleed holdup tanks.
The discharge of activity to the atmosphere as a result of reactor coolant bleed was determined for two situations: (1) continuous bleed over life, and (2) dilusion 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 come out in the void space of the reactor coolant bleed holdup tanks and go into the vaste gas decay tanks. The coolant activity levels were those co=puted at the end of the second core cycle with 1 percent failed fuel. Before reaching the reactor coolant bleed holdup tanks, the letdown flow was taken through the demineralizers assuming a 99 percent removal 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 25 gph. Releasing activity at this rate, the 5 total fraction of the MPC at the site boundary is .06. With a 30 day holdup in the vaste gas decay tanks, the concentration is reduced to 1 about .002 of MPC.
In the case of unit shutdown and start-up, it was postulated that a cold shutdovn occurred at a time in lifetime just prior to beginning the use of the deborating demineralizer for boric acid removal. This results in the maximum quantity of coolant bleed during shutdown. As a result of this operation, a bleed quantity of 72,000 gal. is produced. Letdown through the de.mineralizers with a removal efficiency of 99 percent for iodine was asaused. 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 vaste gas decay tank. With a design pressure of 150 psi and a volume of 1500 ft3, the vaste decay tank can hold the total gas volume displaced by this quantity of coelant bleed. The gas displaced from the bleed holdup tank, approximately 9600 ft3, would only pressurize the vaste decay tant to about 100 psig. The gaseous activity could then be discharged over a period of one week to allow dispersion in accordance with the long te;._
atmospheric diffusion model. The average annual concentration at the site 5 boundary, after a holdup of 30 days in the gas decay tank, would be about
.001 of MFC.
The gaseous concentrations in the makeup tank void were determined from j Henry's Law, assuming the tank gas space is in equilibrium with the reactor l coolant. The fraction of activity in the reactor coolant system which collected in the makeup tank was approximately h5 percent for Kr, 30 per cmt for Xe, and 0.lh percent for I. Tne activity levels - used for sources 1 L ., in.the makeup ..g , . , , ,
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11-8 (Revised h-8-68) .
0317
tank correspond to the reactor coolant system activity at the end of the
~b" A second core cycle. :It is- assumed that the tank vill he v .nted'once J a year.e -
to the vaste decay tank. The volume of gas in the mau up tank is about 300 ft3 at 45 psia. 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 to give an average annual concentration of about .0005 of 5 MPC at the site boundary with no allowance for decay.
Calculations similar to those used for the makeup tank vere performed to detemine the activity in the pressuri::er. It was found that the activity in the pressurizer was approximately one-third the activity in the makeup tank. Venting of the pressurizer results in only about 60 ft3 of gas , which can be released from the vaste decay tank over a period of one week to give a yearly average concentration of about .0002 of NPC at the site 5 boundary.
The activity level in the reactor building atmosphere was computed assuming a reactor coolant system leakage to the reactor building air of 10 gpd.
All of the Kr and Xe, and 50 per cent of the I and Cs that leaked from the reactor coolant system, was dispersed throughout the reactor building atmos-phere.
Activity buildup in the reactor building was computed over the 30 days of fuel leakage, i.e. , it was assu=ed that no purge had been made for 30 days.
This quantity of activity was then discharged to the atmosphere, without decay, by way of the reactor building purge system. The concentration at the site boundary averaged over 30 days was computed to be .02 of MPC. 5 Venting the reactor building once each 30 days would give an average yearly concentration of .02 of MPC, at the site boundary. This calcula-tion was based on the use of the short term dispersion model discussed in Section 2 (Table 2-3).
A preliminary analysis has been made to examine the consequences of reactor j operation with steam generator tube leakage and 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 limiting concentration was established by the activity carried with the air ejector exhaust to the plant vent to remain within the allowable discharge limits of 10 CFR 20. At this limiting concentration, the direct dose rate from the condenser is below the permissible value for continued access.
In the vacuum pumps 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 off-gas vent with no radioactive decay. With this con'servative assum- 1 ption, a reactor coolant leak rate of 1 gpm results in a concentration of
.075 of MPC at the site boundary. The analysis was' based on.1 gpm tube leslage continuously over a year. l5 n -
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11-9 (Revised h-8-68)
This evaluation demonstrates that the total yearly average concentration of activity at the site boundary from all modes of release, include:.g pres-surizer vent, reactor building purge, venting of the nakeup tank, startup expansion and dilution, chemical shim bleed, and steam generator tube leakage is a maximum of about .10 of MPC. The evaluation also demonstrates that l5 equipment capacities are adequate to accommodate and store radioactive gases as necessary. Thus, the system design is adequate to insure safe disposal of gaseous vastes.
Since the nuclear unit vill be located immediately adjacent to existing fossil plants, on-site gaseous vaste concentrations were determined to evaluate the potential hazard for fossil plant workers. The analysis was the same as the above site boundary calculations except that the MPC concentrations of the various isotopes were those applicable to restricted areas (10 CFR 20, Appendix B, Table I). Atmospheric dispersion factors for long-term releases were those reported in Section 2. The following average annual concentrations were calculated at a distance of 300 feet from the nuclear unit: l5 Source of Gaseous Wasg Average Annual Concentration (Fraction of MFC)
- 1) Lifetime Shim Bleed .0007
- 2) Cold Shutdown and Startup .0002
- 3) Venting of Makeup Tank .0002 g h) Venting of Pressurizer .0001
- 5) Reactor Building Purge .0110
- 0) Steam Generator Leakage .0280 Total .0h02 5
The total yearly average concentration is thus .0h of MPC at the fossil plants for all modes of gaseous release from the nuclear unit.
11.1.2.5 3 Radioactive Waste Disposal System Failures The possibility of a significant activity release off the site from accidents in either the solid or the liquid vaste disposal equipment is extremely remote. Both of these systems vill be located in shielded, controlled-access areas with provisions for maintaining conta=ination control in the event of spills or leakage. Solid vastes vill be disposed 0319
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l 11-10 (Revised h-8-68) l l
by licensed contractors in accordance with ICC regulations. Liquid vastes vill be sampled prior to discharge and vill be monitored during
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- ,, , discharge;to insure. compliance with.10 CFR 20...A tabulation of potential- -
.. N vaste disposal system failures and their consequences is presented in Table 11-6.
Radioactive gases will be sampled and discharged in compliance with the requirements of 10 CFR 20. In the event of vaste decay tank failure, these gases vould be released to the decay tank compartment, and then released to the plant vent via the normal ventilation system.
The maximum activity in a vaste gas decay tank vill occur following a boron dilution cycle during reactor startup just prior to switching to deborating demineralizer for boren removal. The reactor 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 72,000 gal, of reactor coolant would be let down at this time. It is assumed that the purification demineralizers have a removal efficiency of 99 per cent for iodine and zero removal efficiency for noble gases. The remaining gaseous activity vill be carried with the water to the reactor coolant bleed holdup tanks, where it is assumed that the gases are immediately re-leased from the water and carried with the purge gases to the vaste gas decay tank. This assumption is quite conservative since the gas release rate vill occur due to diffusion from the surface in accor-dance with Henry's Law and occur over a considerable time period.
Similarly, it is conservatively assumed that the gases do not undergo radioactive decay after leaving the reactor coolant system. With these
/'\ assumptions, the following activity is calculated to exist in the vaste 5Y gas decay tank:
Isotope Total Curies Kr 85m Shh.0 Kr 85 h,200.0 Kr 87 300.0 Kr 88 1,000.0 I 131 9.0 I 132 13.h I 133 12.2 I 13h 1.5 I 135 5.7 Xe 131m 570.0 Xe 133m 870.0 Xe 133 79,000.0 Xe 135m 272.0 Xe 135 2,550.0 Xe 138 136.0 The release of the above activity to the environment has been evaluated both for a slow release from a leaking safety valve and for a sudden ,
release due to a complete rupture of the vaste gas tank. The leaking safety valve is considered to ch ' conformity with 10 CFR 20 and the complete tank rupture is analyzed on the basis of 10 CFR 100.
O' V ,
} 11-11 (Revised 1-15-68)
1 For the leaking safety valve calculation, a leak of 0.3 ser per day is con- 2 sidered.
of1x10gtmospheriedilutioniscalculatedusingthelong-term sec/m3 %/Qvalue The resulting concentration at the site boundary is g
about .0001 MPC which is well below 10 CFR 20 limits.
A complete rupture of the tank is assumed to release all the activity as a puff from the plant vent. Atmospheric dilution is calculated using the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7/Q value of 3 x 10-4 sec/m3 The integrated whole body dose at the site boundary is 2.2 rem and the thyroid dose is 2 rem both of which are well within the limits of 10 CFR 100.
O 0321-
~. - '
11-lla (Revised 2-7-68)
I 11.1.3 TESTS AND INSPECTIONS
. Functional,operationaLtests and inspections..of the Waste . Disposal, Systema .
.? v vill be made as required to insure performance consistent Vith the require-ments of 10 CFR 20.
11.2 RADIATION SHIELDING 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY SHIELDING 11.2.1.1 Design Criteria i
Plant operating personnel and the general public must be protected by
, radiation shielding wherever radiation hazards exist. Protection will be in accordance with limits on radiation exposure as outlined in 10 CFR 20. The shielding vill be designed to perform two primary functions:
, (1) to insure that during normal plant 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 provide the necessary protection of operating personnel following a reactor accident so that the accident may be terminated without excessive radiation exposure to the operators or to the general public.
To comply with limits specified in 10 CFR 20, the shielding vill be designed to give the following radiation dose rate levels throughout the plant from direct and scattered radiation:
Full Power Overation Conditions (1% failed fuel)
Location Dose Rate, mrem /hr Office, Control Room, and Turbine Building 0.50 t
Reactor Building:
! Accessible Areas 10.0 Auxiliary Building:
Accessible Areas 1 to 5.0 Maximum Hypcthetical Accident Conditions Location ,, Dose Rate, mrem /hr Inside Control Room 3 rem integrated whole body dose over 90 days l
l Outside Reactor Building )
) See Section 114 (Safety Analysis)
. ) for Integral Dose Rate Curves Site Boundary O
11-12 Q
p 11.2.1.2 Description of Shielding y> cr
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, ; y -
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11.2.1.2.1 Primary Shield The primary shield will be a large mass of reinforced concrete surrounding the reactor vessel and extending upward from the reactor building floor to form the valls of the fuel transfer canal. The preliminary shield thickness ir, 5 ft up to the height of the reactor vessel flange where the thickness is reduced to k.5 ft. The primary shield will meet the following objectives:
- a. To reduce, in conjunction with the secondary shield, the radiation level from sources within the reacter vessel and reactor coolant system to allow limited access to the reactor building during normal full power operation,
- b. To limit the radiation level after shut down from sources within the reactor vessel to permit limited access to the reactor coolant system equipment.
- c. To limit neutron flux activation of component and structural materials.
The neutron and gamma-ray heating of the primary shield vill be dissipated by the concrete shield cooling system. The primary shield concrete vill be cooled to maintain temperatures less than 150 F.
11.2.1.2.2 Secondary Shield The secondary . shield will be a reinforced concrete structure surrounding the reactor coolant equipment, including piping, pumps, and steam generators.
This shield will protect personnel from the direct gamma radiation resulting from reactor coolant activation products and fission products carried away from the core by the reactor coolant. In addition, the secondary shield vill supplement the primary shield by attenuating neutron and gamma radia-tion escaping from the primary shield. The secondary shield vill be sized to allow limited access to the reactor building during full power operation.
The preliminary thickness of secondary shield valls is h.5 ft.
11.2.1.2.3 Reactor Building Shield The reactor building shield will be a reinforced, prestressed concrete con-tainment structure which completely surrounds the nuclear steam supply system. At full power operation, this shield will attenuate any radiation escaping from the primary-secondary shield complex such that radiation levels outside the reactor building vill be less than 0.5 mrem /br. In addition, the reactor building structure vill shield personnel from radia-tion sources inside the building following a Maximum Hypothetical Accident (MHA). The shielding vill be of sufficient thickness to allow personnel a reasonable time period in which to evacuate the immediate vicinity of the reactor building following the MHA without excessive radiation exposure.
11-13 OD3
'~
The curves in Section 1h (Safety Analysis) indicate an integrated direct dose of 9 rem over a period of two hours immediately outside the reactor building following the MEA. Preliminary thicknesses of the reactor building vall and dome are 3.5 ft and 3 ft respectively.
11.2.1.2.k Control Room Shield The control room shielding vill be designed for continuous occupancy for essential control room personnel following a Maximum Hypothetical Accident.
This vould enable full control and shutdown procedures to be carried out without hazard to the control room operators. Preliminary thickness of the control room shielding is 2 ft. This ensures that the integrated whole body dose over 90 days following the MHA vill not exceed 3 rems, ventilation of the control room under post-accident conditions will be controlled as described in Section 9.7.2.
11.2.1.2.5 Auxiliary Shield Auxiliary shielding vill include all concrete valls, covers, and removable blocks which vill shield the numerous sources of radiation occurring in the radioactive vaste disposal, makeup and purification, chemical addition and sampling systems. Typical components which require shielding include vaste noldup tanks, boric acid and vaste evaporators , makeup tank, vaste decay tanks, demineralizers, makeup pumps, vaste drumming area, reactor coolant drain tank, and reactor building sump pump.
11.2.1.2.6 Spent Fuel Shielding Shielding vill te provided for protection during all phases of spent fuel lh removal and storage. Operations requiring shielding of personnel are spent fuel removal from reactor, spent fuel transfer through refueling canal and transfer tubes, spent fuel storage, and spent fuel shipping cask loading prior to transportation. Since all spent fuel removal and transfer operations will be carried out under borated water, minimum vater depths above the tops of the fuel assemblies vill be established to provide radiation shielding protection. Water depths during handling are a minimum of 10 ft in the reacter cavity and fuel transfer canal and 13 ft over rtored assemblies in the spent fuel storage area. The dose rates at the water surface vill be less than 10 mrem /hr. The concrete valls of the fuel t*ansfer canal and spent fuel pit vill supplement the water shielding and will limit the maximu= continuous radiation dose levels in working areas to less than 2.5 crem/hr.
The refueling water and cono-et: salle aAso provide shielding from activated control rod clusters and reactor internals which vill be removed at refueling times. Although dose rates vill generally be less than 2.5 mrem /hr in working areas, certain manipulations of fuel assemblies, rod clusters, or reactor internals may produce short term exposures in excess of 2.5 mre=/hr.
However, the radiation levels vill be closely monitored during refueling operations to establish the allovable exposure times for plant personnel in order not to exceed the integrated doses specified in 10 CFR 20.
.. .t ,. ,
. +, .. c -
ll-lh 324
t s 11.2.1.2.7 Materials and Structural Requirements
. . . .a . .. ... . . , n + r. .- m,. ~ . . , .~.. u. r - k , . . .t The material used for the primary, secondary, reactor building, and auxiliary shields will be ordinary concrete with density of approximately 140 lb/ft3. Since the primary and secondary shielding valls serve as the refueling structure, give support for the reactor coolant components under pipe rupture conditions, and provide missile shielding, they vill be rein-forced and designed to be self-supporting.
Times of occupancy in restricted areas will vary depending on measured radiation levels in each area. Such areas as containment operating floor, reactor vessel head prior to refueling, primary loop compartments after sh;tdown, and spent fuel handling areas will be surveyed prior to access and a time-limit.ed work schedule vill be set up.
11.2.1.3 Evaluation 11.2.1.3.1 Radiation Sources The shielding vill be designed to attenuate neutron and gamma radiation emanating from the following basic sources:
- a. Reactor Core, Interntas, and Reactor Vessel
- b. Reactor Coolant Loops
- c. Radioactive material released during accidents
- d. Auxiliary Systems Equipment
- e. Spent Fuel Elements Source magnitudes are determined for the reactor operating at the maximum expected power level of 25kh MWt with reactor coolant activity levels corresponding to 1 per cent failed fuel. Gamma-ray yield and spectral distributions from prompt fission and gross fission product activity are based on 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 Molern Physics, Vol. 30, No. 2 (April 1958).
The production of N-16 in the reactor coolant is calculated with a B&W code which comptes the integral of the 0-16 (n,p) N-16 cross section over the neutron *' lux 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) l N-16' crcss section used is that reported in WAPD-BT-25. Activities of i individual fission products in the core, reactor, coolant, and reactor auxiliary '
systems are determined 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 leakage from the core to the coolant, and removal from the coolant i by purification and leakage, are calculated.
O .
0325 l
l 11-15
11.2.1.3.2 Neutron and Gamma Shields--Calculation Methods The primary shield preliminary thickness is based on work perfor=ed for the Oconee Nuclear Station using Babcock and Wilcox computer codes which' solve the neutron and gamma-ray attenuation equations for the multi-layer source-shield complex. Neutron penetration in shield regions was calculated using the B&W LIFEX code as a coefficient generator to provide input data into either the TOPIC code or MIST code. TOPIC (IDo-16968) and MIST (IDO-16856) are programs which solve the transport equation using the Carlson SN method in cylindrical and slab geometries respectively, and were used to generate h-group fluxes in the radial and axial directions from the core.
Gamma-ray attenuation was calculated using the Taylor exponential form of buildup with the gamma source strengths divided into 1 Mev energy intervals between 1 and 10 Mev. The equstions for the d! rect gamma flux from the simpler geometric sources (line, disc, truncated cone, and cylinder) were solved by a Basic Geometry Code. For the more complex source-shield con-figurations where non-uniform source distributions may exist, a kernel integration code was used. This program uses a point kernel attenuation along a 14ne-of-sight from the source point to the dose point and computes the ga=ma Ilux by summing over the source distribution. Secondary gamma-ray penetration was calculated using a Secondary Ca=ma program for a laminated, semi-infinite shield array. The aforementioned B&W codes and techniques are described in IDO-2hh67 11.2.1.3.3 MHA 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 release following a reactor accident. For the calculations it was assumed that 100 per cent of the gases , 50 per cent of the halogens , and h
1 per cent of the solid fission products were instantaneously released to the reactor building following a buildup period in the core of 600 full power (2,5hh MWt) days.
The fission product activity was assumed to be uniformly dispersed throughout the reactor building volume, and the reactor building was represented by a cylindrical source for the dose calculations. The integrated dose over various time intervals was computed as a function of distance from the reactor building. The results are given in 14.2.2.h.
11.2.1 3.h operating Limits The radiatitn shielding design, including heating and dose rate profiles, temperature ciistributions, and coolant flow requirements, vill be evaluated during the detailed design of the plant to establish the operative limits.
11.2.1 3.5 Radiation Surveys Neutron and gn=ma radiation surveys vill be performed in all accessible areas of the plant as required to determine shielding integrity. Plans and procedures for radiation surveys during operation and following shutdown vill be formulated during the detailed plant design.
- . . .. . .. ~...
~ ~. .
11-16 .
11.2.2 AREA RADIATION MONITORING SYSTEM 11.2.2.1 Design Bases
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The fixed radiation monitoring system will be designed to indicate and alarm high radiation monitoring levels throughout the plant. Visual presentation of readings, recorded presentation, and an audible / visible alarm at both 2 the detector location and the Control Room will be prcvided for all gamma channels and most atrespheric channels. All instrumentation for the radiation monitoring sy3 tem vill obtain its voltage supply from the 120 volt a-c essential service busses and each detector will have a " Loss of Power" alarm. The normal high sdiation alarm setpoint will be 10 percent above the normal operational reading of the detector. A maximum alarm point will be set to correspond to the MFC value specified in 10 CFR 20.
The maximum alarm point set at 10 CFR values could be either an actual value or a calculated number correspording to 10 CFR E' limits.
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. Near incere instrument space inside the Reactor Building.
- d. Fuel handling bridge in Auxiliary Building.
O e. Auxiliary Building pump area,
- f. Auxiliary Building near sample sink.
- g. Auxiliary Building cask decontamination and loading area.
- h. Auxiliary Lailding in shutdown cooler area.
- i. Auxiliary Building near Reactor Building component cooling water Coolers.
J. Radio-Chemistry laboratory.
- k. Cable and computer room.
- 1. Machine shop,
- m. Control room.
Air Partic' late u and Radio Gas Detectors to be mounted as follows:
. a. Plant Vent s
- b. Inside Reactor Building
(~% c. Radio-Chemistry Laboratory l V
11-17 (Revised 2-7-68) 0327
- d. Outside Air Monitor
- e. Condenser Vacuum Pump Discharge gg
- f. Control Room and Auxiliary Building
- g. Spent Puel Area Detector ranges will be determined depending upon the normal background at the detector locations and the calculated levels for abnormal conditions.
Radioactive test sources will be available to allow the overall system per-formance to be verified at regular intervals.
11.2.2.3 Evaluation Area radiation monitor detectors will 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 vill be sufficiently high to indicate the accessibility of the Reactor Building following a serious accident inside.
The incore instrument area vill be monitored, and a local alarm will be provided to warn if a high radiation level exists or is created while incore assemblies are being manipulated.
The sample sink area in the Auxiliary Building vill be equipped with a detector to alarm an abnormal condition in connection with system sampling.
Alarms vill be actuated in the control room and at the detectors if an abnormal change in radiation background occurs.
The radiation monitoring system shall be checked and calibrated at least once per month. When any portion of the radiacion monitoring system requires maintenance, that unit shall be completely checked and calibrated immediately after completion of maintenance.
11.2.3 HEALTH PHYSICS The plant superintendent is responsible for radiation protection and con-tamination control. All personnel assigned to the plant and all visitors will be required to follow rules and procedures established by administrative control for protection against radiation and contamination.
Under supervision of the plant superintendent, the administration of the radiation protection program vill be the responsibility of the plant Health Physicist. It will be the responsibility of the Health Physics section to train plant personnel in radiation safety; to locate, measure, and evaluate radiological problems; and to make recommendations for control or elimination of radiation hazards. The Health Physics section vill function in an sjvisory capacity to assist all personnel in carrying out their radiation safety qg>-
11-18 01328
o responsibilities and to audit all aspects of plant operation and maintenance d i to assure' safe conditions and'c'ompliance vith the' AEC'an' d other;' federal * ^
and state regulations concerning radiation protection.
"**?
Administrative controls will be established to assure that a31 procedures and requirements relating to radiation protection are followed by all plant personnel. The procedures that control radiation exposure vill be subject to the same review and approval e.s those that govern all other plant procedures (see Section 12 5, Administrative Control). These procedures will include a Radiation Work Permit system. All work on systems or locations where exposure to radiation or radioactive materials is or may be involved will require an appropriate Radiation Work Permit.
RADIATION WORK PERMITF A Radiation Work Permit shall be obtained by all personnel prior to entering a Control Area or performing any work on radioactive or contaminated material or equipment.
In the event that the safety of the plant or its personnel are endangered, entry may be made into a Control Area simultaneously with monitoring per-sonnel. A Radiation Work Permit shall be completed as soon as possible after correction of the situation.
Radiation Work Permits shall be issued routinely by the Shift Supervisor.
These permits shall show:
%)
- a. The nature of the work to be performed.
- b. Expected duration of vork.
- c. Names of persons to perfor= the work,
- d. Signature of authorizing Shift Supervisor,
- e. Signature of an individual from the Health Physics Group who shall ensure that:
- 1. Designated persounel are within their permissible exposure limits.
- 2. The area has been adequately surveyed prior to entry.
- 3. Adequate protective clothing and supplies are available at the control point. ,
- k. Monitors are available for the work.
All such permits shall be filed with the Health Physics group for future reference.
O 0329 11-19
11.2.3.1 Personnel Monitoring system The personnel monitoring program shall insure that the reco=mendations O and regulations of tne Atomic Energy Commission are followed for all involved personnel. All peraonnel entering a Control Area shall wear a film badge or its equivalent. Exposures shall be maintained within the limits established in 10 CFR 20. In addition, those persons who ordinarily work in restricted areas or whose job requires frequent access to these areas will have pocket chambers, self-reading dosimeters, pocket high-radiation alarms, vrist badges, and finger tabs readily available for use, when required by plant conditions. This personnel monitoring equipment will also be available on a day-to-day basis for those persons , employees , or visitors not assignesl to the plant who have occasion to enter Restricted Areas or to perform work involving possible exposure to radiation. Records of radiation exposure history and current occupational exposure vill be maintained by the Health Physics group for each individual for whom personnel monitoring is required.
The external radiation dose to personnel vill be determined on e. daily and/or weekly basis, as required, by means of the pocket chamber and dosimeter.
Film badges will be processed monthly or more frequently when conditirns indicate it is necessary.
11.2.3.2 Personnel Protective Eculpment Special " protective" or "anticontamination" clothing vill be furnished and vorn as necessary to protect personnel against contact with radio-active contamination. Change Rooms vill be conveniently located for proper utilization of this protective clothing. Respiratory protective equipment vill also be available for the protection of personnel against airborne radioactive contamination and vill consist of full face filter h nasks, self-contained air-breathing units, or air-supplied masks and hoc.is . The first line of defense against airborne contamination in the work area is the ventilation system. However, respiratory protective equipment will be provided should its use become necessary.
Maintenance of the above equipment vill be in accordance with the manu-facturer's recommendations and rules of good practice such as those published by the American Industrial Hygiene Association in its "Respir-atory Protective Devices Manual." The use and maintenance of this equipment vill be under the direct control of the Health Physics group, and personnel vill be trained in the use of this equipment before using it in the performance of work.
11.2.3.3 Change Room Facilities Change room facilities vill be provided where personnel may obtain clean protective clothing required for plant work. These facilities vill be divided into " clean" and " contaminated" sections. The " contaminated" section of the change rooms will be used for the removal and handling of contaminated protective clothing after use. Showers , sinks , and necessary monitoring equipment also vill be provided in the change areas to aid in the decontamination of personnel.
. .f, , , .... , < -
, ..t 11-20 h L
n Appropriate written procedures will govern the proper use of protective 2
og- . ' clothing;.where.and, how. its 'iseto be vorn and removed;. and how ther change o ""'e-room and decontamination facilities for personnel, equipment, and plant areas are to be used.
In order to protect personnel from access to high radiation areas that may exist temporarily or semipermanently as a result of plant operations and maintenance; varning signs, audible and visual indicators, barricades, and locked doors vill be used as necessary. Administrative procedures will also be written to control access to high radiation areas. The Radiation Work Pemit System will also be utilized to control access to high radiation areas.
11.2.3.4 Health Physics Laboratory Facilities The plant will include a Health Physics Laboratory with facilities and equipment for detecting, analyzing, and measuring all types of radiation and for evaluating any radiological problem which may be anticipated.
Counting equipment (such as G-M, scintillation, and proportional counters) vill be provided in an appropriate shielded counting room for detecting and measuring all types of radiation as veil as equipment (such as a multi-channel analyzer) for the identification of specific radionuclides. Equip-ment and facilities for analyzing environmental survey and bioassay samples will also be included in ' he Health Physics Laboratory. Maintenance and use of the Health Physics Laboratory facilities and equipment vill be the responsibility of the Health Physics group.
11.2.3.5 Health Physics Instrumentation Portable radiation survey instruments vill be provided for use by the Health Physics group as well as for operating and maintenance personnel.
A variety of instruments will be selected to cover the entire spectrum of radiation measurement problems anticipated at the plant. Sufficient quantities will be obtained to a.ow for use, calibration, maintenance, and repair. This will include instruments for detecting and measuring alpha, beta, gamma, and neutron radiation. In addition to the portable radiation monitoring instruments, fixed monitoring instruments, i.e.,
count rate meters, vill be located at exits from restricted areas. These instruments are intended to prevent any contamination on personnel, material, or equipment from being spread into unrestricted areas. Appropriate monitoring instruments will also be available at various locations within the restricted areas for contamination ecntrol purposes. Portal monitors will also be utilized, as appropriate, to control personnel egress from restricted areas or from the plant.
The plant vill have a permanently installed remote radiation and radio-activity monitoring system for locations where significant levels can be expected. This system vill monitor airborne particulate and gaseous radioytivity as well as external radiation levels. This system will
,preselt an' audible alarm and radiation level indication in the area of conggp:p in addition to the control room.
O u
i I
0131
" .N " 11-21 I
11.2.3.6 Medical Prorrams Facilities for screening personnel for contamination will be available on site with outside services utilized as bachup and support for this progrr.m.
A medical examination program appropriate for radiation workers vill be conducted to establish and maintain records of the physical status of each employee. Subsequent medical examinations will be held as determined necessary for radiation workers. Medical doctors, preferably in the local area, vill be used for this program. The health Physics group will be responsible for the program and will assist the physicians to preserve the health of the employees concerned and to confirm the radiation control methods employed at the plant.
11.3 REFERENCES
(1) Frank, P. W. , g al_. , RadiocheTistry of Third PWR Fuel Material Test -
X-1 Loop NRX Reactor, WAPD-TM-29, February 1957.
(2) Eichenberg, J. D., e_t,al., Effects of Irradiation on Bulk U0 , 2WAPD-183, October 1957.
(3) Allison, G. M. and Robertson, R. F. S. . The Behavior of Fission Products in Pressurized-Water Systems. A Review of Defect Tests on UO 2 Fuel Elements at Chalk River, AECL-1338, 1961.
(h) Allison, G. M. and Roe, R. K., The Relcase of Fission Gases & Iodines From Defected UO 2 Fuel Elements of Different Lengths, AECL-2206, June 1965.
(5) Duke " aver Company, Preliminary Safety Analysis Report, Volume II, 1966.
O
.* *. ks e e ','g* .r'- s
. 0332 I
- 11-22
Table 11-1 Radioactive Waste Quantities Per Unit
,, . . .p . , - ;. a '. , ... .~
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Quantity Per Year fas_te Source Per Nuclear Unit Assumptions and Comments Liquid Waste Reactor Coolant System:
Startup Expansion 128,000 gal h Cold Startups Startup Dilution 88,000 gal 2 Cold Startups at beginning of life and 1 cold startup at 100 and 200 full power days respec-tively.
Lifetime Shim Bleed 176,000 gal Dilution from lh60 to 175 ppm System Drain 45,000 gal Drain to level of outlet nozzles for refueling operations Sampling and Laboratory Drains 22,500 gal 12 samples per week at 5 gal.
per sample Purification Demineralizer Sluice 160 ft3 3 80 2 ftfg/ft3 resin sluice./ year replacemen Spent Fuel Pool Demineralizer 3
Sluice 42 ft 3 21 2ft p/ft/gearreplacementat resin sluice 1 Deborating Demineralizer Regen-3 eration and Rinse 2,500 ft 1 Regeneration per year per demineralizer at 20 ft3 /ft3 resin regeneration Miscellaneous System Leakage h5,000 gal 5 gph leakage Laundry 55,000 gal 150 gpd Showers 110,000 gal 10 showers per day at 30 gal per shower
- Caseous Warte (a)
Off-Gas from Reactor Coolant System 1,350 ft 3 Degas at 25ccH per liter 2
concentration Off-Gas from Liquid Sampling Th ft3 Degas at 25ccH2 per liter concentration q
(G~
11-23 (Revised 1-15-68) ---
0333 m,.__ __ - - , _ - _ , _ . -
Table 11-1 (cont'd)
Off-Gas from Makeup Tank 900 ft3 Vent once per year Off-Gas from Pressurizer 60 ft3 Vent once per year Solid Waste Purification Ecsin 3 80 ft Resin replacement once per year Spent Fuel Pool Demineralizer Resin 20 ft 3 Resin replacement once per year Evaporator Condensate Demineralizer Resin 3 2 ft Resin replacement once per year Evaporator Bottoms 800 ft3 concentrated to 20 per cent solids (a) Excludes reactor building and plant ventilation.
l O
i i
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11-2h
Table 11-2
, . . . .....v.~. x .
.. . ., .. . ... . c. . . . . , .. . . . : . ..:.
Escape Rate Coerficient, Element sec-1 Xe 1.0 x 10-7 Kr 1.0 x 10-7 I 0 Br 2.0 x 10 0 Cs 2.0 x 10 0 Rb 2.0 x 10 0 2.0 x 10 Mo h.0 x 10-9 Te h.0 x 10-9 Sr 2.0 x 10-10 Ba 2.0 x 10-10 Zr 11 Ce and other rare earths 1.0 x 10 1.0 x 10- 11 Table 11-3 Reactor Coolant Activities For a Unit Containing One Per Cent Defective Fuel Isotope Activity.JJc/ml Isotope Activity,JJc/ml Kr 85m 2.0 I 131 33 Kr 85 15.5 I 132 4.9 Kr 87 1.1 I 133 4.5 Kr 88 3.7 I 13h 0.55 Rb 88 3.7 I 135 2.1 Sr 89 0.057 Cs 136 0.81 Sr 90 0.0028 Cs 137 77.0 Sr 91 0.057 Cs 138 0.Th Sr 92 0.018 Mo 99 1.2 Xe 131m 2.1 Ba 139 0.088 Xe 133m 3.2 Ba 1ho 0.076 1 Xe 133 290.0 La 1h0 0.026 1.0 l Xe 135m Y 90 0.0007 Xe 135 9.h Y 91 0.CN3 Xe 138 05 Ce 14h 0.002T Tables 11.-2, 11-3 O) 11-25 0.35
_ ~ - . . _ - - . , _ - . . . , _ _ .
-% , .=
Table 11-4 Waste Disposal System Component Data P
Reactor Coolant Drain Tank Number 1 Volumc, cu ft (each) 1000 (7,500 gal) l5 Material Carbon Steel, Corrosion-Resistant Lining Deborating Demineralizer Humber 2 l5 Type Semiautomatic Regeneration Material Carbon Steel, Corrosion-Resistant Lining Reactor Coolant Bleed Holdup Tank Number 3 !5 Volume, cu ft (each) 11,000 (82,500 gal)
Material Carbon Steel, Corrosion-Resistant Lining Miscellaneous Waste Holdup Tank Number 1 l5 Volume, cu ft (each) 2,700 (20,000 gal)
Material Carbon Steel, Corrosion-Resistant Lining Waste Neutralization Tank
- Number 1 Volume, cu ft h00 (3,000 gal)
Material Carbon Steel, Corrosion-Resistant Lining Spent Resin Storage Tank Number 1 Volume, cu ft (each) h50 (3,375 gal)
Material Carbon Steel, Corrosion-Resistant Lining Evaporator Condensate Storage Tank l (Reactor Coolant Cendensate)
Number 2 Volume, cu ft (each) l5 1500 (11,250 gal)
Material Carbon Steel, Corrosion Resistant
. , , , , . Lining - -
TABLE 11-h 11-26 (Revised h-8-68) 0336
Table ll-h (Cont'd)
. . , , ,. ..e ja. . , s y . . .. - u , . . . n . g n i
.s , ., . ,'
. . . . m, . s.~ w.v . .:.na :,i::~ ,-P.~ i-i :t r. > ;+ . ~: > .w Evaporator Condensate Storage Tank (Miscellaneous Waste Condensate)
- Number 1 l5 Volume, cu ft (each) 500 (3,750 gal) 1 Material Carbon Steel, Corrosion-Resistant Lining Waste Evaporator (Reactor Coolant Waste) 4 Number 1 l
! Process Rate, gpm (each) 75 Material Stainless Steel Waste Evaporator (Miscellaneous Waste)
Number 1 Process Rate, gpm 7.5 Material Stainless Steel i
Evaporator Condensate D eineralizers Number 2 l5 Material Stainless Steel Reactor Building Sump Pump , ,
Number 1 Capacity, gpm (each) l5 200 Material Stainless Steel Waste Transfer Pump (Reactor Coolant Waste)
Number 2 Capacity, gpm (each) 100 Material Stainless Steel Waste Transfer Pump (Mi'acellaneous Waste)
Number 2 Capacity, gpm (each) 50 Material Stainless Steel Auxiliary Building Sump Tank Q[ Number 1 Capacity, gal 35.00 A Material Carbon Steel corrosion resistant lining
- TABLE 11-4 (Cont'd) 11-27 (Revised 4-8-68)
,. . . - . . - . . - -- . . . . . . - - - . - - -~.
Table 11-4 (Cont'd) Auxiliary Building Sump Tank Pump Number 2 P Capacity, gpm (each) 50 Material Stainless Steel Evaporator Feed Pump (Reactor Coolant Waste) Number 2 Y Capacity, gpm (each) 7.5 Material Stainless Steel Evaporator Feed Pump (Miscellaneous Waste) Number 2 Capacity, gpm (each) 75 Material Stainless Steel Evaporator Condensate Pump (Reactor Coolant Waste) ~ Number 2 Capacity, gpm (each) 20 Material Stainless Steel Evaporator Condensate Pump (Miscellaneous Waste) Number 2 Capacity, gpm (each) 20 Material Stainless Steel Evaporator Vacuum Pump Number 2 E Capacity, cfh (each) 12 Material Carbon Steel Waste Gas Compressor Humber 2 b Capacity, cfb (each) k0 Material Carbon Steel Waste Gas Decay Tank Number Volume, cu ft (each) 3 f 1500 Material Carbon Steel .~ . TABLE 11-4 (Cont!d) e 11-28 (Revised 4-8-68) 0338
TABLE 11-5 .... ).r.~. < , , . - . . = . 4. ~ . :m .s ...x.. . * . s. n - - - ' . .r.: : . ' i%>., : , e .+s.' - G:t :; - ' e :.'w ' N*vM % s ~' - h ~ - Maximum Activity Concentrations in the Plant Effluent 5 for Unit Operating With One Per Cent Failure Fuel Liquid Waste Yearly Average Concentrations in Circulating Water Discharge, Operation Fraction of MPC Lifetime Shim' Bleed 0.0002 Discharge of Miscellaneous Wastes 0.00001 1 Gaseous Wastes Yearly Average Concentration Operation at Site Boundary, Fraction of MFC Lifetime Shim Bleed (30 Day Holdup) .002 1 Startup Expansion and Dilution .001 5 Venting of Makeup Tank .0005 Venting of Pressurizer .0002 Reactor Building Purge .02 Steam Generator Tube Leakage of 1 gpm .075 TABLE 11-5 11-29 (Revised h-8-68) 0.D9
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O Table 11-6 Waste Disposal System Failure Analysis Component Failure Co=ments and Consecuences Reactor Building Sump Fails to close Backup isolation is provided Drain Valve (inside or on opposite side of reactor outside) building. Reactor Building Drain Fails to open Continuous drainage is not Line Valve (inside or required; the valve is located outside) for maintenance during operation. Reactor Building Sump Fails to operate Continuous operation is not Pump required; located for maintenance during operation. Reactor Coolant Drain Fails to operate Continuous venting is not Tank Vent Valve required; relief protection is provided for tank Fails to close Vent gas is conveyed to vaste gas decay tank and discharged through filters to plant vent. Waste Gas Vent Filters Rupture or lose High activity level monitored g efficiency and alarmed if unsufficient W plant vent dilution is avail-able. Waste gas is diverted to vaste gas decay tanks. Waste Gas Decay Tanks Leak rr rupture Building purged to plant vent through filters. Tanks . are protected by relief valves. Reactor Coolant Bleed Leak Leakage is collected in aux-Holdup Tanks iliary building drain sump for process or disposal; building is continuously purged to plant vent. Evaporator Train Fails to operate Continuous operation is not required; vaste gas decay tanks provide for vaste col-lection during maintenance. Deborating Demineralizers Exhausted resin Spare unit placed in service while original unit is regen-erated. Startup time is in-creased near end-of-life depending on balance between rod worth and boric acid required. lh l 11-30 i 0340
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