ML19309C548
| ML19309C548 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/01/1967 |
| From: | JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO. |
| To: | |
| References | |
| NUDOCS 8004080746 | |
| Download: ML19309C548 (38) | |
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TABLE OF CONTEh"IS Section Pgge, e
11 RADIOACTIVE WASTES AND RADIATION FROTECTION n.1 11.1 RADICACTIVE WASTES 11-1 11.1.1 DESIGN BASES 11-1 11.1.1.1 Perfemance Objectives 11-1 11.1.1.2 Radioactive Waste Quantities 11-1 11.1.1.3 Waste Activity 11-1 11.1.1.h Diseesal Methods 11-2 11.1.1.5 shielding 11-3 l
11.1.2 SYSTEM Dr. SIGN 11-3 !
i n.l.2.1 Liquid Waste Disposal System 11-3 11.1.2.2 Solid Waste Dispesal System 11-4 11.1.2.3 Gaseous Waste Disposal System u-5 n.l.2.h System Radiation Monitoring 11-5 n.l.2.5 Design Evaluation ll-ti n.1.3 TESTS AND INSPECTIONS 11-11 11.2 RADIATION SHIELDING 11-11 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY SHIELDING 11-11 11.2.1.1 Design Criteria ll-l' 11.2.1.2 Descriptien of Shielding 11-12 11.2.1.3 Evaluation n-lk O
u_, .000I 230"
11.2.2 AREA RADIATION MONITORING SYSTEM 11-16 11.2.2.1 tesign Bases 11-16 l 11.2.2.2 teserietien 11-16 !
11.2.2.3 Evaluatien 11-17 i
11.2.3 EEALTH PHYSICS 11-17 1
11.2.3.1 Persennel Monitoring System 11-19 ;
11.2.3.2 .Persennel Protective Ec_uitment 11-19 11.2.3.3 Change Roem Facilities 11-19 l
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11.2.3.h Health Physics Laboratory Facilities 11-20 11.2.3 5 Health Physics Instrumentation 11-20 11.2.3.6 Medical Programt 11-21 11.3 REFERDICES u-21 O
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LIST OF TABLES Table No. Title Page 11-1 Radioactive Waste Quantities 11-22 11-2 Escape Rate Coefficients for Fission Product Release 11-2h 11-3 Reactor Coolant Activities Free One Per Cent Defective Fuel 11-2h 11-4 Waste Disposal System Ccmponent Data 11-25 11-5 Maximum Activity Concentrations in the Station Effluent With One Per Cent Failed Fuel 11-28 11-6 Waste Disposal System Failure Analysis 11-29 0
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000f232'l.11-111
LIST OF FIGURES Figure No. Title 11-1 Waste Disposal System O
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11 RADI0 ACTIVE VASTE AND RADIATION PROTECTION 11.1 RADIOACTIVE WASTES 11.1.1 DESIGN BASES 11.1.1.1 Perfomance Objectives The vaste disposal system vill be designed to provide controlled hacdling and disposal of liquid, gaseous, and solid vastes which vill be generated ^
during plant operation. The design criteria are to insure that station personnel and the general public are protected against excessive exposure to radiation frca vastes in accordance with limits defined in 10 CFR 20.
11.1.1.2 Radioactive Waste Quantities The estimated volumes of radioactive vastes generated durin6 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 hand 11n6 equipment has been detemined on the basis of fissica product leakage through clad defects in 1 per cent of the fuel. The activicy levels were ecmputed assuming full power operation of 2,535 MWt for one core cycle with no defective fuel folleved by operatica over the second core cycle with 1 per cent defective fuel. Continuous reactor coolant purification at a rate of cne 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 chan6es in deminerali:er 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 en the use of " escape rate coefficients" (sec -1) as detemined frca experiments involving purposely defected fuel elements (References 1, 2, 3, h) . 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 perfomed with a digital ecmputer code which solves the differential equations for a five-member radioactive chain for buildup in the fuel, release to the coolant, removal from the cociant by purification and leakage, and collection on i a resin or in a holdup tank. The activity levels in the reactor coolart durin6 full power cperation at the end of the second core cycle are shcvn in Table 11-3 The liquid vaste generated by leakage, sampling, and dertinerali:er sluice or rinse is assumed to have an activity concentration equal to the cen-centration in the rector coolant. Reactor coolant bleed vill be taken frca the devnstream side of the purificatica deminerali::er. It is assutsed to have the same act:l.vity concentration as the reactor coolant reduced by l
the decontamination factor of the purification deminerali:er. Laund:7 and shever vastes are assumed to contain negligible amcunts of radioactivity.
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Gaseous activity vill be generated by the evolution of radioactive gases from liquids stored in tanks throughout the station. T$ese include such items as the pressurizer and letdown. stora6e tank which are vented to the vaste gas disposal system. The activity of the gases is 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.h Disposal Methods Liquid vastes frem the station vill be handled in two separate streams 2 usin6 two evaporator chains. Reactor coolant bleed vill be fed through one chain and miscellaneous vastes, which include reactor building sump drains, reactor coolant drains , and floor drains, vill be processed throu6h the other chain. The treatment of the liquid vastes will be in one of the following ways:
- a. Reactor Coolant Bleed
- 1. Collected, monitored, demineralized, and stored in the reactor coolant vaste condensate storage tanks.
- 2. Collected, monitored. concentrated, and either reclaimed through boric acid recycle or discharged to vaste drum-ming area for packaging and off-site disposal.
- 3. Condensate resulting frem the concentration operatica vill either be reclaimed as desineralized water or 2 g
discharged with the cooling tower blevdevn effluent.
- b. Miscellaneous Wastes
- 1. Collected, monitored, demineralized, and stored in the miscellaneous vaste condensate storage tanks.
- 2. Collected, =enitored, ccncentrated, packa6ed, and shipped off-site.
- 3. Condensate resulting from the concentration operation vill be either reclaimed as demineralized water or discharged with the coolin6 tower blevdown effluent.
Gaseous vastes are disposed of using one of two =ethods:
- a. Continuous dilution and discharge throu6h vaste gas filters to the station vent when activity levels permit.
t, . Diversion to vaste gas holdup tanks with sampling and cen-f,, trolled subsequent release through vaste gas filters to the station vent.
Solid radioactive vastes vill be accumulated and packaged in special drums suitable for ICC-approved shipment off-site to a licensed vaste disposal g facility.
11-2 (Revised 10-2-o7) dt'.
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n.1.1.5 shielding O Shielding for the components of the vaste disposal system vill 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 vin be located in the auxiliary building. The shield design criteria for the auxiliary building is a dose rate of 1 to 2 mrem /hr in cen-tinuously accessible areas and 5 to 15 mrem /hr in areas requiring limited access. The components of the vaste disposal system vill be shielded by concrete vans and floors varying thicknesses depending on the magnitudes of the sources in each component and on the access requirements in a particula-l area. In some areas local shielding in the form of lead or removable concrete - blocks vill be utilized to facilitate maintenance or repair operations. 11.1.2 SYSTEM DESIGN
- 11.1.2.1 Liquid Waste Disposal System Liquid vaste handling vill be divided into two separate vaste processing chains. One chain vill process the reactor coolant bleed stream and the other will handle all misee naneous liquid wastes. The system flow dia6 ram with necessary instrumentation and controls is shown in Figure 11-1.
Waste disposal system component data is given in Table 11-4. Reactor coolant will be received from the makeup and purification system and will be the largest single source of operational liquid waste to be handled. This liquid vill be received as a result of reactor coolant expansion and O operational requirements for reduction of reactor coolant boric acid content. l l It vin be either conveyed to reactor coolant bleed holdup tanks for storage l or passed through deborating domineralizers for boric acid removal and re- I turned as unborated maketp to the makeup and purification system. The deborating demineralizers vill be used only for boric acid concentrations below 1000 ppm to limit the rate at which resins are used up. Se reactor coolant bleed holdup tanks will be sized to contain one reactor coolant system volume each. The contents of each tank will be periodically sampled to determine their radioactive content. These tanks will feed the vaste batch tank which in turn supplies the vaste evaporator or concentrator. The contents of the vaste batch tank will be pumped continuously through the evaporator using the evaporator feed pumps and returned to the batch tank in a closed loop so that the vastes in the loop become progressively concentrated. When vastes are sufficiently concentrated, the evaporator feed pumps vill be shut down and the concentrated vastes will automatically drain to the vaste batch tank. Residual vastes remaining in the evaporator vin be flushed out by returning small amounts of condensate through the evaporator body. This backvashing 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 sampled to determine their content. The vastes say then be disposed of by either pump-ing to the vaste drumming area or recycling through domineralizers to re-claim boric acid. The evaporator condensates will be collected in a con-. densate test tank where they are sampled to determine quality and activity., level. Condensates vill then be pumped through cation and evaporator con-l densate demineralizers to the evaporator condensate storage tanks for ' I
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-3 (Revised 10-2-67)
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s. reuse or =ay be pumped to the river for disposal after mixing with the 2 cooling tower blevdown effluent. Gaseous vastes vill be removed con-tinuously from both the vaste batch tank and evaporator 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
=enitored for activity, heldup for decay as required, and then released at a controlled rate through the station vent. A sonitor located in the gaseous discharge line to the station vent will be equipped with an indicator, and alarm to annunciate a high activity level. The high level alarm actuates an interlock to stop the discharge of gaseous effluents l2 from the vaste gas system.
The second evaporator chain vill process liquid vastes collected by the miscellareous vaste tank and auxiliary building su=p tank. The 2 miscellaneous vaste tank vill collect.vastes from the reactor building st=p, reactor coolant drain tank, miscellaneous coolers, and liquid samples. Th! auxiliary building su=p tank vill collect a variety of liquid vaste including demineraliser rinse; chemical tank drains ; waste gas soisture separator drains ; and laundry, shower, and floor drains. The liquid vt stes .vi.11 be pumped into a neutralising tank where the pH of the solution is adjusted as necessary to prevent foaming and sa=ples are taken to determine activity. Wastes vill t. hen be transferred to 2 a vaste batch tank for cycling through an evaporator or =ay be pumped to the evaporator condensate storage tanks through the cation and evaporator condensate demineralizers. When vastes are sufficiently concentrated, the concentratec will be collected in the vaste batch tank and subsequently pumped to the vaste drumming area for packaging and disposal. Condensate vill be collected in a condensate test tank, sampled for activity, and 2 subsequently either reused as demineralised water or discharged to the river after mixing with the cooling tower blevdown effluent. Gaseour vastes vill be removed by vacuum pumps from the evaporator and batch tank and passed to the vaste gas decay tank for ultimate release through the plant vent. Both evaporato chains vi31 be designed to give decontamination factors higherthanlo{andvilllesizedtoprocessvastesataratewellin excess of the expected vat.te accumulation rate. As indicated above, all 1.. quid vaste vill be sampled and analysed for l radioactive concentration prior to disposal. It discharge to the environment is permissible, a flow iniicator and appropriate valving vill permit l2 controlled release from tne evaporator condensate storage tanks. The j flow rate and activity or all liquids discharged from the vaste disposal ; system vill be indicated and alarmed. The high activity alarm vill 2 l actuate an interlock to stop the discharge in the event of excessive activity release. 11.1.2.2 Solids Waste Disposal System i Solid vastes vill be placed in ICC-approved containers for the vaste l material. Loaded containers vill be =enitored for surface radiation I I levels and stored in a shielding area prior to shipment to an off-sito disposal facility. .. Evaporator concentlate frem the evaporator that does not contaia re- , psable boric acid vill be pumped into a shipping cask for off-site disposal. l l d Opent ' resins frem the de::dneralizers will be sluiced to a spent resin .i 7 gQh ll k (aevised 10-2-67) \_
i l l storage tank, and the sluice water vill be transferred from the tank to the miscellaneous vaste holdup tank. The spent resin storage tank vill hold one complete charge of resins from the reactor auxiliary systems. Spent resin vill be transferred from the storage tank to special drums for disposal. In the drums, radioactive resins vill be mixed with concrete and vermiculite and allowed to solidify. The activity of the spent resins and the shielding capability of the drum and shipping cask vill determine the mixture proportions in the drum. The insulating properties of concrete and vermuculite protect the drum from excessive gamma heating. Other miscellaneous solid vastes such as filters, clothing, laboratory equipment, pieces of metal, and paper vill be disposed of using a baler and light metal shipping containers. 11.1.2.3 Gaseous Waste Disposal syatem Gaseous vastes vin be removed continuously from both evaporator chains ) during the concenuation operation. Vacuum pumps vill maintain a constant j vacuum in the batch tanks and evaporators and vill draw off vnste gases to the gas decay tanks via the vaste gas moisture separators. Gaseous vastes in the decay tank vill be monitored for activity and held up for decay and then released at a predetermined rate through absolute and l charcoal filters to the plant vent. A high activity alarm and indicator l vill be located in the discharge line and provide automatic shut off I of releases at a preselected level. 1 11.1.2.h System Radiation Monitoring The cooling water systems that remove heat from potentially radioactive sources vill be monitored to detect accidental releases. A radiation monitor vill be located in the intermediate cooling loop, the nuclear services closed cooling loop, the spent fuel cooling system, and in the liquid vaste discharge header. In addition, a monitor win be located in the plant effluent line before final discharge inwo the river. Reactor coolant letdown flow vin be monitored to detect a gross fuel I assembly failure. A smaller fuel assembly leak vill be detected by regular laboratory analysis of reactor coolant satples. Air samples from the reactor building and the station vent 5:in be monitored for particulate, iodine, and gaseous activity. These radiation monitors are comr.ercially available equipment. The required characteristics win be establir.hed during detailed station design. The maximum sensitivity of detectors when combined with appropriate dilution factors vill insure safe limits of release. ) l 000i. ?3y a, i k l,
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11-5 (Revised 10-2-67) I l I
1 f 11.1.2.5 Design Evaluation l All analyses on liquid and gaseous vaste disposal vill be perfor=ed on the O 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 de=enstrate the capability of safet station opera- - tion within the limits of 10 CFR 20 with quantities of radioactive fission products in the system. A s*w/ of the various operatiens considered in the analyses , and the total concentrations resulting in the station effluent from operation of the unit with failed fuel, are given in Table 11-5 The activity concentrations result-ing are given as fractions of the Maximum Permissible Concentration (MPC) for unrestricted areas , i.e. , the concentration of each radioactive 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 nomal mode of plant operation vill be to treat contaminated wastes in the vaste disposal system and store the evaporator condensates in the eva-porator condensate storage tanks for reuse as demineralized water supply. However, to demonstrate safety in the event of abnormal operation of the plant, the effects of liquid vaste discharges to the environment vere ana-lyzed for several situations. The first operation considered in the analysis was the continuous release of liquid vastes corresponding to reactor coolant letdown to storage for boric acid reduction. These vastes were processed through the vaste disposal system without holdup or decay and the evaporator condensates were discharged to the river after mixing with the cooling tower blevdown effluent. The activity level in the station effluent was determined by assuming that reactor coolant system liquid was processed through the eva-porator drain at the average coolant bleed rate of 25 gph and the condensates were discharged centinuously for a period of 278 full power days. Letdown through the purification demineralizer was assumed to give a decontamination factor of zero for cesium and 100 for all other nuclides. In addition, a decontamination factor of 100 for cesium was assumed by pusing the liquid vaste through a cation demineralizer installed downstream of the reactor coolant evaporator. k Decontamination in the evaporator unit was assumed to be 10 . The dilution flow vas the cooling tower blevdown of 2000 gpm. The resulting yearly average concentration in the station effluent was 0.005 of the MPC. This demonstrates that activity levels in the station effluent are significantly less than MPC even in the event of continuous vaste releases. In addition to the continuous release analysis, the effect of an inadvertent release of liquid vaste was considered. It was assumed that the entire contents (10,000 gal.) of an evaporator condensate storage tank vere pumped at 50 GPM into the cooling tower blevdown effluent due to operator error. This liquid was reactor coolant condensate which had been processed through the cation j iemineralizer and evaporator train. The instantaneous concentration in the cooling tower blevdown effluent was 0 5 of the MPC. An inadvertent release of rav vaste from a reactor coolant bleed tank or a miscellaneous vaste tank vas not regarded as credible since the discharge from these tanks cannot,be released to the environ =ent without passing through the evapceator ' condensate storage tanks which, therefore, act as a barrier to prevedt direct release to the environment.
, 11-6 (Revised 10-2-67)
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Three reactor coolant bleed holdup tanks , each with a capacity of 11,000 ft3, will be provided for a total stora6e capacity of 33,000 ft3 The maxi-num quantity of coolant letdown for chemical shim, during any 30 day period in life, vill be approxi=ately 600C ft3 Thus, only one tank, or one-third l of the available stora6e capacity is required to provide a 30 day holdup ' period for the coolant which vill be bled down over 30 days. The maximum volume of coolant removed during heatup and dilution to startup from a cold l shutdown vill be 1h,000 ft3 This occurs at the end of the chemical shim period. Two cold startups at this stage vould generate 28,000 ft3 of vaste. Earlier in life the quantity removed would be less than this due to the smaller amount of dilution required. 'Peo cold startups , early in life, ! vill contribute about 3,000 ft3 of liquid vastes. The remaining coolant l removed from the reactor system is the partial drain which occurs ence per ' year during refueling. The coolant is removed in a batch of 6,100 ft3 and returned to the reactor coolant system upon completion of refueling. Thus, it occupies storage capacity only during the period of refueling. The required stora6e volume for refueling operations of 6,100 ft3 is less than 20 per cent of the available capacity. l CJ n G
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000I'34g-11-6a (Revised 10-2-67) i I
i 4 It is extremely unlikely that operating conditions could occur which would n require storage for excessive amounts of liquid vastes. However, even in V the event of two cold startups toward the end of core life, the available storage capacity would accommodate the liquid vastes. This demonstre.tes that the three ta@s 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 wastes will include the miscel-laneous vaste holdup tank (2,700 ft3), the auxiliary building sump tank (hh0 ft3), and the reactor building sump (1,000 ft3), Activity levels in the miscellaneous vaste holdup tank were determined by assuming that all liquid collected in the tank was reactor coolant leakage. Collection was assumed to take place continuously at 25 gpd and the contents were processed through the evaporator chain without holdup. The condensates were discharged to the cooling tower blevdown effluent with a dilution flow of 2,000 gpm. The concentration at the point of discharge averaged over the year was sig-nificantly less than the MPC for unrestricted areas (as shown in Table 11-5) . This concentration vill .normally be far lower since it is intended to reuse the evaporator condensates as a demineralized water supply. The reactor coolant and miscellaneous vaste handling 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 O analyzed on this conservative basis. It is concluded that the capacity of the liquid waste disposal system vill be large enough to permit vide flexibility in station operations while providing a means for safe dis-posal of vastes with activity well below the acceptable limits. 11.1.2.5.2 Gaseous Wastes In determining the activity concentrations in the gaseous effluent, the atmospheric dilution was computed using the model for release as described in Section 2.3. Concentrations were calculated at the exclusion distance under the long term release conditions. The collection of gaseous activity was determined for those components 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) dilution and expansion following shutdown and startup. O 4
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000i 20 ^ 117 (Revised 10 2-67) gIkI (,
For the case of centinuous 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
!anks. The coolant activity levels were those cceputed at the end of the second h
core cycle with 1 per cent failed fuel. Before reaching the reactor coolant blesi holdup tanks, the letdown flow was taken through the demineralizers assuming a 99 cer cent removal efficiency for iodine. "'he activity was released to the ataosphere, without holdup, at a rate equal to the average shis bleed rate over life of 25 gph. Releasing activity at this rate, the total fraction of the !@C st the exclusion distance of 2000 feet is .05. With a 30 day holdup in the reactor coolant bleed tanks, the concentration is reduced to about .005 of GC. In the case of unit shutdevn and startup, it was postulated that a cold shut-down occurred at a time in lifetime just prior to beginning the use of the deborating de=ineralizer for boric acid removal. This results in the =aximum quantity of coolant bleed during shutdown. As a result of this operation a bleed quantity of 9600 ft3 is produced. Letdown through the de= w ralizers with a removal efficiency of 99 per cent for iodine was assu=ed. As *he coolant is let devn to the bleed holdup tanks , all of the Kr, Xe , and I is assuced to come out of the water and go into the vaste gas decay tank. With a decign pressure of 150 psi and a volume of 1500 ft3, the vaste decay tank can hola the total gas volume displaced by this quantity of coolant bleed. "'he gas dis-placed frem the bleed holdup tank approximately 9600 ft3 vould only pressurize the vaste decay tank to about 100 psig. The gaseous activity could 1; hen be discharged over a period of one week to allow dispersion in accordance with the 1cng term atmospheric diffusion model. The average annual concentration at the exclusion distance, after a holdup of 30 days in the gas decay tank, would be about .0007 of the MPC. Two cold startups toward the end of core life vould produce an average annual concentration of about .0014 of the MFC. The gaseous concentrations in the makeup tank void were determined from Henry's Law assuming the tank gas space is in equilibrium with the reactor coolant. The fraction of activity in the reactor coolant system which collected in the
=akeup tank was approximately h5 per cent for Kr, 30 per cent for Xe, and 0.lh per cent for I. The activity levels used for sources in the makeup tank cor-respond to the reactor coolant system activity at the end of the second core cycle. It is assumed that the tank vill be vented once a year to the vaste decay tank. The volume of gas in the makeup tank is about 300 ft3 at 45 psia.
This gas vould only increase the vaste gas decay tank pressere 10 psi. This gas can be discharged to the atmosphere over a period of one week to ena2re dispersion in accordance with the long-term atmospheric diffusion model to give an a~erage annual concentratica of about .0005 of the MPC at the exclusien dis- l tance with no allowance for decay. l Calculations similar to those used for the makeup tank were performed to determine the activity in the pressurizer. It was found that the activity in the pressuri:e Venting of the was approximately pressuri:er one-third results in the 60 only about activitf ft of ngas i the, makeup which cantank. be released from l l the vaste decay tank over a period of one week to give a yearly average concen-I tration of about .0001 of the MPC at the exclusion distance. The activity level in the reactor building at=osphere was ecmputed assuming a reacter coolant system leakage to the reactor building air of 10 gpd. All
,,c of 4 ::co(the lant iKr and Xe, system, wasand 50 perthroughout dispersed cent of thetheI and Cs that reactor leaked building frem the reactor atmosphere. l, ,
11-0 (Revised 10-2-67) 001 2 G
l Activity buildup in the reactor building was computed over the 30 days of fuel leakage, i.e. , it was assumed that no purge had been =ade for 30 days. O- This quantity of activity was then discharged to the atmosphere, without decay, by way of the reactor building purge system. The concentration at the exclusion distance of 2000 feet averaged over 30 days was ecmputed to be 0.02 of the MPC. Venting the reactor building once each 30 days vould give an avera6e yearly concentration of 0.02 of the MFC, at the exclusion boundary. This calculation was based on the use of the short term dispersion model discussed in Section 2 (Table 2 h). A preliminary analysis has been =ade to examine the consequences of reactor cperation 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 vacuum pump exhaust to the station vent to remain within the allovable dischar, limits of 10 CFR 20. At this limiting concentration, the direct dose rate from the condenser is below the permissible value for centinued access. In the vacuum pumps exhaust, the controlling isotope is xenon-133. The analys: 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 conservative assumption, a reactor coolant leak rate of 1 gym results in a concentration of 0.06 of the MPC at the exclusion distance. The analysis was based on 1 gpm tube leakage continuously over a year. () 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 buildin6 purge, venting of the letdown storage tank, startup expansion and dilution, chemical shim bleed, and steam generator tube leaka6e is a maximum of about 0.09 of the MPC. The evaluation aJJo demonstrate that equipment capacities are adequate to accommodate and store radioactive gases as necessary. Thus, the system design is adequate to insure safe disico of gaseous vastes. 11.1.2.5.3 Radioactive Waste Disposal System Failures The possibility of a significant act vity release off the site from accidents in either the solid or the liquid vaste disposal equipment is extre=ely remote Both of these systems will be located in shielded, controlled-access areas with provisions for maintaining contamination control in the event of spills or leaka6e. Solid vastes vill be disposed by licensed contractors in accord-ance with ICC regulations. Liquid vastes vill be sampled prior to discharge j and vill be monitored during discharge to insure ecmpliance with 10 CFR 20. A tabulation of potential ~ vaste disposal system failures and their consequence: ! is presented in Table 11-6. l I Radicactive gases vill be sampled and discharged in compliance with the re . quirements of 10 CFR 20. In the event of vaste decay tank failure, these . gases vould be released to the decay tank compatment, and then released to the station vent via the normal ventilation System. , l CE) 0001'245
- Is*:s s,.;
11-9 (Revised 10-2-67)
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'he maxi =um activity in a vaste gas decay tank vill occur folleving a boron
.ilution cycle durin6 reactor startup just prior to switching to deborating I
.emineraliser for boron removal. The reactor cuolant water activity used :
'or the analysis assumes prior operation for an extended period with failed l
Mel rods , equivalent to exposure of 1 per cent of the fuel. Approximately 1
.600 ft3 of reactor coolant would be let down at this time. It is assumed i hat the purification demineralisers have a removal efficiency of 99 per ;
ent for iodine and zero removal efficiency for noble gases. The remaining l aseous activity vill be carried with the water to the reactor coolant bleed oldup tanks , where it is assumed that the gases are immediately released
'rcm tne water and carried with the purge gases to the vaste gas decay tank. l his assumption is quite conservative since the gas release rate vill occur
.ue to diffusion from the surface in accordance with Henry's Law and occur
'ver a considerable time period. Similarly, it is conservatively assumed hat the gases do not undergo radioactive decay after leaving the reactor colant system. With these assumptions , the following activity is calculated o exist in the vaste 6as decay tank:
Isotone Total Curies Kr 85M Shh.0 Kr 85 h,200.0 Kr 87 300.0 Kr 88 1,000.0 I 131 90 I 132 13.h 12.2 W I 133 [ I 13h 1.5 l I 135 57 Xe 131m 570.0 Xe 133m 870.0 t Xe 133 79,000.0 Xe 135m 272.0 l Xe 135 2,550.0 Xe 138 136.0 he rocm containing the vaste gas decay tanks vill be ventilated, and the
.ischarge vill be to the station vent. The activity frca a vaste gas tank
'ailure is assumed to be released frcm the vaste gas tank rocm at a lov ontrolled rate to the station vent. The discharge from the station vent s conservatively assumed to mix in the wake of the building structures.
.tmospherie dilution is calculated using the two hour meterological model
.iscussed in Section 2 (Table 2 h). The total integrated dose to the whole ody at the exclusion distance is 0.3 rem, and the thyroid dose at the same
.istance is 0.h rem. 'Ihese doses are well below the guideline values of 0 CFR 100.
1 y a,, . 0001 244 11-10
11.1.3 TESTS AND INSPECTIONS O Functional operational tests and inspections of the Waste Disposal System vill be made as required to insure performance consistent with 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 ' Plant operating personnel and the general public must be protected by radiation shielding wherever radiation hazards exist. Protection vill be in accordan:e 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 smd to the general public is within the limits set forth in 10 CFR 20 at- (2) to provide the necessary protection of operating personnel folleving 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: Full Power Oteration Conditions (l". failed fuel) Location Dose Rate, mrem /hr Office, Control Room, and Turbine Building 0.50 Reactor Building: Accessible Areas 25.0 Auxiliary Building: Accessible Areas 1.0 - 2.0 l Maximum Hyecthetical Accident Conditions Location Dose Rate, mrem /hr Inside Control Room 3 rem integrated whole body dose over 90 days Outside Reactor Building )
) See Section lh (Safety Analysis)
) for Integral Dose Rate Curves i Site Boundary ) I O .!1
~
u. 11-11 (Revised 10-2-67)
-0001 245 'I I l
1
]
. 2.1.2 Description of Shielding
. 2.1.2.1 Primary Shield te primary shield vill be a large mass of reinforced concrete surrounding ie reactor vessel and extending upward frem the reactor building floor i form the valls of the fuel transfer canal. The preliminary shield tickness is 5 ft up to the height of the reactor vessel flange where te thickness is reduced to k.5 ft. The primary shield vill meet the 311cving obj eetives :
I a. To reduce, in conjunction with the secondary shield, the radiation level frca sources within the reactor vessel and reactor coolant system to allov limited access to the reactor building during normal full power operation,
- b. To limit the radiation level after shut down frcm sources within the reactor vessel .to pe:=it limited access to the reactor coolant system equipment.
- c. To limit neutron flux activation of cmponent and structural materials.
Le neutron and gamma-ray heating of the priman shield vill be dissipated the concrete shield cooling system. The pri=ary shield concrete vill be
>oled to maintain temperatures less than 150 F.
. 2.1.2.2 Secondary Shield O l te secondary shield vill be a reinforced concrete structure surroundi:2g the tactor coolant equipment including piping, pumps , and steam generators. This i
tield vill protect personnel frcm the direct gamma radiation resulting from ractor coolant activation products and fission products carried away frem te core by the reactor coolant. In addition, the secondary shield vill supplement l te primary shield by attenuating neutron and gamma radiation escaping from te pri=ary shield. The secondary shield vill be sized to allow limited access i the reactor building during full power operation. The preli=inary thickness
' secondaq shield valls is h.5 ft.
. 2.1.2.3 Reactor Building 3hield te reactor building shield vill be a reinforced, prestressed concrete containment
.ructure which empletely surrounds the nuclear steam supply system. At til power operation, this shield will attenuate any rediation esesping frem te primary-secondan shield ccztplex such that radiation levels outside the tactor building vill be less than 0 5 mr/hr. In addition, the reactor building
.ructure vill shield personnel frem radiation sources inside the building illevi.2g a Maximum Hypothetical Accident (MHA) . The shielding vill be of tfficient thickness to allow personnel a reasonable time period in which i evacuate the immediate vicinity of the reactor building following the MHA
.thout excessive radiation exposure. The curves in Section lh (Safety Analysis)
,dicate an integrated direct dese of 9 rem cist a period of two hours immediately, itside the reactor building folleving the MHA. Preliminarf thicknesses of e reactor buildin6 vall and dcne are 3.5 ft and 3 ft respectively.
N' U ,0
^~'0 0 0 1 2 4'6 ~
11-12
11.2.1.2.4 Control Room Shield O The control room shielding vill be designed for continuous occupancy for essential centrol room personnel following a Maximum Hypothetical Accident. This would 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 MEA vill not exceed 3 rems. Ventilation of the control room under post-accident conditions will be controlled as described in Section 9.8.2. - n . 2.1. 2. 5 Auxiliary Shield Auxiliary shielding vill include all concrete vans, covers , and removable blocks which vill shield the numerous sources of radiation occurring in the radioactive vaste disp'osal, makeup and purification, chemical addition and sampling systems. Typical components which require shielding include vaste holdup 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 vin be provided for protection during all phases of spent fuel removal and storage. Operations requiring shielding of personnel are spent fuel -emoval from reactor, spent fuel transfer through refueling canal and transfer tuber., spent fuel storage, and spent fuel shipping O cask Icading prior to transportation. Since all spent fuel removal and transfer operations vill 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 reactor cavity and fuel transfer canal and 13 ft over stored assemblies in the spent fuel storage area. The dose rates at the water surface vill be less than 10 mrem /br. The concrete valls of the fuel transfer canal and spent fuel pit vin supplement the water shielding and vill limit the maxi =um continuous radiation dose levels in vorking areas to less than 2.5 mrem /hr.
"he refueling water and concrete valls also provide shielding from activated control red clusters and reactor internals which vill be removed at refueling' times. Although dose rates vill generally be less than 2.5 sre=/hr in voiking areas , certain =anipulations of fuel assemblies ,
rod clusters, or reactor internals =ay produce short ters exposures 11 excess of 2.5 =re=/hr. Eevever, the radiation levels . rill be c1csely senitored during refueling operations to establish the allovable exposure ti=es for plant personnel in order not to exceed the integrated deses specified in 10 CIR 20. 11.2.1.2.7 Materials and Structural Require =ents
~
ne =aterial used for the primary, seconda y, reactor building, and , auxiliary shields vill te crdina y cencrete with density of approx 1=ately * , Os y lb /*t3 Since the pr'-= y and secondarf shielding valls serve as the r.
~
.000.1 247' ff '- =-u ae.-1,ed =-2-e:
efueling structure, give support for the reactor coolant ecmponents under ipe rupture conditions, and provide missile shielding, they will be rein-g orced and designed to be self-supporting. imes. of occupancy in restricted areas vill vary depending on measured adiation levels in each area. Such areas as containment operating ficor, eactor vessel head prior to refueling, primary loop ccupartments after nutdown, and spent fuel handling areas vill be surveyed prior to access nd a time-limited work schedule vill be, set up. 1.2.1.3 Evaluation 1.2.1.3.1 Radiation Sources he shielding vill be designed to attenuate neutron and gamma radiation sanating frcm the following basic sources :
- a. Reactor Core, Internals, and Reactor Vessel
- b. Reactor Coolant Loops
- c. Radioactive material released during accidents
- d. Auxiliary Systems Equipment e.
e Spent Fuel Elements ource magnitudes are detemined for the reactor operating at the aximum expected power level of 2535 MWt with reactor coolant etivity levels corresponding to 1 per cent failed fuel. Gamma-ray ield and spectral distributions frca prcmpt fission and gross fission reduct activity are based on information in Volume III, Part B, of he Reactor Handbook. The yield and spectral data for capture gammas re taken frcm ANL-5800, Reactor Physics Constants, and the Reactor andbook. Data on activation product gamma rays are derived primarily rcm the Review of Modern Physics , Vol. 30, No. 2 ( April 1958) . The reduction of N-16 in the reactor coolant is calculated with a B&W ode which ecmputes the integral of the 0-16 (n,p) N-16 cross section ver the neutron flux in a water-cooled reactor, subject to variables in colant flow and density and in neutron flux spectra and magnitude. The
-16 (n,p) N-16 cross section used is that reported in WAPD-BT-25 .ctivities of individual fission products in the core, reactor, coolant, nd reactor auxilialy systems are determined by a B&W ccuputer code esigned to predict activities frca a five-member radioactive chain .t any point in the core history. Fission product leakage frca the ore to the coolant and removal frca the coolant by purification and eaka6e are calculated.
! .1.2.1 3.2 Calculation Methods eutron and Gamma Shields he pri=ary shield preliminary thickness is based en work performed for . O he Oconee Nuclear Station using Babcock and Wilcox ccuputer codes which o the negrpn and gamma-ray attenuation equations for the multi-layer ' *, , ry. . 0001 248
.,, 11-lh
l l source-shield ecmplex. 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 cods 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 vere 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 equations for the direct gamma flux from the simpler gecanetric 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 line-of-sight from the source point to the dose point and ccmputes the gamma flux by summing over the source distribution. Secondary gamma-ray penetration was calculated using a Secondary Gamma program for a laminated, semi-infinite shield array. The aforementioned B&W codes and techniques are described in IDO-2hh67. Gilbert Associates, Inc. vill perform the shielding calculations for final sizing of all radiation shielding. 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 1 per cent of the solid fission products were instantaneously released to O 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 frem the reactor building. The results are given in lb.2.2.h. 11.2.1.3.h Operating Limits The radiation shielding design, including heating and dose rate profiles, temperature distributions , 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 gamma 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. O .
.e 's 0'001 749 -
11-15 (Revised 10-2-67)
L1.2.2 AREA RADIATION MONITORING SYSTE4 Ll.2.2.1 Design Bases, 3:e fixed radiation monitoring system for the Three Mile Island Nuclear teactor Facilities vill be designed to indicate, record, and alarm high l adiation monitoring levels throughout the station. The system vill be
- cmprised of radiation detectors and allows for visual presentation of
.eadings, recorded presentation, and an audible / visible alars at the letector location and the Control Room. All instrumentatien for the ra-listion monitorin6 system vill obtain its voltage supply frcm the vital
'nstrumentstion bus and each detector will have a " Loss of Power" alarm.
'he normal radiation alert alarm setpoint will be above the normal oper- [
ttional retding of the detector. A maximum alarm point vill be set to
- orrespond to the MPC value specified .in 10 CFR 20. The maximum alarn
>oint set at 10 CFR 20 values could be either an actual value or a calcu-
.ated number corresponding to 10 CFR 20 limits.
.1.2.2.2 Description u es Gamma Monitoring Jetectors are located as follows:
- a. On each of the fuel handling bridges inside the Reactor Building,
- b. Inside the Reactor Building near the personnel access hatch.
- c. Near incore instrument termination space inside the Reactor Building. O
- d. On fuel handling bridge in Auxiliary Building.
- e. Auxiliary Buildin6 decay heat removal pump area.
- f. Auxiliary Building near reactor coolant waste pump area.
- g. Auxiliary Building near makeup tank.
- h. Auxiliary Buildin6 near intermediate cooling pumps,
- i. Radio-Chemistry laboratory.
t J. Cable Relay Room.
- k. Contaminated =achine shop.
- 1. Control room.
- m. Sample Sink.
- n. Reactor Building Dome l
G m vt, m, m u. 0001 m . 11-16(Revised 10-2-67)
Atmospherie Menitoring Detectors are located as follows:
- a. Reactor Building Purge Duct
- b. Auxiliar/ and Fuel Handling Building Exhaust Duct
- c. Control Room Ventilation Duct
-4
- d. Fuel Handling Ventilation Duct
- e. Auxiliar/ Building Ventilation Duct
- f. Reactor Building Air Sample Line
- g. Condenser Vacuum Pump Exhaust
- h. Site Monitors (2) at 2000 ft site bcundar/
- 1. Sample Sink
- j. Radio-Chemical Laboratorf
- k. Spent Fuel Area
- 1. Waste Gas Decay Tank Discharge Licuid Monitoring Detectors are located as follows:
- a. Letdown Coolant
- b. Intermediate Cooling Water
- c. Nuclear Services Closed Cooling Water
- d. Spent Fuel Cooling Water
- e. Plant Liquid Effluent Line
- f. Liquid Waste Discharge Header i
l l o ' 0001 251 11-16a (Revised 10-2-67)
O DELETED Detector ranges vill be determined depending upon the normal background at the detector locations and the calculated levels for abnormal conditions. Radioactive test sources vill be available to allow the overall system per-formance to be verified at regular intervals. , 11.2.2.3 Evaluation Area radiation monitor detectors vill be located on each of the fuel handling bridges to varn 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. l The upper range of - he 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 vill 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 O- detector to alarm an abnormal condition in connection with system sampling. Alazus will be actuated in the control room and at the detectors if an abnormal change in radiation background occurs. l TL; -diation monitoring system shall be checked and calibrated at least j once per monti.. When any portion of the radiation monitoring system requires maintenance, that unit shall be completely checked and calibrated l immediately after completion of maintenance. 1 11.2.3 HEALTH PHYSICS i l The station superintendent is responsible for radiatica protectica and 1 l contamination control for the Three Mile Island Facility. This responsi- f bility is, in turn, shared by all supervisors. All persoc.nol assigned to l the station and all visitors will be required to follow rules and procedures I established by administrative control for protection against radiation and contamination. O ' 0001 252 i l ff % ,
. t. ,
11-17 (Revised 10-2-67)
ne administration of the radiation protection progra= vill be the responsi-111ty of the station Health Physicist. It vill be the responsibility of lll ae Health Physics section to train station perscnnel in radiation safety; o locate, measure, and evaluate radiological problems ; and to make recem-endations for control or eli=ination of radiation hazards. The Health aysics section vill function in an advisory capacity to assist all personnel a carrying out their radiation safety responsibilities and to audit all ; spects of station operation and =sintenance to assure safe conditions nd compliance with the AEC and other federal and state regulations concer-ing radiation protection. iministrative controln vill be , established to assure that all procedures nd requirements relating to radiation protection are folicwed by all station ersonnel. The procedures that control radiation exposure vill be subject o the same review and approval as those that govern all other station recedures (see Section 12.5, A6sinistrative Control). These procedures ill include a Radi.ation Work Permit system. All verk on systems or locations aere exposure to radiation or radioactive materials is or may be involved ill require an appropriate' Radiation Work Per=it. RADIATION WORK PEPNITS l Radiation Work Permit shall be obtained by all personnel prior to entering Control Area or performing any work en radioactive or contaminated material r equipment. n the event that the safety of the plant or its personnel are endangered, atry may be made into a Control Area simultaneously with monitoring per-onnel. A Radiation Work Pernit shall be completed es soon as possible fter correction of the casualty. adiation Work Permits shall be issued routinely by the Shift Foreman. dese permits shall shev:
- a. The nature of the work to be performed.
- b. Expected duration of vork.
- c. Names of persons to perform the work.
- d. Signature of authorizing Foreman.
- e. Signature of an individual from the Health Physics Group who shall ensure that:
i
- 1. Designated personnel are within their permissible expcsure limits.
l 2. The area has been adequately surveyed prior to entry.
- 3. Adequate protective clothing and supplies are available at the
.,, , control point.
l
- h. Monitors are available for the work. , ,.
t-11 such pe= nits shall be filed by Health Physics for future reference. ' - 11-18
11.2.3.2 Personnel Monitcring System O The personnel monitoring program for the Three Mile Island Nuclear Station t shall insure that the reccamendations and regulations of the Atcznic Energy i Camission are followed for all involved personnel. All personnel enter-ing a Centrol 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 frquent access to these areas vill have pocket chambers , self-reading dos- j imeters , pocket high-radiation alams , vrist badges , and finger tabs readily m; i available for use, when required by station conditions. This personnel monitoring equipment will also be available on a day-to-day basis for those l persons , empicyees, or visitors not assigned to the station who have occassic ! to out.er Restricted Areas or to perfom vork involving possible exposure to radiation. Records of radiation exposure history and current occupational exposure vill be' maintained by the Health physics section for each individua: for whcm personnel monitoring is required. The externt.1 radiation dose to personnel vill be determined on a daily and/or weekly basis , as required, 4 by means of the pocket' chamber and desimeter. Film badges will be processed monthly or more frequently when conditions indicate it is necessary. 11.2.3.2 Personnel Protective Ecuiment Special " protective" or "anticentamination" clothing vill be furnished and vorn as necessar/ to protect personnel aGainst centact with radio-active contamination. Change Rocms vill be conveniently located for proper utili:stion of this protective clothing. O Respiratory protective equipment vill also be available for the protection of personnel against airborne radioactive contamination and vill consist of full face filter masks , self-contained air-bresthing units , or air-supplied masks and hoods. The first line of defense against airborne contamination in the work area is the ventilation system. Ecvever, respiratory protective equipment vill be provided should its use become necessary. Maintenance of the above equipment vill be in accordance with the manu-facturer's recccmendations 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 naintenance of this equipment vill be under the direct centrol of the Health Physics sectica, and personnel vill be trained in the use of this equi; scent before using it in the perfemance of vork. 11.2.3.3 Change Roce Facilities Change recze facilities vill be provided where personnel may obtain clean protective clothing required for station work. These facilities vi.' ' - divided into " clean" and "centaminated" sections. The " contaminate <1 secticn of the change reces vill be used for the removal and handling c.. centan.inated protective clothing after use. Shcvers , sinks , and necessar/
=cnitoring equi;=ent also vill be provided in the change areas to aid in the decenta=;ne.tien of perscenel.
O 0001 254
- w. . n. .
11 'J
Equip =ent decentamination facilities will also be provided at the station for large and s=all ite=s of plant equipment and ec=ponents. g Provision vill also be made for deconta=ination of work areas throughout the station. Appropriate written precedures vill govern the proper u.- of protective clothing; vbere and how it is to be worn and receved, and how the change rocm and decontamination facilities for personnel, equipment, and station areas are to be used. In order to protect personnel frcs access to high radiation areas that may exist temporarily or semipemanently as a result of station operations and maintenance , warning signs , audible and visual indicators , barricades , t.nd locked doors vill be used as necessary. Administrative procedures vill also be written to control access to high radiation areas. The Radiation Work Pemit System vill also be utilized to contrei access to high radiatica areas. 11.2.3.4 Health Physics Laboratory Facilities The station vill include a Health Physics Laboratory with facilities and equi; ment for detecting, analyzing, and measuring all types of radiation and for evaluating any radiological problem which may be anticipated. Counting equipnent (such as G-M, scintillation, and propertional counters) vill be provided in an appropriate shielded counting rocm for detecting and measuring all types of radiation as well as equipment (such as a multi-diannel analyzer) for the identification of specific radienuclides. Equip. h ment and facilities for analyzing environmental survey and bicassay samples vill also be included in the Health Physics Laboratory. Maintenance and use of the Health Physics Laboratory facilities and equipment vill be the responsibility of the Health Physics section. 11.2.3.5 Health Physics Instrumentation Portable radiation survey instruments will be provided for use by the Health Physics section as well as for cpernting and maintenance personnel. A variety of instrue.ents vill be selected to cover the entire spectrum of radiation nessurement problems anticipated at the Three Mile Island Nuclear Station. Sufficient quantities will be obtained to allev for use, calibra-tion, maintenance, and repair. This vill include instru=ents for detecting and measuring alpha, beta, ga==a, and neutron radiation. In addition to the portable radiation monitoring instruments , fixed monitoring instruments , i.e. , count rate =eters , vill be located at exits frem restricted areas . These instruments are intended to prevent any contaminatien on personnel, l material, er equipment frem being spread into unrestricted areas. Appro-priate =enitoring instruments vill also be available at various locations within the restricted areas for contamination centrol purposes. Portal monitors vill also be utill:ed, as appropriate, to control personnel egress frem restricted areas or frcm the station. O W iu 0001 255 11-20
The station vill have a permanently installed remote rsdiation and radio-activity monitoring system for locations where significant levels can be O expected. This system vill monitor r.irborne particulate and gaseous radioactivity as well as external radiation levels. This system vill presen-an audible alarm and radiation level indication in the area of concern in addition to the control rom. 11.2.3.6 Medical Pregrams Facilities and counting equipment for screening personnel for internal exposure vill be available on site with outside services utilized as backup 4 , and support for this program. A caprehensive medical examination program appropriate for radiation workers will be conducted to establish and maintain records of the physical status of each employee at the Three Mile Island Nuclear Station. Subsequent medical examinations will be held as determined necessary for radiation vorkers. Medical doc. tors , preferably in the local area, will be used for this program. The Health Physics section vill be responsible for the program and vill assist the physicians in maintaining medical control of personnel. This program vill be designed to preserve the health of the unployees concerned and to confirm the radiation control methods enployed at the station. 11.3 REFEREiCES (1) Frank, P. W. , g g. , Radiochemistry of Third PWR Fuel Material Test - X-1 Loop NRX Reactor, WAPD-m-29, February 1957. (2) Eichenberg , J. D. , g g. , Effects of Irradiation on Bulk UO , WAPD-183, Oc".1ber 1957. 2 (3) Allison, G. M. and Robertson, R. F. S. , The Behavior of Fission Products in Pressurised-Water Systems. A Review of Defect Tests on UO Fuel Elements at Chalk R$ver, AECL-1338, 1961. 2 (h) Allison, G. M. and Roe, H. K. , The Release of Fission Gases & Iodines From Defected UO 2 Fuel Elements of Different Lengths , AECL-2206, June 19 (5) Duke Power Company, Preliminary S*.fety Analysis Report, Volume II,1966. O 000l 256 11-21
Table 11-1
/
Radioactive Waste Quantities Waste Source quantity per Year
- Assumptions and C Liquid Waste Ft 3 Reactor Coolant System:
Startup Expansion 17,000 h Cold Startups Startup Dilution 11,700 2 Cold Startups at begin: life and 1 cold startup and 200 full power days tively. Lifeti.ne Shim Bleed 23,500 Dilution from 1460 to 17 System Drain 6,100 Drain to level of outlet for refueling operations Sampling and Laboratory Drains 3,000 12 souples per week at 5 per semple Purification Demineralizer Sluice 160 60 P 3/ year replacement i I 2 ft 5/ft3 resin sluice. Spent Fuel Pool Demineralizer Sluice h2 21 ft /3 year replacement ft3 resin sluice Deborating Demineralizer Regen-eration and Rinse 2,500 1Regenergtionperyear4 20 ft 3 /ft resin regener: Miscellaneous System Leakage 6,000 5 sph leakase Laundry 7,300 150 gpd Showers 15,C00 10 showers per day at 30 per shower 1 Gaseous Waste (a) i Off-Gas from Reactor Coolant System 1,350 Degas at 25ccHp per lite: ! cencentration l 1 Off-Gas frem Liquid Sampling Th Degu at 25ccH P'# lit
- concentration 2 *! ,
Off-Gas frca Letdown Storage Tank 900 Vent once per year Off-Gas from Pressurizer 60 Vent once per year , c ., - '; 0001 257 ! l 11-22 (Revised 10-2-67) ! /
Table 11-1 (Cont'd) lid Vaste O urification Resin 80 Resin replacement twice per year pent Fuel Pool Demineralizer Resin 21 Resin replacement twice per year sperator Condensate Demineralizar Resin 2 Resin replacement twice per year sporator Bottoms 800 Concentrated to 20 per cent solid Excludes reactor building and station ventilation. O 1 1 O
~
11-23 0001 258
Table 11-2 Escape Rate Coefficients for Fission Product Release _ Escape Rate Coefficient, Element see -1 Xe 1.0 x 10-I Kr 1.0 x 10-I Br 2.0x10-{0 2.0 x 10- _._ Cs 2.0 x 10-0 Rb 2.0 x 10 4 Mo h.0 x 10-9 Te h.0 x 10-9 Sr 2.0 x 10-10 Ba 11 Zr 2.0 x 10 11 Ce and other rare earths 1.0 1.0 xx 10 10-11 Table 11-3 Reactor Coolant Activities From One Per Cent Defective Fuel Isotore Activity JAc/ml Isotope Activity.uc/ml Kr 85m 2.0 I 131 3.3 Kr 85 15.5 I 132 h.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 Xe 133 290.0 La 1h0 0.026 Xe 135m 1.0 Y 90 0.0007 Xe 135 9.h Y 91 0.00 h3 Xe 138 0.5 Ce lhh 0.0027 O
; . /. I['
- l.', 11-2h (Revised 10-2-67) hf 2 $f)
1 I Table 11 4 Waste Disposal System Comrenent Data O i ractor Coolant Drain Tank l Nu=ber 1 ! Volume, eu ft 1000 l Material Carbon Steel, Corrosion-Resistant Lining 3borating Demineralizer t. i Number 2 Type Semiautomatic Regeneration Material Carbon Steel, Corrosion-Resistant Lining 3 actor Coolant Bleed Holdup. Tank l Number 3 Volume Each, eu ft 11,000 2 Material Carbon Steel, Corrosion-Resistant Lining Lscellaneous Waste Holdup Tank I Number i Volume, eu ft 2700 Material Carbon Steel, Corrosion-Resistant Lining iste Neutralization Tank Number i Volume, eu ft h00 Material Carbon Steel, Corrosion-Resistant Lining sent Resin Stora6e Tank l Number 1 Volume, eu ft h50 Material Carbon Steel, Corrosion-Resistant Lining taporator Condensate Storage Tank 2 (Reactor Coolant Condensate)
- Number 2 I Volume, eu ft 1500 Material Carbon Steel, Corrosion-R'esistant g Lining w 11-25 (Revised 10-2-o7)
Table 11 h (Cont'd) Evaporatcr Condensate Storage Tank (Miscellaneous Waste Condensate) ' Number 2 Volume, cu ft 500 Material Carbon Steel Corresion-Resistant Lining Waste Evaporator (Reactor Coolant Waste) Number 1 Process Rate gpm 7.5 Material Stainless Steel Vaste Evaporator (Miscellaneous Waste)
~
Number 1 Process Rate, gpm 2 Material Stainless Steel Evaporator Condensate Demineralizers Number 2 Material Stainless Steel Reactor Building Sump Pump Number 1 Capacity, gpm 200 Material Stainless Steel Waste Transfer Pump (Reacter Coolant Waste) Number 2 Capacity Each, gpm 100 Material Stainless Steel Waste Transfer Pump Number 2 Capacity Each, gym 50 Material. Stainless Steel Auxiliary Building Sump Tank Pump Number 2 Capacity F.ach, gpm 50 Material Stainless Steel
.,:t (s f- .
000,* gOI ,, 11-26 (Revised 10-2-67)
l Table ll k (Cont'd) uxiliary Buildin6 Sump Tank 2 Number Volume Material Carbon Steel, Corrosion-Resistant Lining vaporator Feed Pump (Reactor Coolant Waste) Number 2 Capacity Each, gym 7.5 Material Stainless Steel vaporator Feed Pump (Miscellaneous Waste) Number 2 Capacity Each, gym 2 Material Stainless Steel vaporator Condensate Pump (Reactor Coolant Waste) Number 2 Capacity Each, gpm Material 20 Stainless Steel g vaporator Condensate Pump (Miscellaneous Waste) Number 2 Capacity Each, 6pm 10 Material Stainless Steel vaporator Vacuum Pump 1 Number 2 l Capacity, cfm 6 Material Carbon Steel I aste Gas Compressor Number 2 Capacity Each, cfm 20 Material Carbon Steel aste Gas Decay Tank Number 2 Volume Each, cu ft Mct,arial 1500 Carbon Steel g
; if 11-27 (Revised 10-2-67) 0001 262
l Table 11-k (Cont'd)
'4aste Gas Filter Ntaber i
?/P' Pre, Absolute, and Charcoal .
1 Filter Combination ! Cation Demineralizers i
, Number 2 Material Carbon Steel, Corrosion-Resistant Lining i i !
J 4 O i .I 1 l. I O -
* ; ,' ll-27a (Revised 10-2-67)
Table 11-5 O Maximum Activity Concentrations in the Station Effluent With One Per Cent Failed Fuel Liquid Waste Yearly Average Concentrations , in Circulating Water Discharge, Oneration Fraction of MPC Lifetime Shim Bleed 0.0050 Discharge of Miscellaneous Wastes 0.0002 Gaseous Wastes Yearly Average Concentrations at Site Boundry Oeeration Fraction of MPC Lifetime Shim Bleed 0.0050 O Startup Expansion and Dilution 0.001h Ventin6 of Makeup Tank 0.0005 Venting of Pressurizer 0.0001 Reactor Building Purge 0.0200 Steam Generator Tube Leaka6e of 1 spm 0.0600 O
., , 0001 264-C# *Ia'!! 11-26 (Revised 10-2-67)
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Table 11-6 Waste Distosal System Failure Analysis O Courconent Failure Ccmment,s and Consequences eactor Building Sump Drain Fails to close Backup isolation is provided on } alve (inside or outside) opposite side of reactor building. eActor Building Drain Line Fails to open Continuous drainage is not require
'alve (inside or outside) the valve is located for =ain-tenance during operation.
eactor Building Sump Pump Fails to operate Continuous operstion is not requir located for =aintenance during operation. eactor Coolant Drain Fails to operate Continuous venting is not requirec
'ank Vent Valve relief protection is provided for tank Fails to close Vent gas is conveyed to vaste gas decay tank and discharged through filters to station vent.
faste Gas Vent Filters Rupture or lose High activity level monitored and efficiency alarmed if unsufficient station vent dilution is available. Waste gas is diverted to vaste gas deca) tanks. Taste Gas Decay Tanks Leak or rupture Building purged to station vent through filters. Tanks are pro-tected by relief valves. teactor Coolant Bleed Leak Leakage is collected in auxiliary j ioldup Tanks building drain sump for process o2 disposal; building is continuousl3 purged to station vent. vaporator Train Fails to operate Continuous operation is not requi: vaste gas decay tanks provide for vaste collection during maintenanc teborating Demineralizers Exhausted resin Spare unit placed in service unile original unit is regenerated. Startup time is increased near end-of-life depending on balance between red vorth and boric acid required. 1 . O p' w' , 0001 265 11-29 (Revised 10-2-67)
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