ML19319C263
| ML19319C263 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/01/1969 |
| From: | TOLEDO EDISON CO. |
| To: | |
| References | |
| NUDOCS 8002110726 | |
| Download: ML19319C263 (31) | |
Text
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D-B TABLE OF CONTENTS Section Pace 11 RADI0 ACTIVE WASTES 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 Sources and Quantities of Radioactive Wastes 11-1 11.1.1.3 Waste Activity 11-2 . 11.1.1.h Methods of Disposal 11-9 . 11.1.2 SYSTEM DESCRIPTION AND OPERATION 11-10 11.1.2.1 Liould Waste System 11-10 11.1.2.2 Gaseous Waste System 31-11 11.1.2.3 Solid Waste System 11-12 11.1.3 DESIGN EVALUATION 11-12 11.1.h TESTS AND INSPECTIONS 11-12 11.1.5 CODES 11-13 11.2 RADIATION PROTECTION 11-13 11.2.1 RADIATION ZONING AND ACCESS CONTROL 11-13 11.2.2 RADIATION SHIELDING 11-14 11.2.2.1 Desien Bases 11-lh 11.2.2.1.1 Radiatio 2 Exposure of Materials and Ccmponents 11-14 11.f.2.1.2 General Design Considerations 11-15 13.2.2.1.3 Specific Design Values 11-15 11.2.2.2 General Descriptions and Evaluations 11-15 11.2.2.2.1 Shield Building 11-15 11.2.2.2.2 Containment Vessel Interior 11-15 11.2:2.2.3 Auxiliary Building 11-16 0317 11-1
D-B Section Page 11.2.2.2.h Turbine Building 11-16 11.2.2.2.5 General Plant Yard Areas 11-16 11.2.3 RADIATION MONITORING SYSTEM 11_17 11.2.3.1 Design Basis 11-17 11.2.3.2 /irborne Radiation Monitoring System n-17 11.2.3.2.1 Stack Gas Monitors 11-17 11.2.3.2.2 Contairment Vessel Monitors n-17 11.2.3.2.3 Was'te Gas Monitor n-17 11.2.3.2.4 Condenser Air Ejector Monitor 11-18 11.2.3.2 5 Ventilation Systems Monitors n-18 11.2.3.3 Waterborne Radiation Monitoring Systems n-18 - 11.2.3.3.1 Radvaste Liquids Monitoring Systems 11-18 11.2.3.3.2 Component Cooling Water System Monitors 11-18 11.2.3.3.3 Reactor Coolant System Activity Monitor
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11-18 11.2.3.4 Area Radiation Monitoring System 11-18
11.3 REFERENCES
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. LIST OF TABLES Table No. Title Pg 11-1 Radioactive Waste Quantities 11 h 11-2 Escape Rate Coefficients for Fission Product Release 11-6
. 11-3 Rec.ctor Coolant Activities and Decontamination Factors Required to Meet 10 CFR 20 Standards for Water in Unrestricted Areas 11-7 11 h Tritium Activity in Reactor Coolant with Bleed Rec-*cled 11-8 O
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D-B LIST OF FIGURES ' 3 Figure No. Title
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11-1 Clean Radvaste Processing System 11-2 Miscellaneous Liquid Radioactive Waste 11-3 Waste Gas Processing System ll h Process Monitoring 3 0320
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- 11 RADI0 ACTIVE WASTES AND RADIATION PROTECTION 11.1 RADI0 ACTIVE WASTES 11.1.1 DESIGN BASES 11.1.1.1 Performance Objectives The Radioactive Waste System is designed to provide controlled handling and disposal of liquid, gaseous, and solid vastes from the Davis-Besse Nuclear Power Station. The principal design criterion is to insure that station personnel and the general public are protected a6ainst exposure to radioactive material in accordance with the regulations of 10 CFR 20.
11.1.1.2 Sources and Quantities of Radioactive Wastes The various types of radioactive vastes to be handled arc:
- a. Liquid Wastes
- 1. Clean liquid vaste ii. Miscellaneous liquid vaste iii. Detergent vaste
- b. Gaseous Waste
- i. Hydrogenated vaste gases
- 11. Aerated vaste gases i
- c. Solid Wastes The major sources of cle&n liquid vaste are bleed-off of the reactor coolant during a reduction in reactor coolant boron concentration, an increase in coolant volume due to heat-up of the reactor system, and partial replacement of reactor coolant prior to refueling.
Liquid vastes other than from the reactor coolant system are considered as detergent vastes and as miscellaneous vastes and are collected separately. The detergent vastes may contain oil and detergents. The sources of these vastes include:
- a. Sa=ple system drains
- b. Decontamination area drains
- c. Spent fuel storage area drains ,
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- d. Equipment drains
- e. Low point piping drains l 1
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- f. Containment sump
- g. Auxiliary Building Sumps 48 11-1
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- h. Component cooling system drains /
- 1. Boric acid preparation area drains J. Laundry and hot showers The sources of gaseous vastes are expected to be reactor system vents, equip-1 ment and tank vents, purging from the sampling system, the degasifier, the make-up tank, and the evaporators. There are also large quantities of displaced cover gases that may be radioactive.
The sources of solid vastes are expected to be spent demineralizer resins , filter elements and/or pre-coat material, contaminated equipment, and paper, rags , plastic sheeting, etc. used in decontamination and contamina-tion control. The estimated volumes of radioactive vastes generated during station operation : are listed in Table 11-1. 11.1.1.3 Waste Activity Activity accumulation in the reactor coolant system and associated waste handling equipment has been determined on the basis of fission product leakage through clad defects in 1 percent of the fuel. The activity levels were computed assuming full power operation of 2772 MWt for two core cycles with no defective fuel followed by operation over the third core cycle with 1 percent defective fuel. The pins that fail are assumed to have been in the core which operated for a h60-day first cycle and a 310-day second cycle at power of 2772 MWt with 2/3 of the power produced by U-235 and 1/3 by Pu-239 The quantity of fission products released to the reactor coolant during steady state operation is based on the use of " escape rate coefficients" (sec-1) as determined from experiments involving purposely defected fuel elements (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 vere 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 decay, purification and bleed. Continuous reactor coolant puri-fication at a rate of one reactor system volume per ds.v vsm med with a zero removal efficiency for Kr, Cs, Y, Mo, and Xe, and a 99 percent removal efficiency for all other nuclides. Activity levels are relatively insensitive to small changes in demineralizer efficiencies , e.g. , use of 90 percent instead of 99 percent vould result in only about 10 percent increase in the coolant activity. Removal by bleed occurs only during the first 253 days of a cycle. After this
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e time, (when the boron concentration is belov 180 PPM) deborating demineralizers are used to further reduce the boron concentration. The activity of important fission product nuclides (except tritium) are shown in Table 11-3,at various times during the third cycle in which the core continues to operate at 2772 MWt. w a - 11-2 y
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r D-B Reactor coolant bleed is taken from the downstrea= side of the purification
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demineralizers. It is assumed to have the same activity concentration as the i reactor coolant reduced by the decontamination factor of the purification demineralizers. t Gaseous activity is generated by the evolution of radioactive gases from the liquids as they are processed through the degasifier and to a lesser extent as they are stored in tanks throughout the station. The degasifier and these tanks are vented to the gaseous radvaste system. The activity of the gases is dependent on the liquid activity. Since it is not practical to remove tritium from the reactor coolant and it has a half life of 12.3 years , the concentration vill continue to build up during the life of the station assuming recycling of the coolant. The three significant sources of tritium are ternary fission, the boron reactions , and productions from the lithium used in the pH control agent. Thirty percent of 8 the tritium produced by fission is conservatively assumed to diffuse through tha clad. The boron concentration decreases with operating time as shown in Section3andthelithiumisheldataconstant2ppmof9g.9% 7Li and 0.1% 6Li. The tritium thus produced is diluted over 39,5h0 ft ,the volu=e of systens used to process and store reactor coolant,and the water loss is assu=ed to be zero during the life of the station. Table ll h shows the tritiu= concentration buildup during the first cycle and at the end of each succeeding cycle. In actual practice tritium concentration vill never reach the levels in Table ll h for several reasons. Only a small percentage of tritium is
\ expected to diffuse through the zircalloy fuel element cladding. Some of the reactor coolant vill be mixed with the refueling water. Finally some of the processed vaste vill be discharged inte the discharge water system. l8 n
.. .t.sca Amendment No. 8 o t$
11-3 M .'
D-B Radioactive Waste Quantities Table 11-1 Quantity Assumptiens . ./ Wante Source per Year and Ccements Liquid Waste Reactor Coolant System Start-Up Expansion 96,000 gal l h cold start-ups Start-Up Dilution 1h6,000 gal ' 2 cold start-ups at beginning of life, and 1 cold start-up at 100 and 200 full (ultimate) power days, respectively 72,000 gal J 2 hot start-ups at peak xenon at 100 and 200 full power days, respectively - 8 Lifetime Shim Bleed 195,000 gal J Dilution frem 1230 to 50 ppm boron System Drain (Refueling) 61,k00 gal " Drain to level of outlet nozzles System Drain (Maintenance) 8h,000 gal Incl. drain of 1 steam , generator Sampling and Laboratory 3,000 gal J 12 samples per wk at 5 gal per sample 3 Demineralizer Sluice h,500 gal J 2 ft 3/ft3 resin 8 Regeneration Wastes 15,000 gal , 20 'ft3 /ft3 resin Area Washdowns 110,000 gal . 5 gpm hose, 1 hr per day Miscellaneous System Leakage h5,000 gal " 5 gph leakage Showers and Laundry 155,000 gal ' 10 showers per day at 30 gal per shower. 120 gpd laundry Gaseous Waste Off-Gas from Reactor Coolant 3,h00 ft 3 Degas at h0 cc H p r liter 2 " System concentration Off-Gas from Liquid Sa=pling 120 ft 3 Degas at h0 cc H2 per liter concentration Off-Gas from Makeup Tank 900 ft3 vent once per year s Off-Gas from PressurizerI. 60 ft3 Vent once per year Amendment No. 8 ll-h {
D-B Quantity Asstaptions , t l Waste Scurce per Year and Comments Solid Waste Demineralizer Resin ' 300 ft3 Resin replacement once per 1, year Miscellaneous (filter Ele- 900 ft3 1-1/2 55-gal drum per week ments, Clothing, Rags, Etc.) plus 300 ft3 per refueling period 4 0
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Table 11-2 Escape Rate Coefficients for Fissien Product Release Escape Bate Coefficient, Element see-1 Xe 1.0 x 10-7 Kr 1.0 x 10-7 I 2.0 x 10-0 Br 2.0 x 10-0 Cs 2.0 x 10-0 , Rb 2.0 x 10-0 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 1.0 x 10-11 Ce and Other Rare Earths 1.0 x 10-11 O i C l - r-
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D-B TABI.E I13 REACTOR COOL. ANT ACTIVITY, M c/mi BASED ON EFECTIVE FUEL l2 Third Cycle Operating Time, effective full power days Isotope Half. Life 50 100 150 200 253 275 310 Kr85m 4.36 h 1.7 1.7 1.7 i.7 1.7 1.7 1.7 Kr 85 10.57 y 7.1 10.7 11.9 10.7 7.7 8.6 10.0 Kr 87 78 m .94 .94 .94 .94 .94 .94 .94 Kr88 2.77 h 3.0 3.0 3.0 3.0 3.0 3.0 3.0 Rb 88 17.8 m 3.0 3.0 3.0 3.0 3.0 3.0 3.0 - Sr 89 53 d .041 .044 .044 .045 .045 .045 .045 Sr 90 28 y .0031 .0033 .0035 .0037 .0039 .0039 .0041 Sr 91 9.7 h .052 .052 .052 .052 .052 .052 .052 Sr 92 2.6 h .019 .019 .019 .019 .019 .019 .019 Y 90 64.8 h .13 .29 .46 .64 .83 .91 1.05 Y 91 58.3 d .12 .18 .19 .21 .17 .21 .25 Mo99 68 h 6.0 5.9 5.9 5.9 5.8 6.0 6.0 Xe131m 12 d 2.I 2.5 2.5 2.4 2.2 2.5 2.7 Xe 133m 2.3 d 3.1 3.1 3.1 3.1 3.1 3.1 3.1 Xe 133 5.27 d 270. 270. 270. 270. 259. 280. 280. Xe 135m 15.6 m 1.05 1.05 1.05 1.05 1.05 1.05 1.05 Xe 135 9.2 h 6.7 6.7 6.7 ' 6.7 6.7 6.7 6.7 Xe 138 17 m .57 .57 .57 .57 .57 .57 .57 1 131 8d 3.6 3.6 3.6 3.6 3.6 3.6 3.6 I132 2.4 h 5.4 5.4 5.4 5.4 5.4 5.4 5.4 1133 20.8 h 4.2 4.2 4.2 4.2 4.2 4.2 4.2 1134 52.5 m .56 .56 .56 .56 .56 .56 .56 I 135 6.68 h 2.2 2.2 2.2 2.2 2.2 2.2 2.2 Cs 136 12.9 d .72 .82 .82 80 .73 .S5 .90 Cs 137 27 y 16. 29. 37. 40. 35. 41. 53. Cs 138 32.9 m .82 .82 .82 .82 .82 .82 .82 Ba 137 m 2.6 m 15. 27. 35. 37. 31. 38. 47. Ba 139 85 m .091 .091 .091 .091 .091 .091 .091 Ba 140 12.8 d .070 .073 .073 .073 .073 .073 .073 La 140 40.5 h .024 .024 .074 .024 .024 .024 .024 - Ce 144 290 d .0029 .0029 .0030 .0030 .0031 .0031 .0031 ( ll-T Amendment No. 2
D-B Table 11 h Tritium Activity in Reactor Coolant with Bleed Recycled Initial Cycle (433d) Time Ternary Fission Boron Activation 2 ppm Li Total (days) ue/ml ue/ml ue/ml ue/ml 1 .009 .0009 .0002 .01 10 .09 .009 .002 .10 50 .k2 .oh .02 .h8 8 100 .8h .08 .03 95 200 1.65 .1h 1.8h
.05 300 2.h6 .18 .07 2.71 h00 3.2h .20 .09 3.53 h33 3.51 .21 .10 3.82 Equilibrium Cycle (292d)
End of Ternary Fission Boron Activation 2 ppm Li Total Cvele # ue/ml ue/ml ue/ml ue/ml 1 3.51 .21 .10 3.82 2 5.h 30 .16 5.86 5 99 53 .30 10 10.73 8 15.0 79 .k5 16.2h 20 21.3 1.1 .63' 23.03 30 24.6 1.3 .T3 26.63 ho 26.7 1.h 79 28.89 f 1 l 1 0328
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D-B
. 11.1.1.k Methods of Disposal Four methods are defined in the treatment of the radioactife vastes.
- a. The clean liquid vastes consist of liquids such as reactor coolant T ii fue'l~ pool' coolant, which are relatively lov in
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chemical impurities and suspended solids content. Processing consists of degassing, storin6, filtering, demineral-izing and evaporating. The end products, concentrated boric acid and demineralized water, are normally stored for later reuse in the reacter cycle.
- b. The dirty liquid vastes consist of liquids of largely varying types and origins such as radioactive laboratory drains, bcilding sumps, and decontamination drains. These liquids are relatively high in chemical impurities and suspended 1 solida content but low in radioactivity. Normal processing consists of storage and filtration, and, if necessary, evaporation. The end products are nor= ally discharged from
,the plant, but the evaporator distillate may be reused.
- c. The gaseous vastes consist of the discharges from all poten-tially radioactive systems. Processing consists of compression into decay tanks, retention for a period of 30-60 days, release 1 through high efficiency filters, and discharge to the atmos-phere through the station vent.
- d. The solid wastes consist of all potentially radioactive solids vastes such as de=ineralizer resins , spe o filter elements, clothing and rags. Processing consic .f storage and packaging, as appropriate, for later .if-site disposal.
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s. D-B 11.1.2 SYSTEM LESCRIPTION AND OPERATION
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11.1.2.1 Liquid Waste System The flow diagram of the Clean Liquid Radioactive Waste Syst[m is shown in Figure 11-1. The major source of clean liquid vaste is the reactor coolant letdown that occurs during plant startups and dilution operations. Minor sources include leakoff, drain, and relief flows from valves and equipment inside the containment which contain reactor coolant. These smaller quantities are accu =ulated in the reactor coolant drain tank before processing. Hydrogenated liquid vastes pu= ped from the reactor coolant drain tank and re-leased from the makeup and purification system are sprayed into the 1 degasifier. The dissolved hydrogen and fission gases flash out of solution and are sent to the Waste Gas Disposal System.' The degasifier pump operation is controlled automatically by a signal from a level controller. Pressure and level instrumentation with alarus are provided on the deganifier to inform the operator of any calfunction. . The liqu'd is puzged from the degasifier through a filter and a boron saturated 3 lmixedbediunexdangedemineralizertooneofthetwocleanvastereceiver tanks. A nitrogen blanket in the tanks is automatically maintained above atmospheric pressure to prevent air in-leakage. This cover gas is released to a vaste gas collection header or into the other receiver tank when it is displaced. No flashing occurs in these tanks, and any transfer f hydrogen or fission gases from the liquid to the cover gas is by the slev 1 process of molecular diffusion. The vastes are then fed to an evaporator . where they are separated into their two reusable constituents: demineralized : water and concentrated boric acid. The distillate is passed through a 2 l polishing demineralizer into one of two clean vaste monitor tanks. From here it is reused, recycled, or discharged. The concentrated boric acid is sent to a concentrate storage tank from which it is taken to boric acid storage for reuse, or routed to the miscellaneous liquid vaste system for further concentration prior to disposal. At several points in this cycle, alternate flow paths are provided. These allow for the recirculation through, or the bypassing of, a demineralizer or evaporator. This flexibility, along with sampling at various sta6es in the cycle, permits the operator several options in insuring the adequate process-ing of the vaste. Near the end of core life, the coolant, instead of being processed by an evaporator is diverted through a deborcting demineralizer. Normally the flow is then dire .ted back to the make-up tank but the option exists to pass it through part, or all, of the Clean Liquid Radvaste System. A flow diagram of the Miscellaneous Liquid Radioactiva Waste System is shown < in Figure 11.2. The major sources for this system are showers , laundry, area washdowns, boric acid being disposef of , and any regeneration vastes. These, 1 depending on their composition, are e- .llected in either the miscellaneous or detergent vaste drain tanks. The contents of these are then monitored and' may be released' through filters directly to the discharge water . - 8 system. If further processing is required, the vastes are fed to an evaporator t x - 0'330 Amendment No. 8 11-10
D-3 and concentrated to 25% solids by weight. Neutralisation of the evaporator
,3 feed may be necessary in order to obtain this high concentration level. The distillate" Ices to the misjell e - - vaste monitor tank where a final check 1 on activity is made before storage for reuse or release to the discharge water system. The evaporator bottoms are tr - fe med to a concentrate '
tank where they are stored prior to dru= ming for off-sitTdrs)6's117- [I Before any liquid is released from a vaste tank to the discharge water system, a sa=ple is taken and its level of activity is determined. If it l8 does not meet established limits, it is recycled until it does. A final check is made on the vaste as it is discharged through c.o in-line radiation monitor. If its activity e .:eeds preset values, an alarm is annunciated and isolation valves automatically shut off the discharge. Some mode of re-cycling is then necessitated. A record vill be kept of both the amount and level of activity of all discharged effluent. The valves in the inlet line to a vaste tank are closed whenever the tank is being discharged so that filling and discharging to the discharge l8~ vater system cannot be done simultaneously. This insures batch processing. 11.1.2.2 Gaseous Waste System The Gaseous Waste System processes potentially radioactive hydrogenated and aerated vaste gases. A diagram of the system is shown in Figure 11-3 Sources of hydrogenated, radioactive vaste gas include the reactor coolant 1 drain tank, degasifier, the make-up tank and the quench tank. Hydrogen and fission gases stripped in the degasifier and vented from other tanks flov to the vaste gas header and then to the vaste gas surge tank. A nitrogen blanket in this tank tutomatically maintains a slight positive l5 pressure in the system. The gases are then compressed into one of three 1 decay tanks, which are sised for 30-60 days holdup. A sa=ple is removed from the tank and its activity level determined. If it is sufficiently lov, the gases are discharged to the vent through a HEPA filter. If it is high, the gases are allowed to decay until future sampling shows that they are suitable for controlled release to the atmosphere. The nitrogen cover gas displaced from the receiver tanks is also handled by this system. Since it should have little activity, an effort is made to keep it separate from the hydrogen and fission gases. This is done bv for.~ing one of the compressors and tanks into a separate processiv; chain. 1 The cover gas that has been compressed can be reused or vented srter sampling. Should any significant .contartination occur, then '.c must be l5 processed similarly to the normal hydrogenated vaste. If orly one compressor 1 is available at any time, all gases will be collected at a common header. . All released gases must pass a radiation monitoring system and if their activity exceeds a set point, an alarm is annunciated and the isolation valves in the discharge line ill automatically close.
,. Aerated radioactive vaste gases from the =iscellaneous and detergent vaste
'( - drain tanks will be processed separately from the hydrogenated vaste gases 5 to prevent the possibility of explosive mixtures. These lov level gases are
. Q,3 simply collectedand released to the vent. - - ~
g 11-11 Amendment No. 8
D-B 11.1.2.3 Solid Waste System Solid vastes are placed in ICC-approved containers appropriate for the vaste material. . Leaded containers are monitored for radiation levels and stored in a special area prior to shipment to an off-site disposal facility. Radio-active p ent "esins sluiced from demineralizers are collected and stored in the rent resin tank until a quantity sufficient for disposal is accumulated. The tank is sized for at least a one-year accumulation and is arranged with pumpout connections for resin transfer to a shipping cask for disposal. All 5l soft solid vastes, such as contaminated clothing, rags, viping towels, paper, gloves, and shoe coverings will be compressed into the containers by a baler. Hard solids such as wood, metal, glass, plastics, concrete and ceramics vill be put into the containers without compressing. Items too large for the containers vill be stored, packaged and disposed of as appropriate. 11.1.3 DESIGN EVALUATION . The possibility of an accidental release of activity from the radvaste system , is minimized by reuse of much of the liquid vastes. Liquid and stored gaseous vastes are sampled prior to discharge to the environment. Solid vastes are disposed of by licensed contractors in accordance with ICC regulations. All liquid radioactive vastes flow to storage tanks prior to discharge to the environment and cannot be discharged to the environment by gravity (i.e., the effluent must be pumped out). All actuator operated valves which control the discharge of radioactive material into the environment fail in the closed T position on loss of actuating force or signal. ) Radioactive geses are continuously monitored durin6 discharge in compliance with reqvtrements of 10 CFR 20. Standby units (pumps, ion exchangers, and compressors) permit continuous processing in the event of equipment failures or lautine maintenance. 11.1.k TESTS AND INSPECTIONS Functional operational tests and inspections of the radioactive vaste system vill be made as required to insure performance consistent with the require-ments of 10 CFR 20. Routine surveillance vill be conducted for detection of system leaks. Radiation detectors and monitors vill be periodically checked for calibration. Alarm circuits and automatic features of flow diversion for vaste liquid and gaseous effluents will be periodically tested. Each component is inspected and cleaned prior to installation into the system. Demineralized water is used to flush all portions of the system. Pre- . operational tests include calibration of instruments, testing of automatic controls, and verification of alarm set points. All flow paths are checked for capacity and mechanical operability. All pumps are run to demonstrate head and capacity. i Amendment No. 5
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11-12 On!:o 4
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/ ) 11.1 5 CODES The water processing system components are designed and fabricated in accor-dance with the following codes and standards:
- a. Tanks (including demineralizers) _
Tanks conform to Section VIII of the ASME Boiler and Pressure Vessel Code. All vetted surfaces are fabricated of a corrosion resistant material,
- b. Pumps Pumps conform to the standards of the Hydraulic Institute ,
and all vetted surfaces are fabricated of a corrosion . resistant material. Pump motors conform to standards of NEMA, IEEE, and USASI.
- c. Piping and Valves Piping and valves conform to code requirements for pressure piping. All pipes carrying radioactive vaste are
' USASI B317, Class III piping. Piping systems, where required, are fabricated of corrosion resistant material.
11.2 RADIATION PROTECTION 11.2.1 RADIATION ZONING AND ACCESS CONTROL The following list identifies the different zones used for the Davis-B' esse Nuclear Power Station. Design Dose Rate (mrem /h on a Designation h0 h/veek basis) Description A f, 0 5 Uncontrolled, unlimited access B 3, 2. 5 Controlled, unlimited access. h0 h/veek C 5, 15 Controlled, limited access for routine
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tasks 1 D- < 100 Controlled, limited access for short periods 2 E Controlled occupancy for very short
> 100 f
()"3'}(( periods. Occupancy during emergencies. l1 A. __ Normally inaccessible. l2
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b?'- Amendment No. 2 11-13 . _ . . _ _ _
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'JNCONTROLLED areas are those that can be occupied by plant personnel or
- visitors on an unlimited time basis with a minimum probability of health hazard from radiation exposure. ~
1 CONTROLLED areas are those where higher radiation levels and/or radioactive ! conta=ination which have a greater probability of radiation health hazard to individuals can be expected. These areas can be entered only by individuals who have passed through the plant access cont ol station. Normally, only individuals directly involved in the operacion of the plant will be allowed i to enter these areas. LIMITED ACCESS areas are those that have radiation levels of less than 100
. mrem /h and which can be entered either through open passa6es or unlocked doors. These areas are identified by radiation caution signs at strategic locations.
INACCESSIBLE areas are those where dose rates above 100 mrem /h can be expected. These areas are either blocked off completely or can be entered only through . locked doors. Access is supervised from the access control station and the
. station control room.
In case of emergency, personnel vill be able to use escape routes which involve the minimum exit time. 11.2.2 RADIATION SHIELDLNG 11.2.2.1 Design Bases 3
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The basis for the shielding design for nor=al plant operation is the " Code of Federal Regulations," Title 10, Chapter 1, Part 20, entitled " Standards for Protection Against Radiation." All areas of the plant are subject to these regulations. The areas are zoned according to their expected occupancy by plant personnel and radiation ex-posure levels under normal operating conditions. The maximum whole body exposure for station personnel is 1.25 rem per calendar quarter. For the general public, the mi- doce is not more than 0.5 rem for one calendar year. No individual vill receive more than 25 rem of whole body exposure during the course of any accident, in accordance with 10 CFR 100. The control room is designed to limit whole body exposure to 5 rem durfng the course of the MHA. This provides an allowance for excursions into other areas of the station to i attend to critical equipment. l ! , 11.2.2.1.1 Radiation Exposure of Materials and Components . l No regulations similar to those established for the protection of individuals exist for raterials and components. Materials are selected on the basis that radiation exposure vill not cause significant changes in their physical pro-perties which adversely affect their operation during the design life of the plant. Materials for equipment required to operate under accident conditions are selected on the. basis of the additional exposure received. 11-lh ,
D-B
; 11.1 5 CODES The water processing system components are designed and fabricated in accor-dance with the following codes and standards:
- a. Tanks (including demineralizers)
Tanks conform to Section VIII of the ASME Boiler and Pressure Vessel Code. All vetted surfaces are fabricated of a corrosion resistant material.
- b. Pumns Pu=ps conform to the standards of the Hydraulic Institute and all vetted surfaces are f abricated of a corrosion resistant =aterial.
Pump motors conform to standards of NEMA, IEEE, and ANSI. l3
- c. Piping and Valves Piping and valvra conform to code requirements for pressure piping. All pipes carrying radioactive vaste are ANS B31.T, Class III piping. Piping systems, where 3 require , are-fabricated of corrosion resistant material.
11.2 RADIATION PROTECTION 11.2.1 RADIATION ZONING AND ACCESS CONTROL The following list identifies the different zones used for the Day's-B' esse Nuclear Power Station. Design Dose Rate (mrem /h on a Designation h0 h/veek basis) Description A <05 Uncontrolled, unlimited access B < 2.5
, Controlled, unlimited access. ho h/veek C < 15 Controlled, limited access for routine tasks .
D < 100 Controlled, limited access for short periods 2 E > 100 Controlled occupancy for very short
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periods. Occupancy during cmergencies. l1
\ Nor= ally inaccessible. l2
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. 11-13 0335 Amendment No. 3
D-B UNCONTROLLZD areas are those that can be occupied by plant personnel or visitors on an unlimited time basis with a =inimum probability of health hazard from radiation exposure. CONTROLLED areas are those where higher radiation levels and/or radioactive contamination which have a greater probability of radiation health hatard to individuals can be expected. These areas can be entered only by individuals who have passed through the plant access control station. Normally, only individuals directly involved in the operation of the plant will be allowed to enter these areas. LIMITED ACCESS areas are those that have radiation levels of less than 100 mrem /h and which can be entered either through open passages or unlocked doors. These areas are identified by radiation caution signs at strategic locations. INACCESSIBLE areas are those where dose rates above 100 mrem /h can be expected. These areas are either blocked off completely or can be entered only through . locked doors. Access is supervised from the access control station and the
. station control room.
In case of emergency, personnel vill be able to use escape ro'utes which involve the minimum exit time. 11.2.2 RADIATION SHIELDING 11.2.2.1 Design Bases The basis for the shielding design for normal plant operation is the " Code of Federal Regulations," Title 10, Chapter 1, Part 20, entitled " Standards for Protection Against Radiation." All areas of the plant are subject to these regulations. The areas are zoned according to their expected occupancy by plant personnel and radir. tion ex-posure levels under normal operating conditicas. The maximum whole body exposure for station personnel is 1.25 rem per calendar quarter. For the general public, the maximum dose is not more than 0 5 rem for one calendar year. No individual vill receive more than 25 rem of whole body exposure during the course of any accident, in accordance with 10 CFR 100. The control room is designed to limit whole body exposure to 5 rem during the course of the MHA. l This provides s. allowance for excursions into other areas of the station to attend to cS;tical equipment. 11.2.2.1.1 Radiation Exposure of Materials and Components 1 No regulations similar to those established for the protection of individuals l exist for materiala and components. Materials are selected on the basis that radiation exposure vill not cause significant changes in their physical pro-perties which advermely affect their cceration during the design life of the plant. Materials for equipment required to operate under accident conditionse are selected on the basis of the additional exposure received. u-14 O
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0342 DAVIS-BESSE NUCLEAR POWER STATION WASTE GAS SYSTEM 1 _ ^- FIGURE 11-3 I AMENDMENT NO. 3
PROCESS MONITORING ANNUNCIATOR RECORDER C
, ONTROL WATERBORNE MONITORS CLEAN WASTE DISCHARGE - s INDICATOR $ $ $
MONITOR ' WITH ALARM MISCELLAN. .)US WASTE , s INDICATOR $ $ $ DISCHARGE MONITOR WITH ALARM REACTOR COOLANT SYSTEM w INDICATOR h h ACTIVITY WITH ALARM CCW SYSTEM MONITOR ' INDICATOR h WITH ALARM AIRBORNE MONITORS GASEOUS WASTE DISCHARGE s INDICATOR h & $ MONITOR WITH ALARM CONTAINMENT 4TMOSPHERE s INDICATOR h h MONITOR (PARTICULATE & GAS) WITH ALARM STACK MONITOR s INDICATOR h h h (PARTICULATE & GAS) WITH ALARM CONDENSER AIR EJECTOR ' v INDICATOR f VENT LINE MONITOR WITH ALARM PENETRATION BUILDING A INDICATOR h k VENTILATION MONITOR WTTH ALARM FUEL HANDLING AREA N INDICATOR h h VENTILATION MONITOR WITH ALARM ) RADIOACIIVEWASTE AREA s INDICATOR h VENTILATION MONITOR %1TH ALARM AREA MONITORING CONTROL ROOM AREA MONITOR s INDICATOR
%TTH ALARM CONTAINMENT FUEL HANDLING s INDICATOR h h AREA MONITOR WITH ALARM PERSONNEL LOOC AREA s IND'rATOR h h MONITOR %TiH ALARM IMORE SENSING EQP. s INDICATOR h h AREA MONITOR WITH AIARM AUXILIARY BUILDING FUEL HANDLING BRIDGE AREA MONITOR s
INDICATOR WITH ALARM
+ 0343 H SAMPLE SINK AREA s INDICATOR h h MONITOR WITH ALARM -.
CASK DECONTAMINNTION AND s INDICATOR h h LOADING AREA MONITOR WITH ALARM RADIOACTIVE WASTE SYSTEM w INDICATOR h AREA MONITOR WITH ALARM DAVIS-BESSE NUCLEAR POWER STATION PROCESS MUNITORING FIGURE 11-4 l m
D-B 11.2.2.1.2 General Design Considerations t The shielding design considers three conditions:
- 1. Full core power operation at 2772 MWt. This also includes shielding requirements for certain off-normal conditions such as the release of fission products from leaking fuel elements.
- 2. Shutdown. This condition deals mainly with the radioactivity from the suberitical reactor core, with radiation from spent fuel bundles during on-site transfer, and with the residual activity in the reactor coolant system and neutron-activated materials.
- 3. A hypothetical accident in which 100% of the noble gases, 50% of the halogens and 1% of the other fission products are released from the reactor core (TID 1h8hh}.
11.2.2.1.3 Specific Design Values The material used for most of the station shield is ordinary concrete and concrete block with a bulk density of about 1h3 lb/ft3 Only in a very few instances vill steel or water be utilized as primary shielding materials. 11.2.2.2 General Descriptions and Evaluations 11.2.2.2.1. Shield Building The shield building serves two main shielding purposes:
- 1. During operation, it shields the surrounding station structures and yard areas from radiation originating at the reactor vessel and the primary locp components. Together with addi-tional shielding in the interior and in the valls of the containment vessel, the concrete shell will reduce radiation levels outside the shell to below 0 5 mrem /h in uncontrolled areas.
- 2. In the event of an accident, the shielding vill reduce station and off-site radiation intensities, emitted directly from released fission products, to acceptable emergency levels.
The concrete roof of the shield building vill effectively reduce contributions due to sky shine. 11.2.2.2.2 Containment Vessel Interior During operation, most areas inside the' containment vessel are inaccessible
, ,because of dose rates greater than 100 mrem /h and contamination of the
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- atmosphere by activation and/or fission products. The reactor vessel, which is the major radiation. source, is surrounded by a heavy concrete biological' shield. A concrete shield also surrounds equipment that carries reactor ecolant water.
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D-B Inside the contain=ent vessel, shielding is provided around the reactor in-ternals storage pool. This shielding is designed for personnel protection S during storage of activated reactor internals and for protection during / transfer of spent fuel elements to the transfer. tube, 11.2.2.2.3 Auxiliary Building All radioactive areas can be reached through service corridors which vill be entered from the access co trol station. For normal equipment operation, none of the high radiation area need to be occupied since all manually operated valves of contaminated equipment vill have reach rods which penetrate through the shield valls into the corridor or vill have remote manual operators. Gages and other instruments which need visual checking from time to time can be inspected from the corridors or on the local or central control boards. The different systems are isolated from each other by individually shielded chambers. Syste=s can be iaolated for maintenance or repair with no signifi-cant radiation interference from other systems. Heavy concrete shielding is provided around the vaste gas decay tanks wherever they are adjacent to access areas. The :ounting room has been shielded in order to reduce background radiation as much as possible. For the room contr.ining the ventilation system, concrete block is used to shield against radiat. ton from the ventilation filters. TL. relative closeness of fully aceassible and uncontrolled areas requir s especially h'eavy shielding for the area surrounding the spent fuel pool. The control room has concrete shielding for those sides which are in direct )
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line of sight with the shield building. The integrated whole body ga==a dose inside the control room vill be less than 5 rem over a period of 30 days following an MEA. ll.2.2.2.h Turbine Building The turbine building is fully accessible and uncontrolled with dose rates much less than 0.5 mrem /h during normal plant operation as well as during shutdown. In the event of an MEA, access to the turbine building is controlled. 11.2.2.2 5 General Plant Yard Areas The radiation shielding design of the shield building and auxiliary buildings protects all plant yard areas from excessive radiation exposure. All yard areas which are frequently occupied by plant personnel receive a radiation field of less than 0.5 mrem /h. 0345
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D-B [ 11.2.3 RADIATION MONITORING SYSTEM 11.2.3.1 Design Basis The radiation monitoring system is designed to:
- 1. Continuously detect and record the level of radiation in the plant effluents released to the environment.
- 2. Provide operating personnel with a continuous indication and records of the gamma radiation levels in selected plant areas.
3 Protect operating personnel from exposure to excessive ra'diation levels or radioactive concentrations by alarm annunciation and in so=e cases automatic action of protective equipment in the event that such limits are exceeded.
- To fulfill these design criteria the radiation monitoring system consists of interrelated subsystems as described in Figure 11.h. These are identified as the Airborne Radiation Monitoring System, Waterborne Radiation Monitoring System and Area Radiation Monitoring System.
11.2.3.2 Airborne Radiation Monitoring System 11.2.3.2.1 Stack Gas Monitors A continuous sample is drawn from the stack via an isokinetic sa=ple probe and
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activity is monitored by a beta-gz=na air particulate monitor consisting of a detector and removable filter paper asse=bly and a gas ga--a monitor. Tne sample is then returned to be discharged through the stack. Radioactivity is indicated and recorded in the main control room. An alarm is initisted upon the detection of activity above the specified limit and closure of the con-tainment ventilation valves is initiated. 11.2.3.2.2 containment vessel Monitors A continuous air sample is drawn from the containment vessel by an air particulate monitor consisting of a beta-ga=na detector, removable filter paper assembly ^ and a gas gm=na monitor. Radioactivity is indicated and a recorded in the main control room and an alarm and closure of the containment purge valves is initiated upon the detection of activity above the specified limit. 11.2.3.2.3 Waste Gas Monitor This system consists of an off-line -onitor which uses a gacma detector. The systeb,jputinually indicates and records radioactivity in the gas aad initiates
-ly; an alarm ~1n the control room if the specified activity level is exceedei . The system' automatically closes the vaste discharge header valve if an alarm r- situation does occur.
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D-B 11.2.3.2.h Condenser Air Ejector Monitor The air ejector vents the non-condensable gases from the condenser to the j stack. An off-gas monitoring system continuously monitors for the presence of radioactivity in the non-condensable gases which would indicate a primary to secondary leak in the steam generators. The sampler vill be located in the air ejector vent line and consists of an off-line beta-gamma detector which vill indicate and record in the main control room.- An alarm vill be initiated in the event tnat the radioactivity in the non-condensable gases exceeds a preset limit. 11.2.3.2.5 Ventilation Syste=s Monitors The fuel handling, radvaste and penetration areas are ventilated. Each is equipped with a ventilation radiation monitoring system which records the activity levels in the vent ducts from each area and actuates an alarm in the control room when the act'ivity levels reach a preset level. 11.2.3.3 Waterborne Radiation Monitoring Systems 11.2.3.3.1 Radvaste Liquid Monitoring System The radvaste liquid monitor initiates an alarm and terminates the release of radvaste effluent when the activity level in the liquid effluent exceed a preset limit. An in-line monitor is placed on the vaste discharge line. In the case of abnormally high radiation, the vaste discharge valve vill be closed.
. 11.2.3.3.2 Component Cooling Water System Monitors This system has two beta-ga=ma monitors in the inlet lines to the CCWS pumps.
Should the radioactivity level in the system rise above a preset limit, the atmospheric vent valve of the head tank is automatically closed. The system vill then operate unvented with relief to the radioactive vaste system for overpressure protection. 11.2.3.3.3 Reactor Coolan+, System Activity Monitor This monitor vill detect gross increases in the activity of the reactor
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coolant. It vill measure activity in the letdown flov from the reactor coolant system into the makeup and purification system. 11.2.3.h Area Radiation Monitoring System This system consists of beta-ga=ma detectors placed at appropriate locations in the following areas:
- a. One detector near the fuel handling bridge inside the containment ,
- b. Inside containment near the personnel access hatch.
- c. Inside containment near in-core monitoring equipment. gg
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- d. Near fuel handling bridge in auxiliary building.
- e. Auxiliary building near sample sink.
- f. Auxiliary building cask decontamination and loading area.
- g. Control room.
- h. Radioactive vaste system erea.
- 1. Selected areas to be determined later, i.e. , passageway, etc.
Readout for each detector will be provided in the centrol room. High radiation alarm signals for each detector will be furnished to the control rocm and to each detector location. Sources will be available to allow the overall system performance to be verified at regular intervals. Detector ranges will be determined depending upon the normal background at the detector locations and the expected radiation levels for abnormal conditions. ~ The multichannel area radiation monitoring system monitors the radiation intensity of areas in the plant where it is possible for operating personnel to be sutject to abnor= ally high gamma radiation. The selection and number of points are coordinated with the plant access controA so that operating personnel are not able to enter an unmonitored area in which they could be exposed to an excessive dose. (
- I fif ; f; O 0348 11-19
D-B
11.3 REFERENCES
(1) Frank, P. W., et al, Radiochemistry of Third PWR Fuel Material Test - X-1 Loop NPX Reactor, WAPD-TM-29, February, 1957
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(2) Eichenberg, J. D., et al, Effects of' Irradiation on Bulk UO , WAPD-183, 2 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 UO2 Fuel Elements at Chalk River, AECL-1338, 1961. (h) Illison, G. M. and Roe, H. K., The Release of Fission Gases & Iodines From Defected UO2 Fue: Elements of Different Lengths, AECL-2206, June, 1965
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