ML19329D993

Chapter 11 of Rancho Seco PSAR, Radwastes & Radiation Protection. Includes Revisions 1-4

ML19329D993
Person / Time
Site:	Rancho Seco
Issue date:	10/31/1967
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
References
NUDOCS 8004090517
Download: ML19329D993 (31)

Text

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/ 's TABLE OF CONTENTS ( ) ~' 11. RADIOACTIVE WASTES AND RADIATION PROTECTION Section Pace 11.1 RADIDACTTVE WASTE HANDLING 11.1-1 11.1.1 DESIGN BASES 11.1-1 11.1.1.1 Performance obiectives 11.1-1 11.1.1.2 Radioactive Waste Quantities 11.1-1 11.1.1.3 Waste Activity 11.1-1 11.1.1.4 Disposal Methods 11.1-2 11.1.1.5 Shieldine 11.1-6 11.1.2 SYSTEM DESIGN AND EVALUATION 11.1-6 11.1.2.1 Liquid Waste Disposal System 11.1-6 11.1.2.2 Solids Waste Disposal System 11.1-7 11.1.2.3 Gaseous Waste Disposal Svstem 11.1-7 11.1.2.4 Process System Radiation Monitorine 11.1-8 11.1.2.5 Desien Evaluation 11.1-8 11.1.3 TESTS AND INSPECTIONS 11.1-8 11.1.4 TRIT'a'M MANAGEMENT FOR NORMAL OPERATION 11.1-8 11.1.4.1 Tritium Production 11.1-8 11.1.4.2 Tritium Concentration at the S ite Boundary 11.1-11 11.1.4.3 Tritium in Containment Air Durine Refueline 11.1-12 11.1.4.4 Tritium Release From Sprav Pond 11.1-13 /)N 11.1.4.5 Environmental Consecuences of Tritium Leakage I into Secondarv Coolant 11.1-14 11.1.4.6 Tritium Release from Cooling Tower Canal 11.1-14 11.1.4.7 Re ferences 11,1-14 11.2 RADIATION SHIELDING 11.2-1 11.2.1 PRI>MRY, SECONDARY, REACTOR BUILDING, AND AUNILIARY 11.2-1 SHIELDING 11.2.1.1 Desien Criteria 11.2-1 11.2.1.2 Description of Shieldine 11.2-2 11.2.1.2.1 Primary Shield 11.2-2 11.2.1.2.2 Secondary Shield 11.2-2 11.2.1.2.3 Reactor Building Shield 11.2-3 11.2.1.2.4 Control Room Shield 11.2-3 11.2.1.2.5 Auxiliary Shield 11.2-3 4 11.2.1.2.6 Spent Fuel Shielding 11.2-4 11.2.1.2.7 Materials and Structural Requirements 11.2-4 11.2.1.3 Fvaluation 11.2-4 11.2.1.3.1 Radiation Sources 11.2-4 11.2.1.3.2 Neutron and Gamma Shield 11.2-5 11.2.1.3.3 MHA Dose calctilation 11.2-5 11.2.1.3.4 operation Limits 11.2-6 11.2.1.3.5 Radiation Surveys 11.2-6 11.2.2 AREA RADIATION 50 nil 0 RING SYSTEM 11.2-6 11.2.2.1 Desten Bases 11.2-6 11.2.2.2 nesertnti.,n 11.2-6 11.2.2.3 rvalu,eion 11.2-7 000 0105 Amendment 4 11-i-

11.2.3 llEALTil PilYSICS 11.2-8 11.2.3.1 Fadiation Work Permits 11.2-8 11.2.3.2 Personnel Monitoring System 11.2-9 11.2.3.3 Personnel Protective Equinment 11.2-10 11.2.3.4 Chance Room Facilities 11.2-10 11.2.3.5 liealth Physics Facilities 11.2-10 11.2.3.6 liealth Physics Instrumentation 11.2-11 11.2.3.7 Medical Programs 11.2-11

11.3 REFERENCES

11.3-1 LIST OF TABLES Table Number Title Pace 11.1-1 Radioactive Waste Quantities 11.1-3 11.1-2 Escape Rate Coefficients for Fission Product 11.1-4 Release 11.1-3 Reactor Coolant Activity for the 3rd Core Cycle 11.1-5 Based on 17. Fuel Elements Failed O LIST OF FIGURES Finure Number Title 11.1-1 Waste Gas and Miscellaneous Liquids Radwaste System 11.1-2 Coolant Radwaste System 11.1-3 Tritium Concentration at the Site Boundary During 40th Operational Cycle 11.1-4 Tritium Concentration in Containment Air Boundary 11.1-6 Tritium Concentration at the Site Boundary Resulting From Steam Generator Tube Leakage ('OJ 0106 9 11-i1 Amendment 4

y v ' ^ ~ ' 11. ' RADIOACTIVE WASTE'S AND RADIATION PROTECTION 11.I liADI0 ACTIVE WASTE HANDLING 11.1.1 . DESIGN BASES 11.1.1.1 Performance Objectives The waste disposal system will be designed to provide controlled handling and disposal of liquid, gaseous, and solid wastes which will be generated during plant operation. The design criteria are to ensure that pir.nt personnel and the general public are protected agains.t excessive exposure to radiation from wastes in accordance with limits defined fr 10 CFR 20. 11.1.1.2 Radioactive Waste Quantities The estimated volumes of radioactive wastes generated during plant opera-tion are listed in Table 11.1-1. 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 Icakage through clad defects in 1 percent of the fuel. The activity levels were computed assuming full power operation of 2,568 Mwt for one core cycle with no defective fuel followed by operation over the second core cycle with 1 percent defective fuel. Continuous reactor coolant purification at a rate of one reactor system volune.per day was used with a zero removal efficiency for Kr, Cs, Xe, Mo and Y, and a 99 percent removal efficiency for all other nuclides. All Te is assumed to plate out on the system sur-face. Activity levels are relatively insensitive to small changes in demineralizer efficiencies, e.g., use of 90 percent inrtead of 99 percent would result in only about a 10 percent increase in the coolant activity. The quantity of fission products released to the reactor coclant during steady state operation is based on the use of "esespe rate coefficients" (sec -1) as determined from experiments involvi.ng purposeiy defert.ed fuel elements.1,2,3,4 talues of the escape rate coef ficients used in che calcu-lations are shown in Table 11.1-2. Calculations of the activity released from the fuel were performed with a digital computer code which solves the dif ferential equations for a five-member radioactive chain for buildup in the fuel, release to the coolant, removal from the c6olant by purification and leakage, and collection on a resin or in a holdup tank. The activity levels _in the reactor coolant for a unit containing one percent defective fuel during full Tower operation at the end of the second core cycle are shown in Table 11.1-3. 000 010)i. 0. 11.1-1

) c-: s ~ Radioactive Waste' Handling ^ 4 The liquid _wasteLgenerated by leakage, sampilng; and demineralizer sluice f.l~ or rinse is assumed to have an activity concentration equal to the concen- ~ tration in'the reactor coolant. Reacter coolant bleed will be taken from j -the downstream side of'the purification demineralizer. It is assumed to ihave the same activity concentration as the reactor coolant reduced.by the decontamination factor of the purification demineralizer. Shower wastes

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- are assumed to contain negligible a' mounts of radioactivity. - 'rc 2:n r Gaseous activity will;be generated by the evolution of radioactive gases <$.31 ? from liquids stored'in tanks throughout the plant. These include such items as coolant waste tanks, and the makeup tank which are vented to the waste gas disposal system. The activity of the gases is dependent upon 3 the liquid activity.. 11.1.1.4 -Disposal Methods Liquid vastes from the plant will be handled in two separate streams using two evaporator chains. Reactor coolant bleed will be fed through one chain; and miscellaneous wastes which may contain dirt, oil, or chemicals will'be processed through the other chain. The treatment of the wastes will be in one of the following ways: a. Reactor coolant wastes will be de-gassed by flashing, ion-exchanged, stored, and separated into concentrated boric, acid and coolant make-up, both of which will be re-used. I b. Miscellaneous liquid wastes will be collected and evaporated, with the condensate re-used and the bottoms drummed for off-site disposal by an AEC-licensed disposal contractor. TABLE 11.1-1 l RADIOACTIVE WASTE QUANTITIES "Ja.ste Source Quantity Per Year Assumptions and Comments Liquid Waste: Reactor Coolant System: Scartup Expansion 148,000 gal 4 cold startups Startup Dilution 88,000 gal 2 cold startups at beginning of life and 1 cold startup at 100 and 200 full power days respectively 1 i ~ 000-0108 11.1-2 0 -e ~ m 1 u-x -yvv--1---e'e+w - + +- me e +e +ru--= 7^t--i t7?3----wt - Tre TW+- -Mr yw' p y*

~ Rrdio2ctive Wacte llandling TABLE 11.1-1 continued Waste Source Quantity Per Year Assumptions and Comments Lifetime Shim Bleed 176,000 gal Dilution from 1,460 to 175 ppm System Drain 45,000 gal, Drain to level of outlet nozzles for refueling operations Sampling and Laboratory 22,500 gal 12 samples per week at Drains 5 gal per sample Purification Demineralizer 160 ft3 3 80 ft / car replacement Sluice-at 2 ft /ft3 resin sluice 3 Spent fuel Pool Demineral-42 ft3 21 ft / year replacement izer Sluice at 2 ft3/ft3 resin sluice Deborating Demineralizer 2,500 ft3 1 regeneration per year Regeneration and Rinse per demineralizer at 20 ft /ft3 resin regen-3 eration Miscellaneous System 45,000 gal 5 gph leakage Leakage Showers 110,0'00 gal 10 showers per day at 30 gal per shower Gaseous Waste *: Of f-Gas from Reactor Coolant 1,350 ft3 Degas at 25ccll2 Fer System liter concentration Off-Gas from Liquid Sampling 74 ft3 Degas at 25ccll2 rer liter concentration Of f-Gas from Makeur Tank 900 ft3 \\.ent once per year Of f-Cas from Pressurizer (0 ft3 Vent once per jear Solid Waste: Purification Resin 80 ft3 Resin replacement once per year

• Excludes reactor building and plant ventilation 000 0109 11.1-3 Q{

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Radioactive Waste Handling TABLE 11.1-1 continued Waste Source Quantity Per Year Assumptions and Comments Spent Fuel Pool lon-20 ft Resin replacement once' Exchanger Resin per year Evaporator condensate Ion-2 ft Resin replacement once Exchanger Resin per year 3 Evaporator Bottoms 800 ft Concentrated to 20 per-cent solids Radwaste Ion-3 Exchanger Resin 100 ft Four replacement bede per year TABLE 11.1-2 ESCAPE RATE COEFFICIENTS FOR FISSION PRODUCT RELEASE Element Escape Rate Coefficient, sec-1 ~ Xe 1.0 x 10,7 Kr 1.0 x 10-8 1 2.0 x 10-8 Br 2.0 x 10-2.0 x 10,8 Cs 2.0 x 10,g Rb 4.0 x 10,9 Mo 9 Te 4.0 x 10-10 Sr 2.0 x 10-10 Ba 2.0 x 10 1.0x10[ff Zr Ce and other rare earths 1.0 x 10 000 0110 'C ... O

4 Radioactive Ifaste llandling TABLE 11.1-3 REACTOR COOLANT ACTIVITIES FOR A UNIT CONTAINING ONE PERCENT DEFECTIVE FUEL Isotope Activity, pCi/mi Isotope Activity, pCi/mi ' Kr 85m 2.0 1 131 3.3 Kr 85 15.5 1 132 4.9 Kr 87~ 1.1 1 133 4.5 Kr 88 3.7 I 134 0.55 Rb 88 3.7 I 135 2.1 Sr 89 0.057 Cs 136 0.81 Sr 90 0.0028 Cr 137 77.0 Sr 91 0.057 Cs 138 0.74 Sr 92 0.018 Mo 99 5.5 Xe 131m 2.1 Ba 139 0.088 ~ Xe 133m 3.2 Ba 140 0.076 Xe 133 290.0 La 140 0.026 Xe 135m 1.0 Y 90 0.89 Xe 135 9.4 Y 91 0.29 Xe 138 0.5 Ce 144 0.0027 c. Gaseous wastes can be compressed and stored for decay if activity is high, with later release at controlled rates through filters to the environmental stack discharge,.or can be diverted to.the fi,1ters and stack without hold-up where activity is sufficiently low. d. Solid wastes will be accumulated and packaged in suitabic drums for later removal by an AEC-licensed disposal con-tractor. 11.1.1.5 Shielding Shiciding for the components of the waste disposal system will be designed on'the basis of system activity IcVels with nominal 1 percent failed fuel. All components will be located in ti.e auxiliary building. The shicid. design criteria for the auxiliary building is Zone II on controlled areas and Zone III in areas requiring limited access. The piping and equip-ment of"the waste disposal system will be shielded by concrete walls and floors of varying thicknesses depending on the strength of the radiation sources therein and on the access requirements in a particular area. In some areas local shiciding in the form of removable Icad or concrete blocks may be utilized to facilitate ma,intenance or operations. 000 0111 00

~ l - . v, + y I Radicactive Waste Handling 11.1.2' SYSTkM DESIGN AND EVALUATION l .11.1.2.1 Liquid Wa'ste Dis'posal System Liquid waste handling will be divided into two separate waste processing chains. One chain will process the reactor coolant bleed stream and reactor coolant drains, and the other will handle all miscellaneous liquid The conceptual system flow diagram is shown in Figure 11.1-1 and wastes. 11.1-2. Reactor coolant will be received from the makeup and purification system and will be the larg'est single source of operational liquid waste to be handled. This liquid ul11 be received as a result.of reactor coolant expansion and of operational requirements for reducing or increasing reac-tor coolant boric acid content. After flashing and ion-exchange, it will be stored in reactor coolant waste receiver tanks and holdup tanks for, concentrator feed cr. will-be passed through deborating ion-exchangers for coolant make-up. The deborating ion-exchangers will be used only for ~ boric acid concentrations below 1000 ppm in order to limi,t the frequency at which resins must be regenerated. The reactor coolant waste receiver and holdup tanks will be si ed to con-tain one reactor coolant system volume each. The contents of these tanks will be pumped to the concentrator. When the coolant biced has been suffi-cicotly concentrated, the bottoms will be pumped to the concentrated boric ~ acid storage tank. The condensates will be collected in the concentrator condensate tank where they will be sampled to determine quality and activity

• Icvel. Condensates will then be pumped through the deborating ion-exchanger to the demineralized water storage tank for re-use.

The second evaporator chain will process liquid wastes collected by the miscellaneous wastes tank and auxiliary building sump. The miscellaneous liquid wastes will be treated as necessary to prevent foaming and sampics will be taken to determine activity.' Wastes will then be transferred to the miscellaneous wastes evaporator. When wastes are sufficiently concentrated, the concentrates will be collected and subse-quently pumped to the waste drumming area for packaging and disposal by an AEC-licensed disposal contractor. Condensate will be collected in the evaporator condensate tank, sampled for activity, and subsequently re-used as demineralized water af ter passing through the evaporator condensate demineralizer. Both evaporator chains will be designed.to process wastes at a rate well in excess of the expected waste accumulation rate. As indicated above,.no liquid waste will be discharged to the environment. ~ 000 0112 a

m 3 je g ,, a s s ^ 21 .l ~ .' ~ 2 .g ~: Radioactive-Waste Handling 9 11.1.2.2^ Solids' Waste Disposal. System f Evaporator concentrate will be pumped into a shipping container for off-site - disposal.~ Spent' resins frem the demineralizers and ion-exchangers will be ~ sluiced to a spent resin storage tank which will hold one complete charge of resins from the reactor auxiliary systems. Spent resin will be transferred from the storage tank to special casks or. drums for disposal. Other misecl-laneous solid wastes such as filters, clothing, laboratory equipment, pieces of metal, and paper will be disposed of by the use of a baler and light metal shipping containers. All solid wastes will,be transported off-site in approved containers by an AEC-licensed disposal contractor. 11.1.2.3 Caseous Waste. Disposal System Gaseous radioactive wastes are collected mainly at the flash tank and at .the reactor coolant system drain tank during operation and at the purifi-cation system make-up tank during reactor shutdown degasification. Gas from each of these locations is piped to the waste gas surge tank, from where it can be either released directly through high efficiency particu-late and charcoal filters to the plant vent or can be compressed and ~ stored in the waste gas decay tanks. In the waste gas decay tanks, gases are monitored-for activity, held for decay as required, and then released at a controlled rate through the plant vent. A monitor located-in the gaseous discharge line will be equipped with an indicator alarm to annun- ~ ciate high activity. The high activity alarm actuates an interlock to stop the discharge of these gaseous effluents from the waste gas system. Cases, diffusing from liquids at other collection tanks will be swept by. continuous air purge streams into the gas discharge header and through filters'to the plant vent. Because of the inherently slow nature of the diffusion process through liquids, this gas release will not cause sig-nificant ground-1cyc1 concentrations. 11.1.2.4 Process System Radiation Monitoring The component cooling water system which removes heat.from potentially radioactive sources will be monitored to detect accidental releases. Mon-itors will be provided in the component cooling water system which serves the reactor coolant pump seal return coolers, spent fuel cooler, sample coolers, pressurizer relief tank cooling coils, decay heat removal coolers,. and letdown coolers. A high radiation alarm will alert the operator, and the leaking heat exchanger can be isolated. Reactor coolant letdown flow will be monitored to detect a gross fuel assembly failure. A smaller fuel assembly leak will be detected by regu-lar' laboratory analysis of reactor coolant sampics. JAir sampics from the reactor buildings and the plant vent will be monitored for air. particulate, gareous activity, and iodinc activity. 000 0113 L RRml-7

7 S s;: --%._u.._......_.,- k i Radioactive Waste Handling These radiation monitors are commercially-available equipment. The required characteristics will be established during detailed plant design. The minimum sensitivity of detectors, when combined with appropriate dilu-tion factors, will ensure safe limits of release. 11.1.2.5 Design Evaluation The possibility of a significant activity release off the site from either the solid or the liquid waste disposal equipment is.. extremely 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 ~ leakage. Solid wastes will be disposed of by being removed off-site by ap AEC-licensed disposal contractor. No liquids or solids will b.e discharged to the environment. Boric acid and coolant will be purified and re-used. 11.1.3 TESTS AND INSPECTIONS Functional operational tests and inspections of the waste disposal syster. will be made as required to ensure performance consistent with the require-ments of 10 CFR 20. 4 g OL 4 I

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xd 11.2 RADIATION SHIELDING 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY SHIELDING 11.2.1.1 Design criteria Plant operating personnel must be protected by radiation shielding wherever a potential radiation hazard may exist. The shielding must perform two primary functions: ensure that under normal operation the potential radiation dose to operating personnel and the general public is within the limits of 10 CFR 20, and ensure that operating personnel are adequately protected in the event of a power plant accident so that the accident can be terminated without undue hazard to the general public. All plant areas which can be occupied by personnel are classified accord-ing to anticipated access, as follows: Zone I - Normal continuous occupancy with radiation levels not exceeding 1 mrem /hr. Zone II - Periodic occupancy with radiation levels not exceeding 2.5 mrem /hr. Zone III - Limited occupancy with radiation levels not exceeding 15 mrem /hr., when occupied. .m. (~~ - Restricted occupancy consistent with measured Zone IV radiation levels exceeding 15 mrem /hr. The turbine structure, offices, turbine plant service areas, and the control room areas are designated Zone I. Areas such as the local control space in the auxiliary building and the waste disposal area, operating deck of the spent fuel storage building, and the operating deck of the reactor plant during shutdown would generally be designated Zone II. Inter-mittently occupied work areas are designated as Zone III. Typical Zone IV areas include, reactor loop areas after shutdown, drumming areas, and makeup and purification tank areas. The reactor building is accessible l1 for limited times in certain areas such as the in-core instrumentation control centers during normal plant operation. The radiation sources which provide the basis for the shield design are sub-divided into four categories according to their origin or location. a. The reactor core, internals, and reactor vessel. b. The reactor coolant. c. Auxiliary systems equipment. d. Radioactive materials released during accidents. Amendment 1 11.2-1

Radiation Shieldinn The radiation emanating from the reactor vessel consists of neutrons leck-ing from the ccre, prompt fission gammas, fission product gammas, and gammas resulting from the interaction of neutrons with steel and water. Nitrogen 16 is the major radiation source in the reactor coolant during normal operation and it establishes the secondary shield thickness. The shutdown radiation levels in the reactor loop areas are established by activated corrosion products and by assuming that 1% of the reactor fuel is defective allowing fission products to escape into the coolant. The sources in the auxiliary systems and waste disposal systems are established by both the corrosion product activities and the assumed fis-sion product activitics from 1% defective fuel circulating in the reactor coolant. The accident sources are established on the basis of the fisciou products released to the reactot containment following a hypothetical loss of reactor coolant, coupled with a failure of the safety injections system and subsequent core melting. 11.2.1.2 Description of Shielding 11.2.1.2.1 Primary Shield The pricary shield will be a large mass of reinforced concrete surrounding the reactor vessel and extending upward fnam the reactor building floor to iorm the walls of the fuel transfer canal. The preliminary shield thick-ness will be approximately 5 ft. up to the height of the reactor vessel flange where the thickness is reduced to approximately 4.5 feet. The primary shield will meet the following objectives: a. To reduce, in conjunction with the secondary shield, the radiation level from sources within the reactor vessel and reactor coolant system to allow limited access to the reactor building during normal full power operation. b. To limit the radiation level after shut down from sources within the reactor vessel to permit limited access to the reactor coolant system equipment. c. To limit neutron flux activation of component and struc-tural materials. 11.2.1.2.2 Secondary Shield The secondary shield will be a reinforced concrete structure surrounding the reactor crolanc equipment, including piping, pumps, and steam genera-tors. This shield will protect personnel from the direct gamma radiation resulting from reactor coolant activation products and fission products 12 0 11.2-2

rN Radiation Shielding carried away fram the core by the reactor coolant. In addition, the secondary shield will supplement the primary shield by attenuating neutron and gamnia radiation escaping from the primary shield. The secondary shield will be sized to allow limited access to the reactor building during full power operation. The preliminary thickness of secondary shield walls will be approximately 4.0 feet.. 11.2.1.2.3 Reactor Building Shield The reactor building shield will be a reinforced, prestressed concrete containment structure which completely surrounds the nuclear steam supply system. At full power operation, this shield will attenuate any radiation escaping from the primary-secondary shield complex such that radiation levels outside the reactor building will be less than 1 mrem /hr. In addi-tion, the reactor building structure will shield personnel from radiation sources inside the reactor building following a maximum hypothetical acci-dent (MHA). The shielding will be of sufficient thickness to allow person-nel a reasonable time period in which to evacuate the immediate vicinity of the reactor building following the >HL4 without excessive radiation exposure. The curves in Section 14 (Safety Analysis) indicate an integrated direct dose of less than 1 mrem over a period of two hours immediately outside the reactor building following the SMA. Preliminary thicknesses of the reactor building wall and dome are 3.75 ft and 3.25 ft respectively. ( ) 11.2.1.2.4 Centrol Room Shield The control room shielding will be designed for continuous occupancy for essential control room personnel following a maximum hypothetical occident. This would enable full control and shutdown procedures to be carried out without hazard to the centrol room operators. Preliminary thickness of the control room shielding is 1 ft. This ensures that the integrated whole body dose over 90 days following the MHA will not exceed 25 rems. Ventilation of the control room under post-accident conditions will be controlled as described in Section 9.7.2. 11.2.1.2.5 Auxiliary Shield Auxiliary shielding will include all concrete walls, covers, and removable blocks which will shield the numerous sources of radiation occurring in the radioactive waste disposal, makeup and purification, chenical addition and sampling systems. Typical compene,nts which require shielding include waste holdup tanks, boric acid concentrator and miscellaneous waste evaporators, makeup tank, waste gas decay tanks, demineralizers, makeup pumps, vaste drumming area, reactor coolant system drain tank, and reactor building sump pump. 00i 11.2-3

Radiation Chielding 11.2.1.2.6 Spent Fuel Shielding Shielding will be provided for protection during all phases of spent fuel removal and storage. Operations requiring shielding of personnel are spent fuel removal from reactor, spent fuel transfer through refueling canal and transfer tube, spent fuel storage, and spent fuel shipping cask loading prior to transportation. Since all spent fuel removal and trans-fer operations will be carried out under borated water, minimum water depths above the tops of the fuel assemblies will be established to pro-vide radiation shielding protection. Minimum allowed water depths during handling are approximately 10 feet in the reactor cavity and fuel transfer canal and approximately 13 feet over stored assemblies in the spent fuel storage area. The dose rates at the water surfate with a minimum water coverage of 10 feet will be less than 15 mrem /hr. While 13 feet of water limits the dose rate to less than 2.5 mrem /hr at the surface. The con-crete walls of the fuel transfer canal and spent fuel pit will supplement the water shielding and will limit the maximum continuous radiation dose levels in working areas to less than 2.5 mrea/hr. The refueling water and concrete walls also provide shielding from acti-vated control rod clusters and reactor internals which will be removed at refueling times. Although dose rates will generally be less than 2.5 mren/hr in working areas, certain manipulations of fuel assemblies, rod clusters, or reactor internals may produce short term exposures in excess of 2.5 mrem /hr. However, the radiation levels will be closely monitored during refueling operations to establish the allowable exposure times for plant personnel in order not to exceed the integrated doses specified in 10 CFR 20. 11.2.1.2.7 Materials and Structural Requirements The material used for the primary, secondary, reactor building, and auxili-ary shields will be ordinary concrete with density of approximately 140 lb/ ft3 Since the primary and secondary shielding walls serve as the refuel-ing structure: give support for the reactor coolant components under pipe rupture conditions, and provide missile shielding, they will be reinforced and designed to be self-supporting. Times of occupancy in restricted areas will vary depending on measured radiation levels in each area. Such areas as containment operating floor, reactor vessel head prior to refueling, primary loop compartments after shutdown, and spent fuel handling areas will be surveyed prior to access and a tine-limited work schedule will be set up. 11.2.1.3 Evaluation 11.2.1.3.1 Radiation Sources The shielding will be designed to attenuate neutron and gamma radiation emanating from the following basic sources: till 12A 11.2-4 a

Radiation Shielding a. Reactor core, internals, and reactor vessel b. Reactor coolant loops c. Radioactive material released during accidents d. Auxiliary systems equipment e. Spent fuel elements Source magnitudes are determined for the reactor operating at the maximum expected power level of 2568 Mwt with reactor coolant activity levels cor-responding to 1 percent failed fuel. Camma-ray yield and spectral dis-tributions from prompt fission and gross fission product activity are based on information in Volume III, part B, of the Reactor Handbook. The yield and spectral data for capture gammas are taken from ANL-5800, Reac-tor Physics Constants, and the Reactor Handbook. Data on activation pro-duct gamma rays are derived primarily from the Review of Modern Physics, Vol. 30, No. 2 (April 1451). The production of N-16 in the reactor coolant is calculated with a BC.4 code which ccmputes the integral of the 0-16 (n,p) N-16 cross section over the neutron flux in a water-cooled reactor, subject to variables in coolant flow and density and in neutron flux spectra and magnitude. The 0-16 (n,p) N-16 cross section used is that reported in WAPD-BT-25. Activities of individual fission products in the core, reactor, coolant, and reactor auxiliary systems are determined by a 4 B&W computer code designed to predict activities from a five-member radio-active chain at any point in the core history. Fission product leakage from the core to the coolant, and removal from the coolant by purification and leakage, are calculated. 11.2.1.3.2 Neutron and Gamma Shield The preliminary estimates for primary shielding requirements will be sub-sequently confirmed by Bechtel's diffusion and transport computer programs, 11.2.1.3.3 500\\ Dose Calculation The thickness of the reactor building shielding, in accordance with the design dose rate criteria, is based upon radiation 1cvels due to fission product release following a reactor accident. For the calculations it was assumed that 100 percent of the gases, 50 percent of the halogens, and 1 percent of the solid fission products were instantaneously released to the reactor building following a buildup period in the core of 600 full power (2,568 Mwt) days. The fission product activity was assumed to be uniformly dispersed through-out the reactor building volume, and the reactor building was represented i by a cylindrical source for the dose calculations. The integrated dose over various time intervals was computed with the GRACE II computer code s as id function of distance from the reactor building. The results are given j in Section 14. 11.2-5 12 3

Radiation Shielding 11.2.1.3.4 Operating Limits The radiation shielding design, including heating and dose rate profiles, temperature distributions, and coolant flow requirements, will be evaluated during the detailed design of the plant to establish the operatise limits. 11.2.1.3.5 Radiation Surveys Neutron and gamma radiation surveys will 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 shut-down will be formulated during the detailed plant design. 11.2.2 AREA RADI ATION MONITORING SYSTEM 11.2.2.1 Design Bases The fixed radiation monitoring system will be designed to indicate and alarm high radiation monitor!ng levels throughout the plant. An audible / visible alarm at both the detector location and the control room will be 1 provided. Recorded presentation may be required. All instrumentation for the radiation monitoring system will obtain its voltage supply from the 120 volt a-c essential service buses and each detector will have a loss-of-power alarm. The normal high radiation alarm setpoint will be 10 percent above the normal operational reading of the dete; tor. A maxi-mum alarm point will be set to correspond to the MPC va lue specified in 10 CFR 20. The maximum alarm point set at 10 CFR 20 values could be either an actual value or a calculated number corresponding to 10 C FR 20 limits. 11.2.2.2 Description Beta-gamma detectors are located as follows. a. One portal monitor and one hand and foot monitor at control access door b. Inside the reactor building near the personnel access hatch c. Near incore instrument space inside the reactor building d. Fuel handling bridge in spent f eel build ing e. Auxiliary building sump pump area f. Auxiliary building near sample sink ,eg g - UU g. Steam generator blowdown 12 4 9 11.2-6 Amendment 1 i

[#'N Radiation Shielding g. h.- Auxiliary building -in decay heat cooler area 1. Near component cooling water heat exchangers J. Radio-chemistry laboratory k. Primary loop 1. Control room m. Three battary power'ed monitors in reactor building .n. Condenser air ejector Air particulate and radio gas detectors to be mounted in the following places. a. Plant vent b. Inside reactor building c. Rcdio-chemistry laboratory d. Plant boundary e. Control room and auxiliary building Detector ranges will be determined depending upon the normal background at the detector locations and the calculated levels for abnormal conditions. -Radioactive test sources will be available to allow the overall system per-formance to be verified at regular intervals. 11.2.2.3 Evaluation Area radiation monitor detectors will.be located on the fuel-handling bridges to warn personnel if a high radiation level is approached during refueling operations. - A wide range detector will be mounted near the access hatch of the reactor building to indicate radiation levels inside the hatch before it is opened. The. upper range of the detector will be sufficiently high-to indicate the accessibility of_the. reactor building following a serious accident inside. The incore instrument area'will be monitored, and.a local alarm will be provided to warn-if a high radiation level exists or is created while .incore-assemblies-are being manipulated. The sample sink area in the ' auxiliary building will b'e equipped with a g'% detector to alarm an abnormal condition in connection with systen sampling. \\~ - g } 11.2-/

Radiation Shielding l Alarms will be actuated in the control room and at the detectors if an abnormal change in radiation background occurs. The radiation monitoring system shall be checked and calibrated at least once per month. When any portion of the radiation monitoring system requires maintenance, that unit shall be completely checked and calibrated immediately af ter completion of maintenance. 11.2.3 HEALTH PHYSICS The plant superintendent is responsible for radiation protection and con-tamination control. All personnel assigned to the plant and all visitors will be required to follow rules and procedures established by administra-tive control for protection against radiation and contamination. Under supervision of the plant superintendent, the administration of the radiation protection program will be the responsibility of the radiation protection engineer. It will be the responsibility of the health physics section to train plant personnel in radiation safety; to locate, measure, and evaluate radiological problems; and to make recommendations for control or elimination of radiation hazards. The health physics section will i function in an advisory capacity to assist all personnel in carrying out their radiation safety responsibilities and to audit all aspects of plant operation and maintenance to assure safe conditions and ccmpliance with the AEC and other federal and state regulations concerning radiation protection. Administrative controls will be established to assure that all procedures and requirements relating tc radiation protection are followed by all plant personnel. The procedures that control radiation exposure will be subject to the same review and approval as those that govern all other plant pro-cedures (see Section 12.6, Administrative Control). These procedures will include a radiation work permit system. All work on systems or locations where exposure to radiation or radioactive materials is or may be involved will require an appropriate radiation work permit. 11.2.3.1 Radiation Work Permits A radiation work permit shall be obtained by all personnel prior to enter-ing a control area or performing any work on radioactive or contaminated material or equipment. In the event that the safety of the plant or its personnel are endangered, entry may be made into a control area simultaneously with monitoring per-sonnel. A radiation work permit shall be completed as soon as possibic af ter correction of the situation. Radiation' work permits shall be issued routinely by the shift supervisor. These perm'its shall show: L,- 11.2-8

Radiation Shielding a. The nature of the work to be performed. b. Expected duration of work. c. Names of persons to perform the work. d. Signature of authorizing shift supervisor. Signature of an individual from the health physics groups e. who shall ensure that: (1) Designated personnel are within their permissible exposure limits. (2) The area has been adequately surveyed prior to entry. (3) Adequate protective clothing and supplies are avail-able at the control point. (4) Monitors are available for the work. All such permits shall be filed with the Health Physics group for future reference. i 11.2.3.2 Personnel Monitoring System The personnel monitoring program shall ensure that the recommendations and regulations of the Atomic Energy Commission are followed for all involved personnel. All personnel entering a control area shall wear a film badge or its equivalent. Exposures shall be maintained within the limits estab-lished in 10 CFR 20. In addition, those persons who ordinarily work in restricted areas or whose job requires frequent access to these areas will have pocket chambers, s el f-reading dosimeters, pocket high-radiation alarms, wrist badges, and finger tabs readily available for use, when required by plant conditions. This personnel monitoring equipment will also be avail-able on a day-to-day basis for those persons, employees, or visitors not assigned to the plant who have occasion to enter restricted areas or to perfor.a work involving possible exposure to radiation. Records of radia-tion exposure history and current occupational exposure will be maintained by the health physics group for each individual for whom personnel monitor-ing is required. The external radiation dose to personnel will be deter-mined on a da!.ly and/or weekly basis, as required, by means of the pocket chamber and dosimeter. Film badges will be processed monthly or more fre-quently when conditions indicate it is necessary. { 1 .h I l$1)I 11.2-9

Radiation Shielding 11.2.3.3 Personnel Protective Equipment Special protective or anticontamination clothing will be furnished and worn as necessary to protect personnel against contact with radioactive contamination. Change rooms will be conveniently located for proper util-ization of this protective clothing. Respiratory protective equipment will also be available for the protection of personnel against airborne radioactive contamination and will consist of full face filter masks, self-contained air-breathing units, or air-supplied masks and hoods. The first line of defense against airborne contamination in the work area is the ventilation system. However, respiratory protective equipment will be pro-vided should its use become necessary. Maintenance of the above equipment will be in accordance with the manufac-turer's recommendations and rules of good practice such as those published by the American Industrial Hygiene Association in its " Respiratory Protet-tive Devices Manual." The use and maintenance of this equipment will bc under the direct control of the health physics group, and personnel will be trained in the use of this equipment before using it in the performance of work. 11.2.3.4 Change Room Facilities Change room facilities will be provided where personnel may ot tain clean protective clothing required for plant work. These facilities -all be divided into " clean" and " contaminated" sections. The " contaminated" sec-tion of the change rooms will be used for the removal and handling of con-taminated protective clothing after use. Showers, sinks, and necessary monitoring equipment also will be provided in the change areas to aid in the decontamination of personnel. Appropriate written procedures will govern the proper use of protective clothing; where and how it is to be worn and removed; and how the change room and decontamination facilities for personnel, equipment, and plant areas are to be used. In order to protect personnel from access to high radiation areas that may exist temporarily or semipermanently as a result of plant operations and maintenance; warning signs, audible and visual indicators, barricades, and locked doors will be used as necessary. Administrative procedures will also be written to control access to high radiation areas. The Radiation Work Permit system will also be utilized to control access to high radia-tion areas. 11.2.3.5 Health Physics Facilities The plant will include a health physics facility and equipment for detect-ing, analyzing, and measuring all types of radiation and for evaluating any radiological problem which may be anticipated. Counting equipment (such O

~'s Radiation Shielding as G-M, scintillation, and proportional counters) will be provided in an appropriate shielded counting room for detecting and measuring all types of radiation as well as equipment (such as a multi-channel analyzer) for the identification of specific radionuclides. Equipment and facilities for analyzing environmental survey and bioassay samples will also be included in the health physics laboratory. Maintenance and use of the health physics laboratory facilities and equipment will be the responsi-bility of the health physics group. 11.2.3.6 Health Physics Instrumentation Portable radiation survey instruments will be provided for use by the health physics group as well as for operating and maintenance personnel. A variety of instruments will be selected to cover the entire spectrum of radiation measurement problems anticipated at the plant. Sufficient quantities will be obtained to allow for use, calibration, maintenance, and repair. This will include instruments for detecting and measuring alpha, beta, gamma, and neutron radiation. In addition to the portable radiation monitoring instruments, fixed monitoring instruments, i.e., count rate meters, will be located at exits from restricted areas. These instruments are intended to prevent any contamination on personnel, mate-rial, or equipment from being spread into unrestricted areas. Appropriate monitoring instruments will also be available at various locations within (^ 'N the restricted areas for contamination control purposes. Portal monitors will also be utilized, as appropriate, to control personnel egress from y restricted areas or from the plant. The plant will have a permanently installed remote radiation and radio-activity monitoring system for locations where significant levels can be expected. This system will monitor airborne particulate and gaseous radio-activity as well as external radiation levels. This system will present an audible alarm and radiation level indication in the area of concern in addition to the control room. 11.2.3.7 Medical Programs Facilities for screening personnel for contamination will be available on site with outside services utilized as backup and support for this program. A medical examination program appropriate for radiation workers will be conducted to establish and maintain records of the physical status of each employee. Subsequent medical examinations will be held as determined necessary for radiation workers. Medical doctors, preferably in the local area, will be used for this program. The health physics group will be responsible for the program and will assist the physicians to preserve the health of the employees concerned and to confirm the radiation control methods employed at the plant. ~ g. bi 00i 11.2-11

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11.3 REFERENCES

1. Frank, P. W., et al., Radiochemistry of Third PWR Fuel Material Test - X-1 Loop NRX Reactor, WAPD-TM-29, February 1957. 2. Eichenberg, J. D., g d., Effects of Irradiation on Bulk UO3, WAPD-183, October 1957. 1 3. Allison, G. M. and Robertson, R. F. S., The Behavior of Fission Products in Pressurized-Water Systems. A Review of Defect Tests on CO2 Fuel Ele-ments at Chalk River, AECL-1338, 1961. l 4. Allison, G. M. and Roe, H. K., The Release of Fission Gases & Iodines From Defected CO2 Fuel Elements of Different Lengths, AECI-2206, June 1965. l 000 0130 11.3-1 l , _ _ _ _ _ _ _,. _.. _ _ _ _ _ _}}