ML20096F107
| ML20096F107 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 03/30/1992 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 9205200194 | |
| Download: ML20096F107 (135) | |
Text
{{#Wiki_filter:, , ABWR assign occument 3.7 Radiation Protection Ds. sign Description The AllWR design prosides r.uliation protection features that will keep exposures for both plant personnel and the general public well below allowable limits. These .ow exposure conditions are achieved by an integrated approach that recogniws the contribution of both shielding provie ns and ventilation system designs that conuol air borne contaminants. 51onitoring of radiation levels is an integral part of the plant radiation protection strateg. The plant design procides radiation shielding for rooms, corridors and operating areas commensurate with their occupancy requirements and thus _ maintains radiation exposures to plant personnel as low as reasonably achievable. Staintenance of plant compoacnts is achieved without significant radiation exposura from adjacent plant s> ems or equipment by use of shielded cubicles, labyrinth access and prosisions f( temporary shielding. Under accident conditions, plant shielding designs permit operators to perform required safety functions in sital areas of the plant, in addition to protection of operating personnel, the plant design proddes radiation shiciding which maintains radiation exposure to the general public as low as is reasonably achievable. Plant ventilation systems insure that concentrations of airborne radionuclides are maintained at levels consistent wi.h personnel access requirements. In addition, airborne radioactisity monitoring is prosided for those normally occupied areas of the plant in which there exists a significant potential for airborne contamination. Inspection, Test, Analyses and Acceptance Criteria Tables 3.7a and 3.7b provide a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the AIMR plant shielding, ventilation and airborne monitoring equipment. O 17 1 3/30/92 9205200194 920330 PDR ADOCK 05200001 PDR
Table 3.7a: Plant Shielding Design inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Anelyses Acceptance Criteria
- 1. The plant design shall provide radiation 1. An analysis of the expected radir-tion lefeb 1. Maximum expected radiation levels are shielding for rooms, corridors and in each plant area wi!! Se oorformed to well within (25% or less) of the radiation operating areas commensurate with their verify ths adequacy of the shielding zone designation, for each p. ant area, as 2
occupancy requirements to maintain dasign. Tnis analysis shall consider the indicated in Figures 3.7.a through 3.7.bb. radiation exposures to plant personnel as fol!owing: low as reasonably achievabis.
- a. Confiernatory calculations shall consider all significant radiation sources (greater than 5% contribution) for an area. Radiation source strength in plant systems and components will be determined based upon an assumed source term of 100,000 pCurie!second offgas release rate (after 30 minutes decay), a 200 pCurie/ gram-steam N-16 9 trource term at the vessel exit nozzle,
% and a core inventory commensurate with a 4005 MWT equilibrium core at 51.6 kwatt/ liter. All source terms sha!! be adjusted for radiological decay and buildup of activated corrosion and wear products.
- b. Commonly accepted shielding codes, using nuclear properties derived from well known references (such as Vitamin C and ANSI /ANS4.4) shall be used to model and evaluate plant radiation environments.
- 1) For non-complex geometries, point kernal shielding todes (such as QAD or GGG) shall be used.
- 2) For complex geometries, more sophisticated two cr three dimensional ,
transport codes (such as DORT or e TORT) shall be used. w - 9 9 9
p I w -3 4 V - Tatde 3.7a: Plant S.S'nw Design (Continued) . Inspections. Tests, Analyses and Acceptance Criteria
- Certified Design Commitment inspections, Tests. Analyses Acceptance Criteria
, 1. (Cont.)
- c. In any calculatbn. a safety factor shall be applied based upon benchmark j comparisons of the code and data i~ coilccted from known and measured I ' environtnents.
I
- 2. The plant design shall provide shielded 2. Using the methods identified in (1) above. 2. Shielding design (with temporary shielding cubicles, labyrinth access, and space for radiation levels present in areas where installed. where appropriate) is such that
- c temporary shielding to allow for .
maintenanceis performed shat:be radiation from adjacent areas sha!! i maintenance of plant components without evaluated for the contribution from contribute no more than a small fraction 4 significant radiation exposure from adjacent high radiation areas and (10% or less) of the radiation field intensity < adjacent plant systems or equipment. equipment. or less than 0.06 mrem /hr whictever is I farger, in plant areas where maintenance is performed.
- 3. The plant radiation shielding design shall 3. An analysis of the expected high radiation 3. Under accident conditions, radiation permit operators to perform required levels in each area which will or may shielding design allows access, occupancy
] safety functions in vital areas of the plant require occupancy to permit an operator to and egress of vital areas such that j (including access and egress of these aid in the mitigation of or recovery from an personnel radiation exposures do not areas) under accident conditions. accident (vital areal shall be performed to exceed 5 rem to the whole body or its ,' verify the adequacy of the plant shielding equivalent. for the duration of the accident i design. This analysis shall use ' (based on the required frequency of access j calculational methods consistent with (1.b) to each vital areal. For areas requiring continuous occupancy (such as the control ~ above and a radiation source term (adjusted for radioactive decay) based on room). local radiation hot spots shall not th' following:. exceed 15 mrem /hr (averaged over 30 days).
- a. Liquid containing systems: 100% of the core equilibrium nobie gai, inventory,50% of the core equilibrium halogen inventory and 1% of all others are assumed to be mixed in the reactor coolant and recirculation liquids
( recirculated by the residual heat e removal system (RHR). the high 5 t t x.< -y-
- ~ w ..-y.,. m , - - - , ~g s,, r , , , _ , . . -. , _
Table 3.7a: Plant Shielding Design (Continued} Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests Analyses Acceptance Criteria
- 3. (Cont.)
pressure core flooder (HPCF), and the reactor corn isolation cooling (RCIC) systems.
- b. Gas containing systems- 100% of the core equilibriem noble gas inventory and 25% of the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere. For vapor containing systems (such as the main steam lines) these core inventory fractions are assumed to be contained in the reactor o coolant vapor space.
W k 4. The plant design shall provide radiatior. 4. Using the methods identified in (1) above. 4. The radiation dose to the maximal y shielding to maintain radiation exposure to the radiatier. dose to the maximally exposed member of the public is a small the general public as low as is reasonably exposed member of the general public fraction (10% or less) of the dose hmit to a achievab;a. from direct and scattered shall be member of the publ.c hsted in 40CFR190. determined. O O O 4
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Table 3.7b: Ventilation And Airtiorne Monitoring 4
; inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment lespections, Tests, Analyses Acceptance Criteria .
- 1. Plant design shall provide adequate - 1. Expected concentrations of airborne 1. Calculation of radioactive airborne containment of airborne radioactive radioactive matenal shall be calculated by - concentration shall demonstrate that-j - materials and the ventilation system will nuclide for normal plant operations,
, ensure that concentrations of airbome - anticipated operational occurrences for a. For normaily occupied rooms and , radionuclides are maintained at levels - each equipment cubicle, corridor, and areas of the plant (i.e. those areas consistent with personnel access operating area seguiring personnel requiring routine access to operate and
! requirements. access. Calculations shall consider- maintain the plant) equilibrium concentrationsof airborne nuchdeswill
- a. Design ventilation flow rates for each be a small fraction (10% or less) of the area, occupational concentration limits listed t in 10 CFR 20 Appendix B.
- b. Typical leakage characteristics for equipment located in each area and b. For rooms that require irifrequent i access (such as for non-routine ,
j c. A radiation source term in each fluid equipment maintenance) the ; w system shall be determined based ventilation system shall be capable of
- $ upon an assumed offgas rate of reducing radioactive airborne '
4 100,000 Curie /second (30 minute concentrations to (and maintait ing
; decay) appropriately adjusted for them at) the occupational l radiological decay and buildup of concentration limits listed in 10CFR20 i activated corrosion and wear products. Appendix 8 during the periods that occupancy rs required. f
- c. For rooms that seldom require access (such as tank rooms), plant design shell provide sufficient containment and ventilation to ensure airbome contamination does not spread to other areas.
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Table 3.7b: Ventilation And Airborne Monitoring (Continued) inspections, Tests, Analyses ar.d Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criieria
- 2. Airborne radioactivity monitoring shall be 7 An analysis shall be performed to identify 2. Airborne radioactivity monito.ing system provided for those normally occupied the plant areas that require airborne shalt:
areas of the plant in which there exists as radioactivity monitoring. significant potential for airborne a. Have the capability of detecting the contamination (greater than 0.1 per year) time integrated change in concentrations of the most limiting particutate and iodine radionuclides in each area equivalent to the occupational concentratiors limits in 10CFR20. Appendix B for 10 hours.
- b. Provide a calibrated response, representatvie of the concentrations within the area (i.E. air sampling monitors in ventilation exhaust a streams shall collect and isokineoc 9
m samplet
- c. Provide local audible afarms (visual alarms in high noise areas) with variable alarm set points and readout!
annunciation capability in the control room. is 5 9 9 9 -
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-'L " SECTION 12.3 - CONTENTS Title
) Section - Pare J12J.11 Facility Deslan Features ' 123 1 12 3.1.1- Equipment Design for Maintaining Exposure ALARA -123-1
- 123.1.2 Plant Design for Maintaining Exposure (ALARA) - 12 3-3 t 123.13 Radiation Zoning - . 123-5 123 14-- Implementation of AIARA 12 3-6 12 3.1.4.1 Reactor Water Cleanup System . 12 3-6 123.1.4.2- Residual Heat RemovalSystem (Shutdown Cooling htode)- ~ 12 3-7 123.1.43 Fuel Pool Cooling and Cleanup System 123-7 \
- - 123.1.4.4 -' - Main Steam System 123-8
-12 3.1.4.5 - Standby Gas Treatment System 123 8 12 3.2 Shieldine . 12 3-9
' 123.2.1- Design Objectives - 123 '
~ 123.2.2 ' Design Description 123-9 123.2.2.1 General Design Guides -123-9 g 12 3.2.2.2 Method of Shielding Design = 123 a-123.23;
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Plant Shielding Description - 123-11
-1233 Ventilation 12313.1 ~
'1233.1. Design Objectives ' 12 3-13.1
^ 1233.2 Design Description 12 3-13.1
-1233.2.1 Control Room Ventilation = - 12 3-13.1'
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Standard Plant RPVJ SECTION 12.3 a i CONTENTS (Contin'ued) Section Title Pace
;1233.2.2 Dr)well -123-13.2 1233.23 - Reactor Building 123-13.2 123 3.2.4 Radwaste Building 123-13.2 12.3.4' Men Radiation and Airborne Radioactivity Monitors '12 3-14 12 3.4.1 System Objectives 123 14 123.4.2 System Description 123-14 123.43 . System Design 123 14
- 12 3.5 Post Accident Access Reautrements 123-15 ,
- 12J.6 Post Accident Radiation Zone Mans 123 15 12 3.7 - Deleted 12 3-15 l 12J 8 References 123-15 '
TABLES Table Title Eage
.123-1; Computer Codes Used'in Shielding Calculations 123 l6
--123 2- 1 Typical Nickel and Colbalt Content of -
Materials 123-17 12 3-3 Area Re.liation ' Monitor, Reactor Building - 123-17.1-
- 123-4 '- 1 Area Radiation Modtor, Control Building 123-17.2 -
113-5: Area Radiation Monitor, Service Building - 123-17.2-
- 123 Area Radiation Monitor, Radwaste Building 123-173 4 12 3-7 Area 'iation Monitor, Radwaste Building 123-17.4 123-iii Amendment 20
, . ABWR - _ "
mm Standard Plant nry n SECTION 12.3 ILLUSTRATIONS - Figure Title Page 12 3-1 Reactor Buildin'g Radiation Zone Map for Full Power and Shutdown Operations at Elevation -8200mm (B30 123 18 1123 Reactor Building Radiation Zone Map foi Full Power . and Shutdown Operations at Elevation 1700mm (B2F) 123-19
-12 3-3 Reactor Building Radiation Zone Map for Full Power I and Shutdown Operations at Elevation 4800mm (B1F) 123-20 il23 - Reactor Building Radiation Zone Map for Full Power and Shutdown Operations at Elevation 8500mm (B1M) 123 21 123-5 Reactw Building Radiation Zone Map for Full Power and Shutdown Operations at Elevation 12300mm (IF) 123 22 123-6 Reactor Building Radiation Zor;e Map for Full Power and Siiutdown Operations at Elevation 18100mm (2F) 123-23 123-7 ; Reactor Building Rafiation Zone Map for Full Power .
and Shutdown Operations at Elevation 23500mm (3F) 123-24 123-8 Reactor Building Radiation Zone Map for Full Power
- end Shutdown Operations at Elevation 27200mm (4F) 12 3-25. .
123-9 - Reactor BuPding Radiation Zone Map for Full Power
~
and Shutdown Operations at Elevation 31700mm (4FM) 123-26 123-10 ~ ~ Reactor Building Radiation Zone Map for Full Power
- and Shutdown Operations at Cross Section View A A'
, 123-27 . ,
A c123-11 l ' Reactor Bui' ding Radiatio, Zone Map for Full Power _ _
; and Shutdown Operations at Cross Section View B-B 12 3-28
- 123-12 L Reactor Building. Radiation Zone Map Post LOCA at
- Elesation -8200mm (B3F) 123-29 ,
123-iv m t
' Amendment A
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SECTION 12.3 ILLUSTRATIONS (continued) . FigupeL Title Eags '
= -
' 12313 < _ Reactor Building Radiation Zone Map Post LOCA at Elevation 1700mm (B2F) - 12 3 ,
, 123-14 Reactor Building Radiation Zone Map Post LOCA at Elevation 4800mm (B1F) .123 31 1 123 o Reactor Building Radiation Zone Map Post LOCA at
- Elevation -8500mm (B1M). 12 3-32
.123-16-- - Reactor Building Radiation Zone Mao Post 1 OCA at
! Elevation 12300mm (IF) , 123-33 123-17 Reactor Building Radiation Zone Map Post LOCA at Elevation 18100mm (2F) 12 3-34
-12 3-18"- Reactor Building Radiati ' lone. Map Post LOCA at Elevation -23500mm (3F) 123 35 i 12 3-19 : . Reactor Building Radiation Zone Map Post LOCA at Elevation 27200mm (4F) 123-36 123-20 . Reactor Building Radiation Zone Map Post LOCA at Elevation 31700mm (4FM) ~ 123-37
;123 21 l Reactor Building Radiation Zone Map Post LOCA at Cross Section A-A. 123-38 123-22; ' Reactor Building Radiation Zone Map Post E.OCA at Cross Section B-D 12 3-39
- h -123-23 Deleted [ I 123-24 I Deleied [, l, _ 123 v
Amendment
, , ABWR :wimat. Standard Plaat uv. n SECTION 12.3 ILLUSTRATIONS (continued) Eigurs Title Page 123 25 Deleted 123-26 Deleted 123-27 Deleted 123-28 Deleted 123-29 Deleted 123-30 Deleted .
.123-31 Deleted 123-32 Deleted 123-33 Deleted .
12 3-34 Deleted 12 3-35 Deleted 12 3-36 Radwaste Building Equipment List 123-49 123-37 Radwaste Building, Radiation Zone Map, Normal Operation at Elevation (-)6,500mm 123-51 123-38 Radwaste Building, Radiation Zone Map, Normal Operation at Elevation (-)200mm 123-5.- 123-vi Amendment
. _. . = -,. . ._ _ . , . .-_ _- . ._ _ _ _ .. ._
n y LABWIk'i numt, ,- L Standard Plant -- nty. n SECTION 12.3 E -. ILLUSTRATIONS (continued) Figure Tule Eage 123-39 R'adwaste Building, Radiation Zone Map, Normal Operation at Elevation 7,300mm 12 3-53 123-4d - Radwaste Building, Radiation Zone Map, Normal Operation at Elevation 16,000mm 123-54 123-41 Radwaste Building, Radiation Zone, Normal Operation at CrossSection A A' 123 55 123-42 Control Building, Radiation Zone, Normal Operation at Floor Level (-)13,150mm 12 3-56
- 123-43 Control Building, Radiation Zone, Normal Operation at Floor level (-)7,100mm 123-57
- 123-44 Control Building Radiction Zone, Normal Operation ,
- at Floor Level (-)1,450mm . 123 .
.123-45: Control Building, Radiation Zone, Normal Operation -
at Floor Level 2,900mm 12 3-59
'123-46 Control Building, Radiation Zone, Normal Operation
- at Floor Level 7,350mm 123-60
= 123 Contro; Building, Radiation Zone, Normal Operation at Floor Level 13,295mm 123 61
-- 12 3-48 Control Building, Radiation Zone, Normal Operation, Side View - 123-62
= 123-49; Turbine Building, Radiation Zone, Normal Operation
? at Elevation 53M 123-63 l123 Turbine Building, Radiation Zone, Normal Operation
- [ at Elevation 123M 123-M 123-51 Turbine Building, Radiation Zone, Normal Operation at Elevation 203M 123-65 123-52 Turbine Building, Radiation Zone, Normal Operation l at Elevation 303M - 123-66 123 sii Amendment
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,'o ': Standard Plant nnv. n !
. SECTION 12.3 A ILLUSTRATIONS (continued);
,n Figure Iltle Page 123 53" Turbine Building, Radiation Zone, at ' l LongitudinalSection A A' 123-67 123-54 Control Building, Radiation Zone, Post LOCA, Side View 112348 123 Turbine Building, Radiation Zone, Post LOCA,
; Longitudinal Section 123-69 123-56 Reactor Building, Area Radiation Moaltors,(-)8.2m 12.3-70 12 3 Rcactor Building, Area Radiation Monitors,1.7m & 1.5m 123 71
'123 58- Reactor Building, Area Radiation Monitors,4.8m 1234?.
123 Reactor Duilding, Area Radiation Monitors,123m 123 73
'12340- Reactor Building, Area Radiation Monitors,23.5m 123 ,
i
- 12 3-61
- Reactor Building, Area Radiation Monitors,27.2m -123-75 123-62 l Reactot Building, Area Radiation Monitors,31.7m 123-76
-123 : Reactor Building, Area Radiation Monitors, Section 270/90 123-77
= 123-64 . Control Building, Area Radiation Monitors . , 123-78 123 . Radwaste Building, Area Radiation Monitors, (-)6.5m -123 79
- .123-66 . Radwaste Building, Area Radiation Monitors,(-)0.2m 123-80 2123-67-- ' Radwaste Building. Area Radiation Monitors,7.3m 123 81 i'
. 123-68 Radwaste Building, Area Radiation Monitors,16m 123-82 6 123-69 = Deleted - 123-83
?
123-70 Turbine Building, Level 2, Area Radia: ion ' 7 l- Monitor Elevation 123m 12 3-84 . 123-71 Turbine Building, Level 3, Area Radiation Monitor Elevatma 203m 123-85
]
123-viii 4 Amendment
ABWR uumu - - Sipndard Plant Eto;E SECTION 12.3 ILLUSTRATIONS (continued) Elgure Ilite lhe 123-72 Turbine Building,Ixv - < Area Radiation Monitor Elevation 25.'- 123 86 123-73 Turbine Building, Longitudinal Section AA, Area Radiation Monitors 123-87 t 123-ix Amendment 18
1 i . - ABWR == Standard Plant tuv n 12.3 RADIATION PROTECTION cmrosion resistance) and for which no suitable DESIGN FEATURES alternative low nickel material is available. Cobalt content in the inconel X750 used in the l 12.3.1. Facility Design Features fuel assemblies is limited to 0.05%. The ABWR Standard Plam is designed to meet Stellite is used for hard f acing of the intent of Regulatory Guide 8.8 (i.e., to keep components which must be extremely wear radiation exposures to plant personnel as low as resistant. Use of high cobalt alloys such as reasonably achievable (ALARA)). This section Stellite is restricted to those applications describes the component and system designs in where no satisfactory alternative material is addition to the equipment layout employed to available. An alternative material (Colmonoy) maintain radiation exposures ALARA. Consider- has been used for some hard facings in the core ation of individual systems is provided to area, illustrate the application of these principles. 12.3.1.1 Equipment Design for Maintaining Material application for primary coolant Exposure AIARA piping, tubing, vessel internal surf aces, and other components in contact with the primary This subsection describes specific components l coc! ant is discussed in the following pages, as well as system design features that aid in Typical nickel and cobalt contents of the maintaining the exposure of plant personnel principal materials applied are given in Table during system operation and maintenance ALARA. 12.3 2. Equipment layout to prodde A1 ARA exposures of, plant personnel are discussed in Subsection ' Carbon steel is used in a large portion of the 12.3.1.2. system piping and equipment cutside of the nuclear steam supply system. Carbon steel is (1) Pumps typically low in nickel content and contains a very small amount of cobalt impurity. Pumps located in radiation areas are designed to minimize the time required for Stainless steel is used in portions of the maintenance. Quick change cartridge type system such as the reactor internal components seals on pumps, and pumps with back pullout and heat exchanger tubes where high corrosion features that permit removal of the pump resistance is required. The nickel content of impeller or mechanical seals without the stainless steels is in the 9 to 10.5% range disassembly of attached piping, are employed h and is comrolled in accordance with applicable to minimize exposure time during pump ASME material specifications. Cobalt content is maintenance. The configuration of piping controlled to less than 0.05% in the XM-19 alloy about pumps is designed to provide used in the control rod drives, s u f ficie n t space for efficient pump maintenance. Provisions are made for A previous review of materials certifications slushing and in certain cases chemically indicated an average cobalt content of only 0.15% cleaning pumps prior to maintenance. Pump in austenitic stainless steels, casing drains provide a means for draining pumps to the sumps prior to disassembly, Ni-Cr Fe alloys such as Inconel 600 and thus reducing the exposure of personnel and Inconel X750, which have high nickel content, are decreasing the potential for contamination. used in some reactor vessel internal components. Where two or more pumps conveying highly These materials are used in application. for radioactive fluids are required for opera-which there are special requirements to be tional reasons to be located adjacent to satisfied (such as possessing specific thermal each other, shielding is provided between expansion characteristics along with adequate the pumps to maintain exposure levels Amendment 10 12 3-1
ABWR mm Standard Plant yg n ALARA. An example of this situation is the that could lead to radioactive crud deposi-RWCS circulation pumps. . Pumps adjacent to tion. Connections are available for conden-other highly radioactive aquipment are also sate or demineralized water flushing of the shielded to reduce the maintenance exposure, heat exchangers. For the reactor water for example, in the radwaste system. clean up (CUW) system, separate connections are provided to chemically decontaminate Whenever possible, operation of the pumps both the heat exchangers (both regeneratise and associated valving for radioactive and non regenerat'e) and the pumps. The systems is accomplished remotely. Pump other main heat exchangers (RHR and RIP) are control instrumentation is located outside provided connections by which the exchangers high radiation areas, and motor- or can be flushed with clean water. The last pneumatic-operated valves and valve main heat exchanger, the luci pool heat extension stems are employed to allow exchanger, is downstream of the filter operation from outside these areas, demineralizer and is therefote not subNcted to flows containing significant amounts of (2) Instrumentation fission or s ivation products. In all cases, the pumps directly invelved with the Instruments are located in low radiation heat exchangers are also inline for decon-areas such as shielded valve galleries, tamination with the exchangers.. Instru-corridors, or control rooms, whenever mentation and valves are remotely operable possible. Shicided valve galleries provided to the maximum extent possible in the for this purpose include those for the RWCS, shicided heat exchanger cubicles, to reduce , FPCC, and radwaste (cleanup phase sepa ator, tiie need for entering thest high radiation . spent resin tank, and waste evaporator) areas. systems. Instruments required to be located in high radiation areas due to operations (4) Valves requirements are designed such that removal of these instruments to low radiation areas Valve packing and gasket material are for maintenance is possibiv. Sensing lines selected on a conservative basis, accounting are routed from taps on the primary system for environmental conditions such as in order to avoid placing the transmitters temperature, pressure, and radiation or readout devices in high radiation areas. tolerance requirements to provide a long For example, reactor water level as well as operating life. Valves have back seats to ' recirculation system pressure sensing minimize the leakage through the packing. instruments are located outside the drywell. Straight through valve configurations acre selected where practical, over those which Liquid service equipment for systems exhibit flow discontinuities or internal containing radioactive fluids are provided crevices to minimize crud trapping. Teflon with vent and backflush provisions. gaskets are not used. Instrument lines, except those for the reactor vessel, are designed with provisions Wherever possible, valves in systems for backflushing and maintaining a clean containing radioactive fluids are separated fill in the sensing lines. The reactor from those for "clea.- services to reduce vessel sensing lines may be flushed with the radiation exposure irom adjacent valves condensate following reactor blowdown. and piping during maintenance. (3) Heat Exchangers Pneumatic or mechanically operated valves are employed in high radiation areas, Heat exchangers are cons:ructed of stainless whenever practical, to minimi7e the need for steel or Cu/Ni tubes to minimize the pos- entering these areas. For certain sibility of failure and reduce maintenance situations, manually operated valves are re q uire m e r.t s. The heat exchanger design required, and in such cases extension valve allows for the. complete drainage of fluids stems are provided which are operated from a from the exchar.ger, avoiding pooling effects shielded area. Flushing and drain provi-Amendment 20 1212
s ABWR 2mmixi. 1 SARadlinll'llll11 JiLLil sions are employed in radioactive systems to reduce exposure to personnel du ing maintenance. For ascas in which especially high radiation leve's are encountt .d. valving is reduced
'o the maximum cytent possible with the b tlk of the valve and piping located io ,.n adjacent valve gallery where the radiation levels are lower.
Amendment 20 123 21
/ N 2iMi%At, SandarMlant .._..J a n (5) Piping the radiation exposure of personnel during maintenance. The dampers located in the i
Piping was selected to provide a service cubicles are remotely operated, .hus i life equivalent to the design life of the requiring no acccu to the cubicle during plant, with consideration given to corrosion operation. A pneumatic transfer system is allowances and environmental conditions, employed to remove the radioactive ch,,rcoal l Piping for service in radioactive systems from the filter, requiring entry into the such as the RWC system hase butt welded shielded cubicle only during the connection conn 5ctions, rather than socke! welds, to of the hoses to the SGTS filter unit. reduce crud traps. Distinction is made between piping conveying radioactive and nonradioactive fluids, and separate routing 12.4.1.2 Plant Design for Maintaining bposure is provided whenever possible. Piping (ALARA) conveying highly radioactive fluids is l usually routed through shielded pipe chases This subsection describes features of - and shielded cubicleb. 1lowever, when these equipment layout and design which are employed options are et feasible, the radioactive to maintain personnel exposures ALARA. piping is em sdded in concrete walls and floor s. (1) Penetrations (6) Lighting Penetrations through shield walls are voided whenever possible to reduce the, Lighiing is designed to provided sufficient numbcr of streaming paths provided by these ,
'llumination in radiation areas to allow penetrations. Whenever penetrations are quick and efficient surveillance and required through shield walls, however, they maintenance operations. To reduce the need are located to minimire the impact on for immediate replacement of defective surrounding areas. Penetrations are located bulbs, multiple lighting fixtures are so that the radiation source cannot 'sce*
provided in shicided cubicles. Considera. through the penetration. When this is not tion is also given to locating lighting possible, or to provide an added order of fixtures in easily accessible locations, reduction, penetrations are located to exit thus reducing the exposure time for bulb far above floor level in open corridors or replacement. in other relatively inaccessible areas. Penetrations v.hich are offset through a - (7) Floor Drains shield wall are frequently employed for electrical penetrations to reduce the Floor drains with appropriately sloped streaming of radiation through these floors are provided in shielded cubicles penuraDous, where the potential for spills exist. These drain fines having a potential for Where permitted, the annular region between - conttiining highly radioactive fluids are pipe and penetration sleeves, as well as routed through pipe chases, shielded cicettical penetrations, are filled with cubicles, or are embedded in concrete walls shielding material to reduce the streaming and floors. Smooth epoxy type coatings are area presented 5 these penetrations. The employed to facilitate decontamination when shieIding materia 1s used in these a spill does occur. applications inchtde a lead-loaded silicone foam, with a density comparable to concrete, (8) SGTS Filters and a boron loaded refractory type material for applications requiring neutron as well The SGTS filter is located in 1. separate as gamma shiniditig. There are certain shicided cubicle and is separated oy a penetrations wb- e these two approaches are shield wall from the exhaust fans to reduce not feasibic or are not sufficice.ly Amendment 10 12 M
AlnVR ==. l. Standat31flADt Rtv y effective, la those cases, a shielded l'or situations 19 which radioactive piping enclosure about the penetration as it exits must be routed through corridors or other in the shield wall, with a 90 degice bend of low radiation areas, an analysis is the process ripe as it e mit s t he conducted to ensure that this routing does pe netr ation, is employed. not compromise the existing radiation roning. (2) Sample Statks 11adioactive services are touted separately Sample stations in the pinut provide for the frorn piping containing nonradioactive routine surveillance of reactor water fluids, whenever possible, to minimire the quality. These sample stations are located exposure to personnel duilng maintenance. In low radiation areas to reduce the When such routing combinatione. are required, exposure to operating personnel. Flushing however, drain provisions are r$revided to provisions are included using dem!nerahred remove the radioactive fluid contained in water, and pipe drains to plant sumps are equipment and piping. *Cican" services and provided to minimite the possibility of radioactive piping are required at times to spills. Fume hoods are employed for be routed together in shielded cubic!cs, in airborne contamination control. Iloth such situations, provisions are made for the working areas and fume hoods are constructed valves required for process operation to be of polished stainless steel to case controlled remotely, without need for t . contamination if a spill does occur. Grab entering the cubicle. spouts are located abose the sink to reduce , the possibility of contaminating surrounding Netrations for piping through shield walls areas during the, sampiing process. are designed to minimite the impact on ' suerounding areas. Approrches used to (3) IIVAC Systerns accomplish this objective are described in Subsection 12.3.1.2.1. Major llVAC equipment (blowers, coolers, and the like) is located in dedicated low Piping configurations are designed to radiation arcus in maintain exposures to minimize the number of " dead lege and low personnel maintaining these eqtupment points in piping runs to avoid accumulation ALARA. IIVAC ducting is routed outside pipe of radioactive crud and fluids in the line, chases and does not penetrate pipe chase Drains and flushing provisions are employed walls, which could compromise the whenever feasible to reduce the irnpact of
- hielding. 11VAC ducting penetrations required
- dead legs" and low points.
through walls of shleided cubicles are Systems containing radioactive fluids are located to minimize the impact of the welded to the most practical extent to streaming radiation levels in adjoining reduce leakage through flanged or screwed areas. Addition al IIV A C d e sign connections. For highly radioactive considerations are addressed in Subsection systems, butt welds are employed to minirnire 12.3.3. crud traps. Provisions are also made in radioactive systems for flushing with 3 (4) Piping condensate or chemically cleaning the piping to reduce crud buildup. Piping containing radioactive fluids is routed through shleided pipe chases, (5) Equipment layout shielded equipment cubicles, or embedded in concects; walls and floors, whenever Equipment layout is desigred to redute the possible. " Clean" services such as exposure of personnel requ' red to inspect or compressed air and demineralized water are maintain equipment. " Clean
- pieces of not routed through shleided pipe chases. equipment are located separately from those Amendment 10 ti L4
I ABWR mai. : StattdanifhtnL ntst! l which are sources of radiation whenever CRD removal under the reactor pressure possible. For systems that have components vessel and in the CRD maintenance room, that are major sources of radiation, piping and pumps are located la separate cubicles Appropriately sloped floor drains are to reduce exposure from these components provided in shielded cubicles and other during maintenance. These major radiation areas where the potential for a spill exists sources are also separately shielded from to limit the extent of contamination. Curbs each other. are also provided to limit contamination and simplify washdown operations. A cask
- 16) Contamination Control devontamination vault is located in the reactor building where the sperr: M. sk Contaminated piping systems are welded to and other equipment may be cleani e The CRD the most practical extent to minimite leaks maintenance room it. used for disassembling through screwed or flanged fittings. For control rod delves to reduce the systems containing highly radioactisc contamination potential, fluids, drains an hard piped directly to equipment drain sumps, rather than to allow Consider tlen is given in the design of the contaminated fluid to flow across the floor plant for reducing the effort required for to a floor drain. Certain valves in the decontamination. Epoxy type wall and floor main steam line are also provided with coverings have been selected which provide leakage drains piped to equipment drain smooth surfaces to case decontamination sumps to redu> contamination of tbr steam surfaces. Expanded metal-type floor tunnel. Pump casing drains are employed on gratings are minimited in favor of smooth, radioactive systems whenever possible to surfaces in areas where radioactive spills '
remove fluids from the pump prior to could occur. Equipment and floor drain disassembly. In addition, provisions for sumps are stainless steel lined to reduce flushing with condentate, and in especially crud buildup and to provide surfaces easily contamirsted systems, for chemically decontaminated, cleaning th" 'quipment prior to maintenance, ate providect 12.3.1.3 Radiation Zoning The llVAC system is designed to limit the Radiation zones are estabibNd in all areas extent of airborne contamination by of the plant as a function of both the access providing air flow patterns from areas of requirements of that area and the radiation low contamination to more contaminated sources in that area. O pe r ating _ activities, areas. Penetrations through outer walls of inspection requirennts of equipment, the building containina. radiation sources maintenance activities, W abnormal operating are scaled to prevent miscellaneous leaks conditions are considered in determining the into the environment. The equipment drein appropriate zoning for a given area. The sump vents are fitted with charcoal relationship between radiation rone designations canisters or piped directly to the radwaste and accessibility requirements is presented in llVAC system to remove airborne contaminants the following tabulation: evolved from discharges to the sump. Wet transfer of both the steam dryer and Zone separator also reduces the likelihood of Desig- Dose Rate Access contaminants on this equipment being nailen LtnRam /hr) Descriotion released into the plant atmosphere. In areas where the reduction of airborne A 10.6 Uncontrolled, unlim-contaminants cannot be eliminated ited access efficiently by llVAC systems, breathing air B <1 Controlled, unlim-provisions are provided, for example, for ited access Amendment 10 12 3-5
ABWR = = . SantirdHant nix.n Zone 12.3.1.4 Implementation of AIARA Desig. Dose Rate n,1tha (mrem /ht) Ihstintien in this subsection, the implementation of design considerations to radioactive $3 stems for C <$ Controlled. limited maintaining personnel radiation exposures as low acceu,20 hr/wk as reasonably achievable is described for the following five systems: D < 25 Controlled. limited acct s,4 hr/wk (1) Reactor water cleanup system; E < 100 Controlled, limited (2) Residual heat removal system (shutdown access,1 hr/wk cooling mode); F > 100 Controlled access, (3) Fuel pool cooling and cleanup system; Authorization remired. (4) Main steam; nnd The dose rate applicable for a articular rone (5) Standby gas treatment system is based on operating experience and represents design dose rates in a particular zone, and 12.3.1.4.1 Reactor Water Cleanup Sptem should not be interpreted as the expected dose rates which would apply in all portions of that This system is designed to operate rone, or for all types of work within that zone, continuously to reduce reactor water radioactive
- or at all periods of entry into the zone. !.arge contamination. Components for this system are -
BWR plants have been in operution for two located outside the contalwent and include decades, and operating experience with similar filter demineralizers, a backwash receiving design basis numbers shows that only a small tank, regenerative and nonregenerative heat fraction of the 19CFR20 maximum permissible dose exchangers, pumps, and associated valves. is received in such zones from radiatior. sources controlled by equipment layout or the structural The highest radiation level components shiciding provided. Therefore, on a practical include the filter demineralize s, heat basis, a radiation roning approach as described exchangers, and backwash receiving tank. The above accomplishes the as low as reasonably filter demineralliers are locate well 1233.2.4 F adwaste llullding Access into the drywell is not permitted during normal operation. The ventilation system The radwaste building is divided into two inside merely circulates, without filtering, the zones for ventilation purposes. The control air. The only airflow out of the drywell into ac. room is one rone, and the remainder of the build-cessible areas is minor leakage through the wall. ing it, the other zone. The air pressure in the first zone is maintained slightly above atmo-During maintenance, the drywell air is spherie, while the air pressure in the second purged before access is allowed, zone is maintained slightly below atmospheric. Air in the second zone is drawn from outt.ide the 12JJ.23 Reactor flullding building and distributed to various work areas within the building. Air flows from the work The reactor building IIVAC system is divided areas and is then discharged via the reactor into three zones, which are separated by building stack. An alarm sounds in the control leaktight, phy+al barriers. The tones include: room if the exhaust fan fails. The exhaust now is monitored for radioactivity, and if a high ac-(1) secondary containment (this area contains livity level is detected, the potentially radio. equipment that is a potential source of ra- active cells are automatically isolated, but dioactivity and if a leak occurs, the other airflow through the work areas continues. accessible areas of the building are not con-taminated); if the exhaust now high radiation alarm con-tinies to annunciate after the tank and pump Amendment 12 11 M
AllWR wmi, . Stalidard I'lanL tu u rooms are isolated, the work area branch exhaust ducts are seicetively manually isolated to locate the involved building area. Should this technique fail, because the airborne radiation has spread throughout the building, the control room air conditioning continues, but the air con. ditioning for the balance of the building is shut down. The work area's exhaust air is drawn through a filter unit consisting of a particulate filter, a llEPA filter, a charcoal filter, and then another llEPA filter, before being discharged to the reactor building stack. The n' is mcnitored for radioactivity, and if a high level is de. tected, supply and exhaust is terminated, and the SGTS is started. Maintenance provisions for the filters are similar to those for the control building flVAC sptem. See Subsection 9.4.6 for a detailed discus. . sion of the radwaste building ilVAC system. . c Amendment 12.3-13.3
ABWR 2mmm Slandard Plant nvn 12.3A Area 1(adiation and Airborne 123..tJ Sperm Design Itadioactivity Monitors The area radiation monitoring detectors This section defines and describes the area provided in each plant building are listed in radiation system that monitors the gamma Tables 12.3 3 through 12.3-7 along with area radiation levels throughout the plant except location maps shown in Figures 12.3-56 through within the containment. The gamma radiation 12.3 73, Also, these tables specify the levels within the containment (drywell and sensitivity range of cach channel as designated suppression chamber) are monitored continuously below along with requirements for local area by the containment atmospherie monitoring system alarms. (CAhtS) as described in Subsection 7.6.2. 1 our gamma sensitive ion chambers (two per divisions 1 The channel sensitivity covers the following
& 2) are provided by CAhtS to monitor for airborne ranges:
radioactivity up to 107 rads per/hr. Those four sensors are located at the penetrations a) Range 10 2 to 102 mR/hr 11 listed in Table 6.2 8. The area radiation (iiigh Sensithity) monitoring system is classified as non. safety. b) Range 10-1 to 10 3 mR/hr hl 12J.4.1 System Objecthes (hiedium Sensitivity) The pt rpose of the area radiation monitoring c) Range 1 to 10 4 mR/hr L(Low system is to warn plant personnel of excessive Sensitisity) , gamma ray levels in service areas including the arcas where nuclear fuelis stored or handled, to d) Range 102 to 10 6 mR/hr LL(Low record and indicate the monitored gamma radiation Low Sensitivity) levels in the control room at selected locations within the various plant buildings, and to c) Range 10*l ta lt7 8 mR/hr VL provide audible local alarms at key locations (Very low Sensitivity) where abnormal radiation levels could endanger plant personnel. There are two radiation detectors that are located in the fuel storage and handling area, 12.3.4.2 System Description one is positioned to monitor the radiation near the fuel pool and the other is placed in the The area radiation monitoring system fuel handling area to monitor the radiation that consists of gamma sensitive detectors, associated may result from accidental fuel handling. digital radiation monitors, auxiliary units, Criticality detection monitors for this area are local audible warning devices and multipoint not needed to satisfy the criticality accident recorders. The actector signals are digitized requirements of 10CFR70.24, because the AI!WR and eptically multiplexed for transmission to the design utilizes specialized high density fuel radiation monitors. Each monitor has two storage racks that preclude the possibility of adjustable trip circuits for alarm initiation, criticality accident under normal and abnormal one high radiation level trip and one downscale conditions. The new fuel bundles are stored in trip. The downscale trip circuit operates on racks that are placed at the bottom of the fuel loss of power or when gross equipment failure storage pool. A full array of loaded fuel occurs. Auxiliary units are provided in local storage racks are designed to be suberitical by areas for radiation indication and for initiating at least 5% delta k. Refer to Sections 9.1 and the sonic alarms on abnormal levels. The 9.2 for details, electronics are powered from the non 1E vital 120 Vac source while the recorders are powered from The detectors and radiation monitors are the 120 Vac instrument bus, responsive to gamma radiation over an energy range of 80 kev 7 hieV. The energy dependence Amendment 18 12114
ABWR mwm - - SLtuidFrd Plant tu v n will not exceed 20% of point from 100eV to 3 hicV. The overall system design accuracy is within 9.5% of equivalent linear full scale recorder output for any decade. The trip alurm setpoints will be established in the field following equipment installation at the site. The exact settings will be based on sensor location, back ground radiatloa levels, expected radiation levels, and low occupational radiation exposures. linch channelis calibrated based on a pseudo input signal to confirm accurate monitor response. The detectors are calibrated using standardized traceable radioactive source in order to establish the linearity and sensitivity of the channel for subsequent calibration. The area radiation monitoring system is designed to accommodiate periodic surveillance testing. The area radiation monitoring instru- , mentation is designed and properly located to , provide early detection and warnleg for personne! protection to insure that occupational radiation exposures will be as low as is reasonably achieved (AIARA)in accordance with guidelines stipulated in Reg Guide 8.2 and 8.8. The area radiation monitoring system in. cludes instrutnentation provided to assess the radiation conditions in crucial areas in the reactor building (the RHR equipment areas) where access may be required to service the safety related equipment during post LOCA per Reg Guide 1.97. Amendment 18 12.3-14 1
DOM 2 w.tmat. Slandard Plant luy 9 12.3.5 Post Accident Access from the clean access corridor at the 4S00 lesci Requirements (BIF) and up three floors to the 23500 level (3F). There are two access corri dors, clean The locations requiring access to mitigate the and dirty, with contamination in those areas consequences of an accident during the 100-day limited to air inflitration from the environment post accident period are the control room, the and penetration leakage from the PASS system. technical support center, the remote shutdown in addition, the lines penetrating the PASS room panel, the primary containment sampic station are doubly valved permitting line isolation in (post accident sample system), the health physics the event of any potential rupture. Sources of facility (counting room), and the nitrogen gas radiation therefore are limited to minor leakage supply bottics. Each area has low post LOCA and gamma shine including the stack monitor room radiation levels. The dose evaluations in which contains only instrumentation and Subsection 15.6.5 are within regulatory associated penetrations for monitoring stack guidelines. e f flue nt. Access to vital areas through out the reactor 12.3.6 Post. Accident Radiation building / control building / turbine building Zone Maps complex is controlled via the service building. Entrance to the service building and access to The post accident radiation zone maps for the the other areas are controlled via double locked areas in the reactor building are presented in secured entry ways. Access to the reactor Figures 12.312 through 12.3 22. The zone maps building is via two specific routes, one for represent the maximum gamma dose rates that clean access and the second for controlled exist in these areas during the post-accident, access. During a event such as a design basis period. These dose rates do not include the . accident, the service building / control building airborne contribution in the reactor builJing. are maintained under filtered ilVAC at a positive pressure with respect to the environment. Air Post-accident tone maps of the control infiltration is minimited by positive flow via building and turbine building are presented in doubic entry ways. Therefore, radiation exposure Figures 12.3 54 and 55 respectively. The zone is limited to gamma shine from the reactor maps are designed to reflect the criteria building, turbine building, main steam line establis'.ed in Subsection 3.1.2.2.10. access corridor, and skyline. 'I his shine is mirimited by locating highly populated areas 12.3.7 Deleted below ground. During a design basis accident event, access 12.3.8 References to remote shutdown panel, nitrogen bottles, and the PASS and monitor systems is controlled from 1. N. M. Schaeffer, Reactor Shic/ ding for the service building via the controlled access Nuclear Engineers, TID 25951, U.S. Atomic way. These corridors are not mdntained under Energy Commission (1973). filtered positive pressure so that per.onal protection equipment (radiation protection suits, 2. J.11.11ubbell, Photon Cross Sections, breathing gear, etc.) will be required in the Attenuation Corfficients, and Encrgy access corridor. Primary contamination would Absorption Coefficients from 10 KcI' to 100 occur from leakage through the PASS system and GrV, NSRDS NHS20, U.S. Department of air infiltration from the environment. Both Commerce, August 1969. pathways are considered minimal and minor contamination under even the most adverse 3. Radi.) logical #calth Handbook, U.S. conditions is expected. Department of Ilealth, Education, and Welfare, Revised Edition, January 1970. The reactor building vital areas are all l located off one of of the two primary 4.access Reactorways Handbook, Volume 111, Part B, E.P. creept the nitrogen bottle areas which are Blinard, U.S. Atomic Energy Commission located on the refueling floor and are accessible (1962). Amendment 12 115
f ' ABWR zw,um. . Standard I'lant iam n
- 5. Lederer, lloilander, and Perlman, Table of Imfoprs, Sixth Editiota, (1968).
- 6. M.A. Capo, Polynomial Appro.tirnation of Garnma Ray Buildup Factors for a Point Imircpic Source, APEX 510, November 1958,
- 7. Reactor Phyics Constants, Second Edition, AhL 58CO, U.S. Atomic Energy Commisslor;,
July 1963.
- 8. ENDF/B Ill aad ENDF/U IV Cross Section Libraries, Brookhaven National Laboratt,ry.
- 9. PDS 31 Cross Section Library, Oak Ridge National Laboratory.
- 10. DLC 7, ENDF/B Phote '-* action Library.
l l' 1 Amendment 18 ggj. gig l
ABWR wim Standard Plant tuv n Tchte 12.31 COMPlTl'ER CODES USED IN SillEl, DING DESIGN CALCULATIONS Computer CodeDesceiption OADF A multigroup, multiregion, point kernal, gamma ray code for calculating the flux and dose rate at discrete locations within a complex source-geometry configuration. GGG A multigroup, multiregion, point kernal code for calculating the conti.bution due to garnma ray scattering in a betrogeneous three dimensional space DOT.4 A discrete ordinates, two dimensional transport code., Multigroup, multiregion neutron or gamma transport , 12.3 16 Amendment 10
ABWR mwat. . StandardflanL . pcv n Table 12.3 2 TYPICAL NICKEL AND COLilALT CONTENT OF MNTEltlALS Nickel Colball Alate .al G W Carbon Steel 0.25 1% of Ni Stainleu Steel 10 1% of Ni Ni Cr Fe (inconel600, 70 1% of Ni ineonel X750) Stellite 6 3 58 Amendment 10 3 37 L_
m _ - - t i L ABWR ~ 2mmu, ; Standard Plant
~
nn n
~l Table 12.3 3 AREA RADIATION MONITORS REACTOR HUILDING Sensitivity Local -
Eg, lxention & Descriotion Elgurtf, Bane dlaDlu 1 Reactor atca (A) 4F 123-62 11 X ! 2 Reactor area (B)-4F 123-62 LL 3 Fuel storage poc) atca (A) 4F 12342 LL X 4 _ Fuel storage pool ana (B)-4F 123-62 LL >
$ R/B 4F south area .12342 11 6 R/B 4F SE area - 123 62- li X ,
7 R/B 3F NW area 123-60 1I 8 R/B 3P SE area 123-60 11 X , 9 CUWintrol panel area B3F 123 M 11 10 ,R/B equipment hatch B2P 123 57 11 X 11 : IICU area (A) B3F . 123 56 - M X 12 IICU area (B) 83F 123-56 M X , 13 SRV/MSIV valve maintenance roorn 3F 123-63 M X 14 - R/B IF SE hatch area 123-49 11 X 15 RPV instrument tack room (A)-B1F 123 58 il X 16E RPV instrument tack room (B) B1F
~
123 58 11 X
'17. R/B BlF SE hatch area 123 58 11 18 TIP drive machine rooci El 1500 123 57. 'M X ,
-19 TIP machine equipment room El1500 123-57 ' L X 20 Core cooling water sampling room M4F 123-61 M X 21 CRD maintenance room B2F 123-57 M X 3 22 R/B B2F SE hatch area 123 57. , li _X 23 R/B B2P NW hatch area 123 57 .I1 X 24 R/B B3F NW arca RIIR 'A* cquip area '123 56 VL- X 25 R/B B3F SE area RIIR *B' equip area 123-56 VL X
-i
- Amendment 18 12.3-17.1
=
s.
+
ABWR mmu - - Siandard PJRHf Ev n Table 12.3 4 .j AREA RADINiiON MONITORS , CONTROL BUILDING i Sensithity b'n, Location & Description l' inure # Range 1- Main Control Room 12341 11 2 Passage Way Underneath Stearn Tunnel 123 41 11 3 RBCW"A' Area El.1315 123-64 Il 4 RDCW'B" Area El.1315 123 64 11 5- RBCW *C" Area.El 1315 123 64 11
. i Table 12.3 5 AREA RADIATION MONITORS SERVICE BUILDING '
Sensithlty N2 IA* cation & Description Elgure # Hanne 1 Service Building Tech. Support Center 123-64 11 Amendment 18 12.3-17.2' ---.-..,._;.-s ._,--.--,1..- ..--.._.__,a.--. _ _ _ . . _ . . . . . - - , , . _ , . _ - . -
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'ABWR :mmva.
i< rv. n Standantflant Table 12.3 6 AltEA RADIATION MONITORS RADWASTE !!UILDING Sensithity local Ng, lecution & Descrintion 1]Eutr.1 llangt Alarms 1 R/W Building Control Room El 16000 12 3-68 11 2 hiaintenance area #1 El 1(000 12346 11 X 3 - hiaintenance Area #2 El16000 12346 11 X 4 R/W Duilding HVAC Ediaust El 1600 123 68 11 5 R/W DuildingTruck Area El7300 123-67 il 6 htSW Compactor Area El7300 123 67 11 7 Corridor to Aux. Building El 7360 123 67 11 X t 8 Equip Rack Area #1 El 0200 12346 II 9 Equip Rock Area #2-El 0200 12344 11 10 R/W Euilding htSW Control Room El-0200 12344 li 11 Rad Waste Sampling Room El 6500 123 6$ 11 12 htSW Equipment Area-El-6500 123-65 11 X 13 R/W Equipment Rack Area #1 El-6500 123-65 11 14 R/W Equipment Rack Area #2 El 6500 123 65 II 123-173 Amendment 18
ABWR zwaam . S.llliliblrd Plarli i<ev ti Table 12.3 7 AltEA RADIATION MONITOltS TUltlllNE IlUILDING Senstthity Local h lic112LL& Descritition 1]gure No. BBKt alDRtu
- 1. Condensate Pump hiaintenance Area 123 70 h1
- 2. Condensate Sampling & Control Area 123 70 h1 X
- 3. Off Gas Sample & Control Area 12 3-70 hi X 4, RFP 1A.1B & 1C Area 123-70 Ii X
- 5. Filter hiaintenance Area 123 71 hi X
- 6. Demineralizer Area 123 71 11
- 7. SJAE A & Recombiner Area 123 71 11
- 8. SJAE B & Recombiner Area 123 71 11
- 9. IIP lleaters & Drain Tank Area 1 123-71 l{
10, llP lleaters & Drain Tank Aren 2 123 71 11
- 11. MSR 1A & IC Area 12 3-72 11 12, htSR 1B & 1D Area 123 72 11
- 13. Turbine Building Operating ihr 12.2 73 11 X
- 14. Equipment htain Access Area 123 73 11 X Amet.Jment IS 12 S 17 4
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APPENDIX 12A CALCULATION OF AIRBORNE RADIONUCLIDES 4 2 r
-ABWR ummat.
mb ' ' ; Standard Plant wo a; 12A.1 CALCULATION OF - R *.
= the kth removal constant for AIRBORNE RADIONUCLIDES the jth source and the ith radionuclide as discussed-
' This= appendix presents a simplified below. .
methodology to calculate the airborne concen-
- trations of radionuclides in a compartment. This = radiorn,clide decay constant methodology is conservative in nat'ure and
( , assumes that diffusion and mixing in a
~ '
Evaluation Parameters compartment it basically instantaneous with respect to those mitigating mechanisms such as The following paramer - u luire evaluation on a
- radioactive decay and other removal mechanisms. case by use basis dictate : S s physical parameters
' The following calculations need to be performed and processes germaine to N ocling process. -
on an isotope by isotope basis to verify airborne concentrations are within the limits of 10CFR20. (1) S is defined as the source rate for radionuclide i 18to the compartment. Typically these sources take (1); For the compartment, all sources of airborne the form of: radionuclides need to be identified such as: (a) Inflow of contaminated air from an upstream
.(a) Flow of contaminated air from other compartment Given the concentration of areas radionuclide i, c., in this air and a flow rate of
*r", the source rate then becomes S. - re ;
4
- (b) Gaseous releases from equipment in the companment (b) Production of airborne radionuclides from equipment. This typically takes two forms, (c) . Evolution of airborne sources from ~ gaseous leakage, and liquid leakage.
- sumps or water leaking from equipment '
(i) For gaseous leakage sources, the source rate is equal to the concentration of
~
(2) Second, the primary sinks of airborne r:dionuclides need to be identified. This will radionuclide i, c, and the leakage rate,"r", primarily be outflow from the compartment or S = tc.. d i but may also t.n the form of condensation onto room coolers. (ii) For liquid sources, the source rate is similar but more complex. Given a liquid (3) . Given the above information the following concentration c and a leakage rate, *r", the equation will calculate a conservative total release krom the leak is rc .= The concentration, fraction of this release which then becomes airborne is typically evaluated by a S.ij gd_ y i partition factor, P which may be conser-
- vately estimated from:
r ' (4 + g g) a
- Where: Noble Gases - P, = 1 C. -= Concentration of the ith' h
' h' radioauclides in the room - All others .P =
I h
- h' V = Volume of roou where: h= saturated liquid S; = The jth source (rate) of the enthalpy 4
ith radionuclide to the room. These sources are discussed saturated liquid h' = enthalpy at one below. atmosphere = 100.10 Kcal/Kg
- 12A.l.t As. .sdment r
5 m-+, , m-.i .,, , - . _ - - , , - - .- ,
ABWR wm. . Stand;ird Plant wn h = saturated vapar (2) The compartment contains a pump carrying reactor enthaipy at one coolant witp a maximum spegified leakage rate of atmosphere = 639.18 0.0000M m per hour at 273.6 C. Kcal/Kg (a) Conservatively it can be estimated bawd upon Therefore the liquid release rate properties from steam tables (see note 1) that becomes, re,P,. under these conditions 44% of the liquid will flash to steam and become airborne Along is defined as the removal rate with the flashing liquid it is assumed that a (2) co R@stant and typically consists of: proportional amount of I-131 will bece me (a) Exhaust rate from the compar. tment. This term considers not (b) Using the design basis iodine concentrations only the exhaust of any initially for rea: tor water from Table 11.1-2 of 0.016 a contaminated ak but also any clean pCi/gm of I-131, it is calculated that the air which ma, oed to dilute the purgp i.4 providing a source of I 131 of 5.0 x compartment air. 10' Ci/sce to the air. (see Note 2) (b) Compartment filter systems are Second, the sinks for airborne material need to be treated by the equation: identified. This example include oaly exhaust which is categorired as flow out of the computment at Rg., e (1 F,) ' r, 150% per hour or 4.2 x 10 per second.
# f where r. = filter system flow Therefore, for an equilibriun, situation, the I-1M rate airborne concentration from this liquid source would be calculated from the following equation.
F= filter efficiency i for radionuclide i A = S / ( A+ R ) + S,/( K+ R ), 2 where (c) Other removal factors on a case by S,= sourg sate in Curies per second = 5.0 case basis which may be deemed x 10 Ci/sec from liquid reasonable and conservative. S= source rate from inflow = 2.4 x 10 2 Example Calculation Ci/sec (Values used below are examples only and = isotope decay const9nt in units of per should not be used in any actual evaluation.) second = 9.977 x 10' /sec This example will look at 11}1 in a R, = R, = removal rate consjant per second compartment 6.1x6.1x7.6 = 282.80 m' = V (exfiltration) = 4.2 x 10 per second First all primary source of radionuclides A = 6.2 x 10 pCi/mi of I-131. needs to be identified and categorized. Notes: (?) Flow into the compartment equals 424.8m' per hour wth thejnput 1 131 concen- tration 1. The assumption of 44% flashing at 273.6 C is equal to 2 x 10' p Ci/ml ({r,om upstream extremely conservative, see Reference 1 for a compartments) or'2.4 x 10' Ci/sec. No discussion of fission product transport. other sources of air either contaminated or 3 clun air are assumed. 2. Water density assumed at 0.743 gy/cm based upon I standard tables for water at 273.6 C. Amendment 12Al-2
, . ABWR mww.
Standan! Plant n,x n 12A.2 References
- 1. Paquette, et al, Volatility of Fission Pre >dacts During Reactor Accidents, Journal of Nuclear Materials, Vol 130 Pg 129-138,1985, 5
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Figure 12.3-22 REACTOR BUILDING RADIATION ZO E MAP POST LOCA AT CROSS SECTION B.B ~ Amendment 15.3-39
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, Gencrtl l'3crir6c Qimpny ABWR nOPRIETARY INFORMATION w.t ooA t.
Standard Plant a.u in am n
- 1. RW/11 LCW SAMP TANK
- 2. RW/Il LCW SAMP PUMP
- 3. RW/Il llCW SAMP TANK
- 4. RW/!) IICW SAMP PUMP ;
- 5. RW/IlllSD SAMP TANK
- 6. RW/11 IISD SAMP PUMP
- 7. LCW COLLECTOR TANK
- 8. LCW SAMPLE TANK !
- 9. LCW COLLECTOR PUMP
- 10. LCW SAMPLE PUMP ,
- 11. LCW DEMINERALIZER ,
- 12. LCW PILTER
- 13. IICW COLLELTOR TANK ,
- 14. IICW RESIN SUPPLY TANK CilANGED TO IICW
- 15. IICW CAUSTIC TANK
- 16. IICW ACID TANK 17.; ilCW COLLECTOR PUMP
- 18. IICW CAUSTIC PUMP -
- 19. IICW ACID PUMP l
- 20. IICW DEMINERALIZER -
- 21. IICW DISTILLATIITANK ,
- 22. IICW DISTILLATE PUMP
- 23. IICW CONCENTRATOR RECIRCULATION PUMP -
- 24. IICW CONCENTRATOR ,
- 25. IlCW CONCENTRATOR llEATER .
- 26. IICW CONCENTRATOR CONDENSER-27.- IISD SAMPLE TANK .
- 28. IISD SAMPLE PUMP
- 29. IISD FILTER ,
- 30. CUW PilASE SEPARATOP e
- 31. SPENT RESIN TANK
- 32. SLURRY AGITATION PUMP !'
- 33. SLURRY PUMP 34, DECANT PUMP
- 35. CONW LIQUID WAST TANK '
- 36. CONW SEAL WATER TANK
- 37. CONW L10UID WAST PUMP
- 38. CONW SEAL WATER PUMP
- 39. SOL WAST SUPPLY TANK
.'40. SOL POWDER llOPPER
- 41. SOL BINDER llOPPER
- 42. SOL BINDER MEASURING IIGi PER
- 43, SOL SOLIDIFICATION AGENT SILO
- 44. SOL SOLIDIFICATION AGENT MEASURING llOPPER
- 45. SOL ADDITIVE WATER TANK 46, SOL MIXING TANK Figure 12.3 36 RADWASTE HUILDING, EQUIPMENT LIST (SiiEET 1 OF 2)
Amendment 12M i
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Generol l'leetnc Company ABWR PaOenicTrav isrOnurriOs 234sioont Standard PlDill cim III net _B
- 47. SOL CLEANING WATER TANK
- 49. SOL CLEANING WATER RECEIVER TANK
- 49. SOL WAST SUPPLY TANK REClRCUIATION PUMP
- 50. SOL WAST SUPPLY PUMP
- 51. SOL VENT BLOWER
- 52. SOL F1LTER BLOWER
- 53. SOL CLEANING WATER RECEIVER PUMP
$4. SOL DECANT PUMP
- 55. SOL DRYER
- 56. SOL MIST SEPARATOR
- 57. SOL CONDENSER 58 SOL PELLETIZER
- 59. SOL PELLET 11LLING MAC111NE
- 60. SOL PARTICl.E FILTER
- 61. SOL llEPA FILTER
- 62. SOL DRUM CONVEYOR
- 63. SOL SIIIELD DOOR
- 64. SOL PELLETIZER CONTROL UNIT
- 65. SOL CAPPING MACillNE
- 66. SOL AIR llEATER
- 67. SOL SOLIDIFICATION AGENT PARTICLE FILTER
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- 68. MSW WAST OIL TANK
- 69. MSW WAST OIL RECEIVE PUMP
- 70. MSV WAST OIL FEED PUMP
- 71. MSW COMBUSTION AIR BLOWR
- 72. MSW OFF GAS BLOWER
- 73. MSW/ UXILIARY EXIIAUST GAS BLOWER
- 74. MSW INCINERATOR
- 75. MSW PRIMARY CERAMIC FILTER
- 76. MSW SECONDARY CERAMIC FILTER
- 77. MSW URY ACTIVE WAST SilOOTER
- 78. MSW AIR PRE-11 EATER 79 MSW INCINERATOR GLOVE 00X
- 80. MSW CERAMIC FILTER GLOVE BOX
- 81. MSW ASil DISCilARGING CONVEYOR
- 82. MSW ASil Dl%CIIARGING EQUIPMENT
- 83. MSW RELIEl GAS FILTER
- 84. MSW llEPA FILTER
- 85. MSW AIR MIXER
- 86. MSW CERAMIC FILTER BURNER
- 87. MSW WAST OIL BURhER 88, MSW INCINERATOR ASII DISCIIARGING BOX
- 89. MSW CERAMIC FILTER BACKDLOW EQUIPMENT
- 90. MSW UOX PULLET STORAGE SYSTEM
- 91. MSW SUPER COMPACTOR
- 92. MSW COMPACTOR
- 93. TANK VENT FILTER
- 94. IIVAC SUPPLY
- 95. liVAC EXilAUST Figure 12.3 36 RADWASTE IlUILDING, EQUIPMENT LIST (SilEET 2 of 2)
Amendment 12 M O li
l t A 5 0.6 mremh s < t0 mrem = 50> a C < 5.0 mrem /hr 1 2 3 4 5 6 sm i 3 R 5$ ; D < 25 mreme E <100 mrem /hr
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E Figure 12.3-39 RADWASTE BUILDING, RADIATION ZONE MAP, NORMAL OPERATION 9 4 AT ELEVATION 7300mm
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N CH HATCH l I l ELEV. 23000 o y .. . . _ _ - _ . - - ;;g C _ __ _E__ C-F C F ax nainrt...ct SToaAce > F- ' 72 73 83 39 ' TA ARtA '$ (({y, }6QQQ F"""""" sE* a C-F C a C eg i _ rp , nAtwEwa'Nct anta 52 STA O 24 z G2s 2s 29 20 ELEV. 7300 Y ' E
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