ML20128P852
| ML20128P852 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 06/30/1985 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19304B384 | List:
|
| References | |
| NUDOCS 8507260521 | |
| Download: ML20128P852 (120) | |
Text
_
12.0 RADIATION PROTECTION 12.1 ENSURINGTHATOCCUPATIONALRADIA110NEXPOSURESAREASLOWASIS REASONABLY ACHIEVABLE (ALARA)
O Quantitative standards of radiation protection are established and maintained by the International Commission on Radiation Protection (ICRP), the National Connittee on Radiation Protection (NCRP), and the former Federal Radiation Council (FRC) which is now part of the Environmental Protection Agency (EPA).
The recommendations of these bodies are reflected in 10 CFR Part 20,
' Standards for Protection Against Radiation."
Specifically, this regulation establishes maximum allowable levels for occupational radiation exposure, in addition to complying with the exposure limits, the regulation specifies that every reasonable ef fort should be made to maintain exposures as low as is reasonably achievable (ALARA).
12.1.1 Policy Considerations O
Refer to the applicant's safety analysis report for a discussion of the applicant's management policy and organizational structure related to ensuring that occupational radiation exposures are ALARA.
The applicant should provide information describing the implementation of
- policy, organization,
- training, and design review guidance provided,in Regulatory Guides 1.0, 0.8, and 8.10 and the training requirements in 10 CFR 19.12.
O Administrative programs and procedures, in conjunction with facility design, should ensure that occupational radiation exposures, both individual and collectively, will be kept below regulatory limits and as low as reasonably achievable as required by 10 CFR 20.1(c).
The applicant's policy should also include a " Radiation Protection Plan" as defined in Section 5 of NUREG 0161.
8507260521 B50712 PDR ADOCK 05000601 K
pon WAPWR-RP 12.1-1 JUNE, 1985 2997e:1d
Westinghousa is committed to helping ensure that occupational radiation exposures are ALARA in pressurized water reactors by providing s" stems and components whose designs take into consideration the radiation exposures
~
associated with operation, inspection, and maintenance.
In keeping with this commitment, Westinghouse has defined the responsibilities of the radiation protection groups and has provided an environment in which the radiation protection functions can' properly perform their duties.
While Westinghouse does not have responsibility for policy considerations related to the operation of systems and components designed and supplied by Westinghouse, a conni cment has been made to gather operational information related to radiation protection aspects of Westinghouse systems and components.
This operational information is then utilized by the radiation protection staff in working with the designers thereby assuring that operational aspects related to the ALARA philosophy are considered during the design stage.
This information is utilized in the development of Westinghouse recommended operation, maintenance, and inspection procedures for the Westinghouse equipment to reflect ALARA practice.
O The policy considerations outlined above ensure that occupational radiation exposures are ALARA in compliance with Regulatory Guides 8.8, Revision 3,
P "Information Relevant to Ensuring that Occupational Radiation Exposures a't Nuclear Power Stations will be As Low As is Reasonably Achievable," and 8.10, Revision 1,
" Operating Philosophy for Maintaining Occupational Radiation x
Exposures As low As Is Reasonably Achievable," and 10 CFR Part 20 (Section 20.l(c)).
12.1.2 Design Considerations The basic philosophy embodied in Westinghouse pressurized water reactor design considerations to ensure that occupational radiation exposures are ALARA can 1
I be expressed as:
i!APWR-RP 12.1-2
.1UNE, 1985 2997e:ld I
1.
Design of systems and components to ensure increased reliability and maintainability, thereby effectively reducing the maintenance requirements on radioactive components.
2.
Design of systems and components to reduce the radiatioIn fields to ensure that operation, maintenance, and inspection activities are performed in the minimum radiation field feasible.
3.
Design of systems and components to reduce the time spent in radiation fields during operation, maintenance, and inspection.
4.
Design of systems and components to accommodate remote and semi-remote operation, maintenance, and inspection procedures.
For the WAPWR, Westinghouse has incorporated many design improvements aimed at reducing operational radiation exposure, including:
1.
Simpler / faster refueling operation via integrated head package (these features are discussed further in Section 12.3).
O 2.
Corrosion resistant steam generators.
9-3.
Steam generator maintenance features.
4.
18-24 month fuel cycles.
5.
Increased CVCS filtered flow.
1 6.
Shielding / layout improvements.
7.
Simplification of major fluid systems (e.g., reduced valve count).
i 8.
Reduction in CVCS safety class (allows much less inspection 4 less exposure).
WAPWR-RP 12.1-3 JUNE, 1985 J
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In translating this design philosophy into practice, Westinghouse calls upon experience f rom past designs operating in the field and upon other relevant field experience as well as laboratory tests.
Many diverse sources of I
relevant field experience and data are used in implementing this design philosophy including:
NRC publications, the Atomic Industrial Forum (AIF)
NESP studies. Electric Power Research Institute studies, internal Westinghouse
- programs, and personal comunications with plant operatcrs.
Internal Westinghouse programs, which involve measuring and recording exposure and radiation level data in operating plants during operation, maintenance, and inspection, as well as personal comunications with operating plant staffs, have been an invaluable source of feedback from operating plants.
Important information is also obtained from Radiation Exposure Management (REM) l seminars.
During 1984 Westinghouse hosted its sixth annual REM seminar for utility health physics and radiation personnel with representation from nearly all of the Westinghouse reactor sites in the United States and attendees from l
several foreign countries.
The information from all of these sources is assimilated and evaluated by the staff responsible for radiation protection l
functions.
The feedback and radiation data from operating plants is used to construct models to predict occupational radiation exposure patterns for various operations, maintenance, and inspection activities on systems and p.
components of the Westinghouse scope of supply.
These models and exposure patterns are further described in Section 12.4.
From these models, the potential for improvements in areas' such as reliability, repair time, and operational techniques related to occupational radiation exposures can be identified for further study.
Recommended design practice and design considerations are comunicated to the system and component designers in three ways within Westinghouse Water Reactor Divisions.
1.
Consultation and personal comunication.
2.
ALARA training programs 3.
Design reviews.
O WAPWR-RP 12.1-4 JUNE, 1985 2997e:ld
The first of these comunications techniques is an informal process employing open comunication and sound engineering judgment.
The second comunication method (ALARA training programs) is a formal training session administered by the cognizant radiation protection personnel.
The program is given to engineers within the various Westinghouse divisions whose responsibility includes design of systems and components associated with nuclear plants.
The aim of the program is to provide design engineers with criteria, design features, operational g'uidelines, and operating plant experience relevant to radiation protection and the minimizing of occupational radiation exposures.
Thus the designer has an awareness of the field conditions and problems imposed by a radiation environment on the operation, maintenance, and inspection of systems and components.
The final communication method (design reviews) is the final check by the radiation protection specialists of the system and component designs to ensure that occupational radiation exposures will be ALARA.
These design reviews are conducted coincident with the safety design review on systems and components required by 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."
O As an aid to the designer in ensuring that ALARA features have been addressed, P
the use of an ALARA design checklist is encouraged. The recomended checklis't includes various ALARA considerations grouped by the basic approaches to exposure reduction, such as time, distance, shielding, and source reduction.
Many of the inherent design features and policy considerations to ensure that occupational radiation exposures will be ALARA throughout the operating life are also applicable during the eventual decomissioning of the plant.
These features include equipment design for ease of accessibility anit maintenance, provisions for remote flushing of equipment, ability to use re.1ote handling equipment, decontamination techniques, and component design features to minimize crud buildup.
Specifications and limitations on cobalt content in equipment components will serve to limit radiation doses from crud buildup during both operation and subsequent decomissioning.
Westinghouse has been active in studies by AIF and NRC to define the impact of decomissioning by O
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. LUNE, 1985 2997e:1d
providing relevant information to the principle investigators.
Westinghouse will continue to be aware of decommissioning impacts in the design of systems and components.
The design considerations outlined above ensure that occupational radiation exposures will be maintained ALARA in accordance with Section C.2 of Regulatory Guide 8.8 and in compliance with 10 CFR Part 20 (Section 20.1(c)).
12.1.3 Operational Considerations Refer to the applicant's safety analysis report for a discussion of the applicant's operational plans and procedures for ensuring that occupational radiation exposures are ALARA.
While Westinghouse does not have ultimate responsibility for policy considerations related to the operation of systems and components designed and supplied by Westinghouse, recomendations and limitations on the operation of the systems and components are supplied to the utility applicant.
These recomendations and limitations are developed taking into account the ALARA philosophy.
The field experience feedback mechanism described in Section t-12.1.2 is also used to develop recomended operation, maintenance, and inspection procedures for systems and components described in RESAR-SP/90 which may contain radioactive materials. These operational considerations are also factored into system and component designs as described in Section 12.1.2.
The operational considerations outlined above ensure that occupational radiation exposures are ALARA in compliance with Regulatory Guides 8.8 and 8.10 and 10 CFR Part 20 (Section 20.1(c)).
O O
WAPWR-RP 12.1-6 JUNE,1985 2997e:1d
i 12.2 RADIATION SOURCES This section discusses and identifies the sources of radiation,that form the basis for shield design calculations and the sources of airborne radioactivity j
used for the design of personnel protection measures and dose assessment.
12.2.1 Contained Sources The shielding design source terms are based upon the three plant conditions of normal full power operation, shutdown, and design basis accident events.
12.2.1.1 Sources for Full Power Operation The primary sources of radioactivity during normal full power operation are direct core radiation, coolant activation processes, leakage of fission products from pinhole defects in fuel rod cladding, and activation of reactor coolant corrosion products.
The design basis for the shielding source terms for fission products in this section is cladding defects in fuel rods producing 0.25 percent of the core thennal power.
The design basis for activation and corrosion product activities is derived from measurements at operating plants and is independent of fuel defect level.
The radionuclide activity levels in the reactor coolant at the design basis level are given in Section 11.1 of RESAR-SP/90 PDA Module 12
" Waste Management", as are the models and assumptions used in determining these sources.
Westinghouse provides, to the applicant, numerous reactor radiation source values for the at-power condition, including:
1.
Neutron particle fluxes at the inside surface of the primary shield concrete at the core midplane.
2.
Gamma ray energy fluxes at the inside surface of the primary shield concrete at the core midplane.
WAPWR-RP 12.2-1 JUNE, 1985 2962e:1d
3.
Gansna ray dose rates at the inside surface of the primary shield concrete.
O 4.
Detailed angular distributions of radiation leakage (neutron and gansna ray) from the reactor pressure for streaming analyses.
The nitrogen-16 activity of the coolant (produced f rom oxygen activation) is the controlling radiation source in the design of the secondary shield and is tabulated in Table 12.2-1, in microcuries per gram of coolant, as a function of transport time in a reactor coolant loop.
The nitrogen-16 source in the pressurizer is given in Table 12.2-2.
Fission and corrosion product activities circulating in the reactor coolant and out-of-core crud deposits comprise the remaining significant radiation sources during full power operation.
The fission and corrosion product activities circulating in the reactor coolant are given in Section 11.1 of RESAR-SP/90 PDA Module 12
" Waste Management".
The fission and corrosion product source strengths in the reactor coolant pressurizer liquid and vapor (s
phases are given in Table 12.2-2.
The isotopic composition and specific
\\
activity of typical out-of-core crud deposits are given in Table 12.2-3.
Typically, 1 milligram of deposited crud material is found in one square centimeter of a relatively smooth surface.
This may be as much as 50 times higher in crud trap areas.
Crud trap areas are generally locations of high turbulence, areas of high momentum change, gravitational sedimentation areas, high-af finity-material areas, and possibly thin-boundary-layer regions.
Systems which process or contain reactor coolant also contain radiation sources during full power operation.
These systems include the chemical and w
volume control system (CVCS) and the boron recycle system (BRS), as described in RESAR-SP/90 PDA Module 13
" Auxiliary Systems".
Table 12.2-4 gives the radiation sources for the CVCS, specifically delineating the sources for:
Ad 1.
CVCS letdown stream.
2.
Mixed bed demineralizers.
3.
Cation bed demineralizer.
WAPWR-RP 12.2-2 JUNE, 1985 2962e:ld
4.
Volume control tank (liquid and vapor phases).
5.
Reactor coolant, seal water injection, seal water return, boric acid, and boric acid polishing filters.
6.
Regenerative,
- letdown, seal injection, and excess 1.etdown heat exchangers.
Radiation sources in the CVCS consist of radionuclides carried in the reactor coolant.
The design of the CVCS ensures that most of the N-16 has decayed before the letdown stream leaves the containment by placing a delay mechanism in the letdown flowpath.
All CVCS system heat exchangers other than the seal O
injection heat exchanger are located in the reactor containment building.
The shielding design is based on the maximum activity in each component:
A.
CVCS Ion Exchangers The mixed bed demineralizer is in continuous use and removes fission products in cation and anion form.
It also is highly effective in removing corrosion products.
The cation bed demineralizer is used intermittently to remove lithium for pH control.
It also is highly effective in removing the monovalent cations, cesium and rubidium.
I The short-lived isotopes are assumed to build up to saturation activities on both beds.
For the long-lived isotopes, the activity i
retained is assumed to 'be evenly distributed between the two demineralizers.
B.
Volume Control Tank The radiation sources in the volume control tank are based on a nominal operating level in the tank of (a,c)
_ in the vapor phase and on the stripping f ractions given in Table 11.1-1 of RESAR-SP/90 PDA Module 12
" Waste Management",
assuming no volume control tank purge.
l l
l WAPWR-RP 12.2-3 JUNE, 1985 2962e:1d I
L
i C.
CVCS Filters OO The design criterion for CVCS filter shielding is based primarily on operating experience.
The source strengths for the reactor coolant filter correspond to an exposure rate of 500 rem /hr at contact; the source strengths for the remaining filters correspond to an exposure rate of 100 rem /hr at contact.
These dose rates are arrived at assuming the filters are homogeneous sources with the dimensions and composition given in Table 12.2-4.
D.
CVCS Heat Exchangers The regenerative, letdown, and excess letdown heat exchangers are located in the containment building.
They provide the cooling for the reactor coolant letdown. Their radiation sources include N-16.
The magnitude of the N-16 source strength is highly sensitive to the location of these heat exchangers with respect to the RCS loop piping.
Therefore, the N-16 source strengths fo'r these heat exchangers are based on the average value in the crossover leg between the steam generator and the reactor coolant pump.
The shielding design takes into account the N-16 decay from the crossover leg to each heat exchanger.
Ta'ble 12.2-5 gives the radiation sources for the BRS, specifically delineating the sources for:
1.
Recycle evaporator feed and condensate demineralizers.
2.
Recycle holdup tanks (liquid and vapor phases).
O 3.
Recycle evaporator feed and condensate filters.
4.
Recycle evaporator vent condenser vapor.
5.
Recycle evaporator concentrates.
O l
WAPWR-RP 12.2-4
. LUNE, 1985 2962e:ld
The shielding design is based on the maximum activity in each component:
A.
BRS Ion Exchangers The recycle evaporator feed demineralizers are located upstream of the holdup tanks and contain mixed-bed resins that remove anion and cation activity f rom the reactor coolant entering the holdup tanks, along with corrosion products not retained by the CVCS mixed bed demineralizers.
The recycle evaporator condensate demineralizer is charged with anion resin to remove any boron and iodine that may be carried over with the evaporator condensate.
B.
Recycle Holdup Tank The recycle holdup tanks are each equipped with a diaphragm.
Gases that flash from the reactor coolant letdown to the holdup tanks are retained under the diaphragm until about 600 cubic feet of gas has accumulated.
The gases are then removed to the gaseous waste system.
The radiation sources in the holdup tanks are based on the unit letting down to a single holdup tank at the maximum letdown flow rate, assuming 50 percent of the gases flash into the vapor phase.
The liquid phase is assumed to contain reactor coolant that has been processed through the CVCS mixed bed and recycle evaporator feed demineralizers.
C.
BRS Filters The recycle evaporator feed filter and condensate filter are located donwstream of their respective demineralizers.
They retain particulates and any resin fines which may escape from the demineralizers.
WAPWR-RP 12.2-5 JUNE, 1985 2962e:ld i
r-,-
The source strengths for the feed filter correspond to a radiation level of 100 rem per hour at contact with the filter housing.
The C
condensate filter source strengths result in radiation levels that are less than I rem per hour at contact with the filter housing...
The filters are assumed to be drained of process fluid and are considered to be homogeneous sources with the dimensions and compositions shown in Table 12.2-5.
D.
Recycle Evaporator The maximum gaseous activity in the evaporator occurs while processing reactor coolant.
The gases are concentrated in the vent condenser portion of the evap' orator.
The source strength of the evaporator concentrates is based on an evaporator process limit of 40 microcuries of long-lived radionuclides per gram of concentrates.
The nitrogen-16 activity is not a factor in the radiation sources for systems and components located outside containment such as the CVCS and BRS due to its short 7.11 second half-life and the greater than 1 minute transport time criteria.
Other auxiliary systems containing radiation sources which require shielding include the. liquid waste processing system (LWPS),
the gaseous waste processing system (GWPS), the solid waste processing system, the spent fuel pit cooling system (SFPCS), and the steam generator blowdown processing system (SGBPS).
The radiation sources for the LWPS are presented in Table 12.2-6.
The liquid waste processing system (LWPS) is considered as several subsystems, based on its intended use during normal operation.
The equipment items normally associated with processing reactor-grade water are the reactor MAPWR-RP 12.2-6
. LUNE, 1985 2962e:1d
coolant drain tank, waste holdup tank, waste evaporator feed filter, and waste evaporator.
The evaporate distillate is directed to the waste condensate tank and may be further processed through the waste evaporator condensate demineralizer and filter.
The waste evaporator concentrates are sent to the drumming room for packaging.
v l
Low activity, nonreactor-grade water is directed to the floor drain or laundry l
and hot shower subsystems.
Normally this water is analyzed, then discharged.
If activity levels prevent this, the water can be processed by a demineralizer and filter, or the waste evaporator.
The equipment included in the subsystem is the floor drain tank and filte.', laundry and hot shower tank and filter, waste monitor tank demineralizer and filter, and two waste monitor tanks. The source strengths for the floor drain tank filter are the same as those for the waste evaporator feed filter, since both filters perform similar functions.
The source strengths for the waste monitor tank demineralizer and filter are based on processing reactor coolant through these components.
Radioactive spent resins discharged from the various demineralizers are retained in the spent resin storage tank.
The short-lived radionuclides are allowed to decay before the resin is directed to the druming station for packaging.
The associated equipment includes the spent resin storage tank and 9-the resin sluice pump and filter.
Radiation sources in the various pumps in this system are assumed to be identical to the liquid sources in the tank from which the pump takes sucti,on.
Radiation sources in the laundry and hot shower tank and filter, and in the waste condensate tank are negligible; these items do not require shielding.
The waste evaporator condensate filter and waste monitor tank filter are located downstream of their respective demineralizers.
The source strengths for the condensate filter result in radiation levels of less than 5 rem per hour at contact with the filter housing.
The maximum radiation level for the waste evaporator feed, spent resin sluice, floor drain tank, and waste monitor tank filters is 100 rem per hour at contact with the filter housing.
O WAPWR-RP 12.2-7 JUNE,1985 2962e:1d
i s
The filters are assumed to be drained of process fluid and are considered to be homogeneous sources with the dimensions and compositions shown in Table 12.2-6.
4 The source strength of the evaporator concentrates is based on an evaporator process limit of 40 microcuries of long-lived radionuclides per gram of concentrates.
For shielding design purposes, the concentrates activity should be assumed in the recirculation pump, concentrates heater, flash tank, concentrates cooler and interconnecting piping.
The radiation sources for the GWPS are presented in Table 12.2-7.
Radiation sources for each component of the GWPS are based on operation with the maximum activity conditions as given in Subsections 11.1 and 11.3 of RESAR-SP/90 PDA Module 12
" Waste Management".
The major equipment items in the GWPS are the refrigerated waste gas dryer, charcoal guard and adsorption beds, gas surge tank, and gas compressor.
The radioactive gases removed from the reactor coolant system at the volume O
control tank are directed througn the refrigerated waste gas dryer, charcoal guard bed and a series of six charcoal adsorption beds.
The effluent is then t-normally discharged via the plant vent although provisions are made for 4
recirculating back to the volume control tank. The gamma ray source strengths for this equipment are derived from plant operation during which the radioactive gases are stripped from the reactor coolant system.
i The gansna ray source strengths for the gas compressor and gas surge tank are based on the maximum sources associated with components which can be routed to the surge tank, i.e., the evaporators, RCS drain tank, pressurizer relief tank and BRS holdup tank.
The radiation sources for the solid waste processing system are presented in Table 12.2-8.
.!O MAPWR-RP 12.2-8 JUNE,1985
)
2962e:1d
}
l The spent resin and evaporator concentrates are packaged at the solid waste drumming station for shipment to a burial or long-term storage f acility, which is generally located offsite.
Prior to shipment, the packaged waste is stored in a drum storage area.
G The initial gamma ray source strength in the spent resin storage tank is assumed to be the same as that in the CVCS mixed-bed demineralizer.
After a 30-day decay period, only the cesium and cobalt isotopes are significant contributors.
The source strength for the evaporator concentrates is based on an evaporator process limit of 40 microcuries of long-lived radionuclides per gram of concentrates.
The initial source strength is based on degassed reactor coolant, less short-lived radionuclides, concentrated by a factor that will yield 40 microcuries per gram at evaporator shutdown.
Table 12.2-8 includes the specific gamma ray source strengths, by energy group, at the time of processing and following a 30 day decay period.
The radiation sources for the SFPCS are presented in Table 12.2-9.
The spent fuel pit cooling system is capable of accomplishing simultaneous cleanup of both spent fuel pit and refueling water.
The major equipment items considered for shielding design in this system are the demineralizers and filters.
The spent fuel pit demineralizer is charged with 75 cubic feet of mixed-bed resin and is used to remove particulate radionuclides that may be present in the spent fuel pit or refueling water.
The specific gama ray source strengths for the demineralizer are based on purification of a refueling cavity with a water volume of 500,000 gallons.
The spent fuel pit filters are located downstream of the spent fuel pit demineralizer and serve to retain particulates and any resin fines which may escape from the demineralizers.
O MAPWR-RP 12.2-9 JUNE, 1985 2962e:1d
The filter source strengths correspond to an exposure rate of 100 rem per hour at contact with the filter housing.
The filter is assumed to be drained of process fluid and is considered to be a homogeneous source with the dimensions and compositions shown in Table 12.2-9.
~
The radiation sources for the SGBPS are presented in Table 12.2-10.
J Although only those portions of the SGBPS which are safety-related are within the scope of the WAPWR NPB and the non-safety-related portions which process O'
the blowdown are the responsibility of the plant specific applicant (see Section 10.4.8 of RESAR-SP/90 PDA Module 6/8), source strengths for a typical Westinghouse system are presented here for shielding design purposes.
The steam generator blowdown processing system maintains the water effluent from the steam generators at a chemical and radiological specification suitable for its recycle into the main condenser or for its discharge.
The major equipment items considered for shielding design in this system are the demineralizers, the SGBPS spent resin storage tank, and the filters.
O O
The blowdown demineralizers contain mixed-bed resins that remove anion and cation activity from the blowdown fluid.
p Spent resins from the blowdown demineralizers are transferred to the spent resin storage tank for temporary storage prior to drumming.
The spent resin sluice pump takes suction from the liquid contained in the spent resin storage tank.
The blowdown prefilter is located upstream of the blowdown demineralizers and is provided to prevent plugging of the resin beds. The blowdown outlet filter is located downstream of the blowdown demineralizer and serves to retain particulates and any resin fines which may escape f rom the demineralizers.
The spent resin sluice filter serves to retain any resin fines contained in Os the sluice water.
d MAPWR-RP 12.2-10 JUNE, 1985 2962e:1d
The filters are assumed to be drained of process fluid and are considered to n
be homogeneous sources with the dimensions and compositions shown in Table U
12-2.10.
12.2.1.2 Sources for Shutdown in the reactor shutdown condition, the only significant sources requiring 1
permanent shielding consideration are the spent reactor fuel, the residual heat removal system, and the incore detector system.
Individual components may require shielding during shutdown due to deposited crud material.
Estimates of accumulated crud are given in Subsection 12.2.1.1 while dose rates may be obtained from Subsection 12.4.1.
The radiation sources in the reactor coolant system, CVCS, and BRS are bounded by the sources given for full power operation with the exception of a short time period (i.e., less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) following shutdown during which the fission product spiking phenomena and crud bursts can result in increased radiation sources.
The spiking phenomena involves the release of a portion of the accumulated water soluble salts from the interior cladding surface (e.g., iodine and cesium) and gases (e.g., xenon and krypton) of defected fuel rorts during the shutdown and s.,/
coolant depressurization (Reference 1).
Crud bursts are the resuspension or solubilization of a portion of the accumulated deposited corrosion products 9-into the reactor coolant system during shutdown such as during oxygenation of the reactor coolant.
However, special shielding considerations to accommodate these increases should be unnecessary due to several factors including:
1.
The spike or crud burst release is of short duration (generally less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).
2.
The CVCS purification loop (i.e.,
letdown through the demineralizer and filters) is generally in operation at full capability during the shutdown (a,c)
The maximum gamma ray source strengths in the residual heat removal system loop are given in Table 12.2-11 for 4 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> af ter reactor shutdown. The residual heat removal system is placed into operation at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a WAPWR-RP 12.2-11
. LUNE, 1985 2962e:ld
shutdown at the maximum shutdown rate.
The system removes decay heat from the AQ reactor for the duration of the shutdown.
The sources given are maximum values with credit for 4 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of fission and corrosion product decay and purification.
Core average gamma ray source strengths are given in Table 12.2-12.
These source strengths are used in the evaluation of radiation levels within and around the shutdown reactor.
These sources are based on a three region core with the regions operated at 435, 870 and 1306 ef fective f ull power days, respectively.
Spent fuel gamma ray source strengths are given in Table 12.2-13.
These source strengths are used in the evaluation of radiation levels for spent fuel handling, storage, and shipping.
These sources are based on one core region operated for 1306 ef fective full power days.
All of these sources may be put on a per unit volume of homogenized core basis by multiplying by the power density of 86.6 watts per cubic centimeter for the reactor described in RESAR-SP/90.
Core average and spent fuel neutron source strengths are given in Table 12.2-14.
The only source material, byproduct material, or special nuclear material requiring shielding consideration for the Westinghouse nuclear power block described in the RESAR-SP/90 is the photoneutron source material used in the secondary source rods.
The gamma ray and neutron source strengths for the secondary source rods, irradiated for 400 days, are given in Table 12.2-15.
Irradiated hybrid incore detector / thermocouple and drive cable maximum gamma ray source strengths are given in Table 12.2-16.
These source strengths are used in determining shielding requirements when detectors are being moved during or following a flux mapping of the reactor core.
These source strengths are given for a detector irradiation period of 30 days and a drive cable irradiation period of 400 days and are given in terms of per cubic centimeter of detector and drive cable.
Irradiated incore detector, drive cable average gamma ray source strengths are given in Table 12.2-17.
These source strengths are used in determining shielding requirements when the detectors are not in use and for shipment when the detectors have failed. The values are given in terms of per cubic centimeter of drive cable af ter an HAPWR-RP 12.2-12 JUNE, 1985 2962e:1d
irradiation period of 400 days.
Irradiated incore flux thimble gama ray source strengths are given in Table 12.2-18.
These source strengths are used in determining shielding requirements during refueling operations when the
~
flux thimbles are withdrawn from the reactor core.
The values are given in terms of per cubic centimeter of Inconel-600 for an irradiation period of 15 years.
The flux thimbles are made of Inconel-600 with a maximum cobalt impurity content of 0.10 weight percent.
12.2.1.3 Sources for Design Basis Accident OV The radiation sources of importance for the design basis accident are the containment source and the residual heat removal system source.
The fission product radiation sources considered to be released from the fuel to the containment following a maximum credible accident are based on the assumptions given in TID-14844 (Reference 2).
These assumptions are consistent with those provided in Regulatory Guide 1.4 and Section II.B.2 of NUREG-0737.
The integrated gamma ray and beta particle source strengths for various time periods following the postulated accident are given in Table 12.2-19.
The post-accident recirculation system and shielding should be designed t'o P
allow limited access to the high head safety injection (HHSI) and the residual heat removal pumps following a maximum credible accident.
The sources are based on the assumptions in TID-14844 with only the nongaseous activity being retained in the sump water, which flows in the residual heat removal lo'op.
Noble gases formed by the decay of halogens in the sump water are assumed to
)
be released to the containment and not retained in the water.
Gamma ray source strengths for radiation sources circulating in the residual heat removal loop and associated equipment are given in Table 12.2-20.
Isotopic fission product sources from the maximum credible accident, based on O
the assumptions in TID-14844, are given in Chapter 15 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System".
O WAPWR-RP 12.2-13 JUNE, 1985 2962e:1d
4 12.2.2 Airborne Radioactive Material Sources 4
O
. O Sources of airborne radioactive material in equipment cubicles, _ corridors, or operating areas normally occupied by operating personnel f rom ~ systems and components described in RESAR-SP/90 may be obtained f rom the reactor coolant O
activities given in Section 11.1 of RESAR-SP/90 PDA Module 12,
" Waste Management".
l Sources resulting from the removal of the reactor vessel head and the movement O
of spent fuel are dependent on a number of operating characteristics (e.g.,
V coolant chemistry, fuel performance, etc.) and operating procedures followed during and af ter shutdown.
The permissible coolant activity levels following depressurization should be based on the noble gases evolved from the reactor coolant system water upon removal of the reactor vessel head.
The endpoint limit for coolant cleanup and degasification should be established based on maximum permissible concentration considerations and containment ventilation system capabilities of the plant.
Operating plant experience has indicated d
that coolant xenon-133 concentrations of less than 0.05 microcuries per gram have posed no problem to the containment atmosphere during vessel head removal.
The exposure rates at the surface of the reactor cavity and spent fuel pool water are dependent on the - purification capabilities of the reactor vessel cavity and spent fuel pool cleanup systems.
A water activity level of less than 0.005 microcuries per gram for the dominant gamma emitting isotopes at the time of refueling has been shown in operating experience to maintain the dose rate at the water surf ace to less than 2.5 millirem per hour.
I 12.2.2.1 Model for Calculating Airborne Concentrations For those regions which are characterized by a constant leakrate of the l
radioactive source at constant source strength and a constant exhaust rate of i
4 O
WAPWR-RP 12.2-14 JUNE, 1985 2962e:1d i
_-..__._._._-,_,-__mm
the contaminant, the peak or equilibrium airborne concentration of the radioisotope in the regions can be calculated, using the following equation:
-A t
1 (LR)$ g (PF)q (1-e Ti )
A C (t) =
g yxTi where j
(LR)$
Leak or evaporation rate of the i
radioisotope in
=
I gm/sec, in the applicable region and th g
activity concentration of the i leaking or evaporating j
A
=
j radioisotope in pCi/gm g,artition factor or the fraction of the leaking activity (PF)g
=
th that is airborne for the i radioisotope O
th total removal rate constant for the i radioisotope in A
=
Ti
-I sec from the applicable region (Kdi + le)
=
-I Adi + Ke are the removal rate constants in sec due th to radioactive decay for the i radioisotope and the exhaust from the applicable region time interval between the start of the leak and the time at t
=
which the concentration is evaluated in seconds free volume of the region in which the leak occurs in cm V
=
MAPWR-RP 12.2-15 JUNE,1985 2962e:1d
TABLE 12.2-2 (Sheet 1 of 2)
O RADIATION SOURCES - PRESSURIZER me 3
Licuid Phase (1500 f t )
O Energy Group Specific Source Strength (Mev/aamma)
(Mev/am-sec)
I"I 0.2 - 0.4 9.5 x 10 O.4 - 0.9 1.6 x 10 4
0.9 - 1.35 6.8 x 10 1.35 - 1.8 3.8 x 10*
4 1.8 - 2.2 3.5 x 10 4
2.2 - 2.6 3.4 x 10 3
2.6 - 3.0 4.7 x 10 3
3.0 - 4.0 1.9 x 10 2
4.0 - 5.0 2.2 x 10 O
3 Liauid Phase (1000 ft )
Energy Group Specific Source Strength 3
(Mev/aamma)
(Mev/cm _,,c) a, 5 (a) 0.2 - 0.4 1.6 x 10 4
0.4 - 0.9 1.1 x 10 3
l 0.9 - 1.35 2.9 x 10 1.35 - 1.8 8.0 x 10 4
1.8 - 2.2 1.3 x 10 2.2 - 2.6 2.8 x 10 2
/
2.6 - 3.0 1.9 x 10 3.0 - 4.0 1.5 x 10 0
WAPWR-RP 12.2-18 JUNE,1985 2962e:1d
TABLE 12.2-2 (Sheet 2 of 2)
RADIATION SOURCES - PRESSURIZER Nitrogen-16 Sources SDecific Source Strenath Discrete Energy Liquid Phase Vapor Phase 3
(Mev/aama)
(Mev/am-sec)
(Mev/cm -sec)
I 0
1.75 4.3 x 10 1.2 x 10 2
I 2.74 3.9 x 10 1.1 x 10 4
3 6.13 7.9 x 10 2.2 x 10 7.12 6.7 x 10 1.8 x 10 i
O V
O l
l O
l_
(a)
Includes 80 key xenon-133.
r l
MAPWR-RP 12.2-19 JUNE,1985 l
2962e:1d
l-1 TABLE 12.2-3 O
ISOTOPIC COMPOSITION AND SPECIFIC ACTIVITY OF i
TYPICAL OUT-OF-CORE CRUD DEPOSITS 1
i Activity (microcuries per milligram) of Deposited Crud
,)
for Effective Full Power Years of Plant ODeration Composition (Nuclide) 1 Year 2 Years 5 Years 10 Years Mn-54 1.0 1.1 1.3 1.4 Fe-59 0.5 0.5 0.5 0.5 Co-58 12.0 12.0 12.0 12.0 Co-60 1.5 2.3 4.0 6.0 In addition to corrosion products, about 1.0 microgram of mixed actinides and fission products may be present for'.each 1 gram of deposited crud.
' O O
' O WAPWR-RP 12.2-20 JUNE, 1985 2962e:1d I
,,,-,-,.___,,___...,-,.-.,,,_.,_m
_.-.._,_____m,...
TABLE 12.2-4 (Sheet 1 of 6)
RADIATION SOURCES - CHEMICAL AND VOLUME CONTROL SYSTEM Letdown Coolant Sources (*
Energy Group Specific Source Strength (Mev/aama)
(Mev/am-sec)
O 4 (b) 0.2 - 0.4 9.5 x 10 0.4 - 0.9 1.6 x 10' 4
O.9 - 1.35 6.8 x 10 1.35 - 1.8 3.8 x 10 4
1.8 - 2.2 3.5 x 10 2.2 - 2.6 3.4 x 10 3
2.6 - 3.0 4.7 x 10 3
3.0 - 4.0 1.9 x 10 2
4.0 - 5.0 2.2 x 10 V
Mixed Bed Demineralizer 3
(75'ft of Resin)
Energy Group Specific Source Strength (Mev/aama)
(Mev/cm -sec) 0.2 - 0.4 2.2 x 10 8
0.4 - 0.9 2.0 x 10 0.9 - 1.35 2.3 x 10 6
1.35 - 1.8 5.8 x 10 1.8 - 2.2 1.7 x 10 2.2 - 2.6 1.0 x 10 2.6 - 3.0 2.3 x 10*
3 3.0 - 4.0 6.7 x 10 2
4.0 - 5.0 7.8 x 10 MAPWR-RP 12.2-21 JUNE, 1985 2962e:1d
i 1-TABLE 12.2-4 (Sheet 2 of 6)
RADIATION SOURCES - CHEMICAL AND VOLUME CONTROL SYSTEM Cation Bed Demineralizer (75 ft of Resin)
's Energy Group Specific Source Strength (Mev/aama)
(Mev/cm -sec) 4 0.2 - 0.4 1.3 x 10 8
0.4 - 0.9 1.8 x 10 i
6 0.9 - 1.35 8.3 x 10 6
1.35 - 1.8 4.3 x 10 1.8 - 2.2 6.5 x 10" I
2.2 - 2.6 2.2 x 10*
4 2.6 - 3.0 2.2 x 10 3
3.0 - 4.0 5.2 x 10 2
4.0 - 5.0 7.5 x 10 Y
i t
! O
.i 1
I l
i O WAPWR-RP 12.2-22 JUNE,1985 l
2962e:1d 1
l l
TABLE 12.2-4 (Sheet 3 of 6)
RADIATION SOURCES - CHEMICAL AND VOLUME CONTROL SYSTEM Volume Control Tank (c)
O VaDor Phase Energy Group Specific Source Strength 3
(Mev/camma)
(Mev/cm -sec) 6 (b) 0.2 - 0.4 1.2 x 10 j
0.4 - 0.9 8.3 x 10 4
0.9 - 1.35 1.6 x 10 1.35 - 1.8 4.8 x 10 4
1.8 - 2.2 8.4 x 10 5
2.2 - 2.6 1.8 x 10 3
2.6 - 3.0 1.0 x 10 3.0 - 4.0 5.9 x 10 i
- ~
Liauid Phase Energy Group Specific Source Strength (Mev/aamma)
(Mev/am-sec) 4 (b) 0.2 - 0.4 7.7 x 10 0.4 - 0.9 8.7 x 10 4
0.9 - 1.35 3.9 x 10 1.35 - 1.8 1.5 x 10 1.8 - 2.2 2.0 x 10*
4 2.2 - 2.6 1.1 x 10 3
2.6 - 3.0 4.3 x 10 3.0 - 4.0 1.3 x 10 2
4.0 - 5.0 1.8 x 10 MAPWR-RP 12.2-D JUNE,1985 2962e:1d
N
,1-
\\.
i g
).
TABLE 12.2-4 (Sheet 4 of 6)
..JK v.,
.s y, ;3
?-
.s
' " '~ \\
3 RADIATION SOURCES - CHEMICAL c
AND VOLUME CONTROL SYSTEM s4.
s
~ (
Reactor Coolant Filter Energy Group Specific Source Strength s
i (Mev/camma)
(Mev/cm -sec)
F4 s..
s.,
\\
5.7 x 107 l
c,'
O.4 - 0.9 0.9 - 1.35 1.5 x 107 as
\\_ w.
I) 1111 Water Injection Ftiter
,t Energy Group Specific Sodrce Strength 3
(Mev/nanwa)
^
(Mev/cm,,,c) l l
O.4 - 0.9 1.4 x 108 0.9 - 1.35 2.8 x 107 Id) l Seal Water Return Filter Ener'gy Group Specific Source Strength 3
e-(Mev/ cama)
(Mev/cm -sec)
.s d'
O 4 - 0.9 1.1 x 107 0.9 - 1.35 3.0 x 106 Boric Acid Filter I
Boric Acid Polishina Filter Energy Group Specific Source Strength (Mev/cm -sec)
(Mev/aanna) 3 i
0.2 - 0.4 1.5 x 105 0.4 - 0.9 8.9 x 105 4
0.9 - 1.35 3.6 x 105 1.35 - 1.8 4.6 Y 104 1.8 - 2.2 3.3 x 103 2.0 x 103 2.2 - 2.6 s.
i s
t 12.2-24 I.
JUNE, 1985 WAPWR-RP
.2962e:1d i
\\ ' '. ~
s0
TABLE 12.2-4 (Sheet 5 of 6)
RADIATION SOURCES - CHEMICAL AND VOLUME CONTROL SYSTEM Regenerative. Letdown. and Excess Letdown Heat Exchanaers Specific Source Strength (Mev/am-sec)
Regenerative (Shell Side),
l O
Energy Group Letdown (Tube Side), and I
(Mev/aansna)
Excess Letdown (Tube Side)
Regenerative (Tube Side) 0.2 - 0.4 9.5 x 104 (b) 7.7 x 104 (b) 0.4 - 0.9 1.6 x 105 8.7 x 104 0.9 - 1.35 6.8 x 104 3.9 x 104 1.35 - 1.8 4.3 x 104 1.5 x 104 1.8 - 2.2 3.5 x 104 2.0 x 104 2.2 - 2.6 3.4 x 104 1.1 x 104 2.6 - 3.0 5.6 x 104 4.3 x 103 3.0 - 4.0 1.9 x 103 1.3 x 103 4.0 - 5.0 2.2 x 102 1.8 x 102 6.0 - 7.0 1.0 x 107 7.0 - 7.5 8.7 x 106 Seal Injection Heat Exchanaer (Tube Side) 5-Energy Group Specific Source Strength (Mev/aammd)
(Mev/am-sec) 0.2 - 0.4 9.5 x 104 (b) 0.4 - 0.9 1.6 x 105 0.9 - 1.35 6.8 x 104 I
1.35 - 1.8 3.8 x 104 1.8 - 2.2 3.5 x 104 O
2.2 - 2.6 3.4 x 104 V
2.6 - 3.0 4.7 x 103 3.0 - 4.0 1.9 x 103 4.0 - 5.0 2.2 x 102 O
O WAPWR-RP 12.2-25 JUNE, 1985 2962e:1d
/
TABLE 12.2-4 (Sheet 6 of 6)
RADIATION SOURCES - CHEMICAL AND'YOLUME CONTROL SYSTEM
+
Note :
(a) The letdown coolant volume is plant layout dependent.
(b) Includes 80 key xenon-13.
(c) These sources correspond to'a nominal operating level in the tank of 360 ft in the vapor phase and 240 ft in the. liquid phase.
(d) Homogeneous sources with the following dimensions and compositions:
Source Dimensions Inches Source Composition Filter Radius Length (Volume Percent) l
- ~
Reactor coolant 3.375 19 67% air, 33% water Seal water return 3.375 19 67% air, 33% water l
Seal water injection zl.375 21 11% air, 89% water Boric acid 3.375 19 67% air, 33% water Boric acid polishing 3.375 19 67% air, 33% water 1
O O
l O
WAPWR-RP 12.2-26 JUNE, 1985 2962e:1d
TABLE 12.2-5 (Sheet 1 of 5)
RADIATION SOURCES - BORON RECYCLE SYSTEM Recycle EvaDorator Feed Demineralizer (75 ft of Resin)
Energy Group Specific Source Strength (Mev/aamma)
(Mev/cm#-sec) 0 0.2 - 0.4 8.2 x 10 0.4 - 0.9 3.1 x 10 6
0.9 - 1.35 4.0 x 10 1.35 - 1.8 7.5 x 10 3
1.8 - 2.2 5.7 x 10 2.2 - 2.6 3.6 x 10 l
Reevele EvaDorator Condensate Demineralizer 3
(30 ft of Resin)
P Energy Group Specific Source Strength 3
(Mev/aama)
(Mev/cm -sec) 4 0.2 - 0.4 1.6 x 10 3
0.4 - 0.9 9.8 x 10 3
0.9 - 1.35 2.0 x 10 2
1.35 - 1.8 9.9 x 10 I
1.8 - 2.2 8.5 x 10 2.2 - 2.6 5.8 x 10
' O WAPWR-RP 12.2-27 JUNE, 1985 2962e:1d
TABLE 12.2-5 (Sheet 2 of 5)
O RADIATION SOURCES - BORON RECYCLE SYSTEM Recycle HolduD Tanks VaDor Phase 3
(600 ft )
Energy Group Specific Source Strength O
3 (Mev/aamma)
(Mev/cm,3,c) 0.2 - 0.4 1.4 x 105 (a) i 0.4 - 0.9 2.0 x 104 0.9 - 1.35 5.7 x 103 1.35 - 1.8 1.8 x 104 1.8 - 2.2 2.8 x 104 2.2 - 2.6 5.4 x 104 2.6 - 3.0 6.2 x 102 3.0 - 4.0 8.8 x 102 4.0 - 5.0 5.5 x 101 Liouid Phase p.
(100.000 cal)
Energy Group Specific Source Strength (Mev/aamma)
(Mov/am-sec) 0.2 - 0.4 8.2 x 104 (a) 0.4 - 0.9 2.7 x 104 0.9 - 1.35 9.7 x 103 i
1.35 - 1.8 1.2 x 104 1.8 - 2.2 1.8 x 104 2.2 - 2.6 3.0 x 104 6-2.6 - 3.0 7.8 x 102 3.0 - 4.0 6.3 x 102 4.0 - 5.0 4.9 x 10I O
WAPWR-RP 12.2-28 JUNE,1985 2962e:1d
TABLE 12.2-5 (Sheet 3 of 5)
O RADIATION SOURCES - BORON RECYCLE SYSTEM ID)
Recycle Evaporator Feed Filter O
' Energy Group Specific Source Strength 3
(Mev/aanna)
(Mev/cm _3,c)
I 0.4 - 0.9 1.1 x 10 0.9 - 1.35 3.0 x 10 i
I)
Recycle Evaporator Condensate Filter Energy Group Specific Source Strength 3
(Mev/camma)
(Mev/cm -sec) 4 0.2 - 0.4 3.0 x 10 0.4 - 0.9 2.0 x 10 3
0.9 - 1.35 4.0 x 10 P
3 1.35 - 1.8 1.9 x 10 2
1.7 x 10 1.8 - 2.2 2.2 - 2.6 1.1 x 10
. O O
j l
l-h!APWR-RP 12.2-29 JUNE,1985 2962e:1d i
~ TABLE 12.2-5 (Sheet 4 of 5)
O RADIATION SOURCES - BORON RECYCLE SYSTEM Recycle EvaDorator O
Vent Condenser Vapor Energy Group Specific Source Strength (Mev/aama)
(Mev/cm -sec) 6 (a) 0.2 - 0.4 2.6 x 10 5
0.4 - 0.9 3.7 x 10 5
0.9 - 1.35 1.0 x 10 0
1.35 - 1.8 3.2 x 10 1.8 - 2.2 5.2 x 10 5
2.2 - 2.6 9.8 x 10 4
2.6 - 3.0 1.1 x 10 4
3.0 - 4.0 1.6 x 10 4.0 - 5.0 1.0 x 10 9-
~
Evaporator Concentrates Energy Group Specific Source Strength (Mev/aamma)
(Mev/am-sec) 5 0.2 - 0.4 1.5 x 10 0.4 - 0.9 8.9 x 10 0.9 - 1.35 3.6 x 10 4
1.35 - 1.8 4.6 x 10 3
1.8 - 2.0 3.3 x 10 3
2.0 - 2.6 2.0 x 10 O
WAPWR-RP 12.2-30 JUNE, 1985 2962e:1d
i i
TABLE 12.2-5 (Sheet 5 of 5)
O RADIATION SOURCES - BORON RECYCLE SYSTEM Notes:
O (a)
Includes 80 key xenon-133.
(b) Homogeneous sources with the following dimensions and compositions:
O Source Dimensions i
inches Source Composition Filter Radius Lenath (Volume Percent)
Recycle evaporator feed 3.375 19 67% air, 33% water j
l Recycle evaporator 3.375 19 67% air, 33% water condensate O
p.
!O O
O WAPWR-RP 12.2-31 JUNE, 1985 2962e:1d
TABLE 12.2-6 (Sheet 1 of 6)
RADIATION SOURCES - LIQUID WASTE PROCESSING SYSTEM --
~
Waste Evaporator Condensate Demineralizer Source Strenaths (for 30 cubic feet of resin)
Energy Group Specific Source Strength (Mev/aama)
(Mev/cm -sec)
O 4
0.2 - 0.4 7.5 x 10 5
0.4 - 0.9 1.7 x 10 4
0.9 - 1.35 4.6 x 10 3
1.35 - 1.8 7.0 x 10 1.8 - 2.2 4.2 x 10 2.2 - 2.6 2.7 x 10 0
2.6 - 3.0 9.0 x 10 5
TOTAL 3.0 x 10 O
g..
'daste Monitor Tank Demineralizer Source Strenaths (for 30 cubic feet of resin)
Energy Group Specific Source Strength 3
(Mev/camma)
(Mev/cm _,,t) 5 0.2 - 0.4 5.8 x 10 6
0.4 - 0.9 3.9 x 10 0.9 - 1.35 1.6 x 10 5
1.35 - 1.8 2.3 x 10 4
1.8 - 2.2 2.3 x 10 2.2 - 2.6 1.0 x 10 TOTAL 6.4 x 10 WAPWR-RP 12.2-32 JUNE,1985 2962e:1d i
l I-
TABLE 12.2-6 (Sheet 2 of 6)
RADIATION SOURCES - LIQUID WASTE PROCESSING SYSTEM..
Reactor Coolant Drain Tank Liouid Phase Source Strenaths (for 175 aallons of liauid)
O-Energy Group Specific Source Strength (Mev/aama)
(Mev/am-sec)
O-I 0.2 - 0.4 9.5 x 104 (a) 0.4 - 0.9 1.6 x 105 0.9 - 1.35 6.8 x 104 1.35 - 1.8 3.8 x 104 1.8 - 2.2 3.5 x 104 2.2 - 2.6 3.4 x 104 2.6 - 3.0 4.7 x 103 3.0 - 4.0 1.9 x 103 4.0 - 5.0 2.2 x 102 TOTAL 4.4 x 105 Reactor Coolant Drain Tank VaDor Phase Source Strenaths (For 23.4 cubic feet of vaDor) i Energy Group Specific Source Strength 3
(Nv/aama)
(Mev/cm,3,c) 0.2 - 0.4 4.1 x 105 (a) 0.4 - 0.9 4.5 x 103 0.9 - 1.35 6.7 x 102
~
1.35 - 1.8 2.0 x 103 1.8 - 2.2 3.5 x 103 2.2 - 2.6 7.4 x 103 2.6 - 3.0 4.1 x 101 3.0 - 4.0 2.1 x 101 s
TOTAL 4.3 x 105 (a)
Includes 80 Kev Xe-133 MAPWR-RP 12.2-33 JUNE, 1985 2962e:1d
TABLE 12.2-6 (Sheet 3 of 6)
RADIATION SOURCES - LIQUID WASTE PROCESSING SYSTEM..
Floor Drain Tank. Waste Monitor Tanks. and
~
Waste HolduD Tank Source Strenaths O
Energy Group Specific Source Strength (Mev/aama)
(Mev/am-sec) 4 0.2 - 0.4 1.6 x 10 0.4 - 0.9 1.4 x 10 4
0.9 - 1.35 5.8 x 10 1.35 - 1.8 1.7 x 10 3
1.8 - 2.2 2.6 x 10 2.2 - 2.6 7.8 x 10 I
2.6 - 3.0 3.0 x 10 5
TOTAL 2.3 x 10 O
Chemical Drain Tank Source Strenaths 9-Energy Group Specific Source Strength (Mev/aamma)
(Mev/am-sec) 4
- 0. 4 - (,. 9 2.4 x 10 2
0.9 - 1.35 6.2 x 10 1.35 - 1.8 6.2 x 10 4
TOTAL 2.8 x 10 O
t O-WAPWR-RP 12.2-34
. LUNE, 1985 2962e:1d
TABLE 12.2-6 (Sheet 4 of 6)
C\\
RADIATION SOURCES - LIQUID WASTE PROCESSING SYSTEM
~;
SDent Resin Storace Tank Source Strenoths O
Energy Group Source Strenath (Mev/aram-sec)
(Mev/aamma)
Resin liquid C')
0.2 - 0.4 2.2 x 107 2.2 x 103 O.4 - 0.9 2.0 x 108 2.0 x 104 0.9 - 1.35 2.3 x 107 2.3 x 103 1.35 - 1.8 5.8 x 106 5.8 x 102 1.8 - 2.2 1.7 x 105 1.7 x 101 2.2 - 2.6 1.0 x 105 1.0 x 101 2.6'- 3.0 2.3 x 104 2.3 x 100 3.0 - 4.0 6.7 x 103 6.7 x 10-1 4.0 - 5.0 7.8 x 102 7.8 x 10-2 TOTAL 2.5 x 108 2.5 x 104 Waste EvaDorator Feed. Floor Orain Tank, and Waste Monitor Tank Filter Source Strenaths Energy Group Specific Source Strength 3
(Mev/aama)
(Mev/cm -sec) 0.4 - 0.9 1.1 x 107 0.9 - 1.35 3.0 x 106 TOTAL 1.4 x 107 SDent Resin Sluice Filter Source Strenaths l O l
Energy Group Specific Source Strength 3
i (Mev/aama)
(Mev/cm _3,c) 0.4 - 0.9 1.1 x 107 O.9 - 1.35 3.0 x 106 l
TOTAL 1.4 x 107 l
l WAPWR-RP 12.2-35 JUNE,1985 2962e:1d
TABLE 12.2-6 (Sheet 5 of 6)
RADIATION SOURCES - LIQUID WASTE PROCESSING SYSTEM
~1 Waste Evaporator Condensate Filter Source Strenaths
\\
Energy Group Specific Source Strength 3
(Mev/aamma)
(Mev/cm -sec) 5 0.2 - 0.4 2.7 x 10 O
5 0.4 - 0.9 5.3 x 10 5
0.9 - 1.35 1.8 x 10 4
1.35 - 1.8 2.4 x 10 1.8 - 2.2 1.7 x 10 3
2.2 - 2.6 1.0 x 10 6
TOTAL 1.0 x 10 Dimensions and Composition of Licuid Waste Processina System Filters O
i Source Dimensions Source Composition Filter Inches (Volume Percent)
Spent resin sluice Radius = 3.375 Air - 67%
Length = 19.
Water - 33%
Waste evaporator feed Same Same Waste evaporator condensate Same Same Floor drain tank Same Same i
Waste monitor tank Same Same O
WAPWR-RP 12.2-36 JUNE, 1985 l
2962e:1d
~
TABLE 12.2-6 (Sheet 6 of 6)
RADIATION SOURCES - LIQUID WASTE PROCESSING SYSTEM'-
Waste Evaporator Gas SDace Source Strenaths Energy Group Source Strength (Mev/aamma)
(Mev/cm -sec)
O 0.2 - 0.4 negl.
0.4 - 0.9 negl.
0.9 - 1.35 negl.
1.35 - 1.8 negl.
1.8 - 2.2 negl.
2.2 - 2.6 negl.
2.6 - 3.0 negl.
3.0 - 4.0 negl.
TOTAL negl.
Waste Evaporator Concentrates Source Strenaths t-Energy Group Source Strength (Mev/camma)
(Mev/am-sec) 5 0.2 - 0.4 1.5 x 10 0.4 - 0.9 8.9 x 10 5
0.9 - 1.35 3.6 x 10 1.35 - 1.8 4.6 x 10 1.8 - 2.2 3.3 x 10 2.2 - 2.6 2.0 x 10 6
TOTAL 1.5 x 10 O
WAPWR-RP 12.2-37 JUNE, 1985 2962e:1d
1 TA8LE 12.2-7 (Sheet 1 of 2)
RADIATION SOURCES - GASEOUS WASTE PROCESSING SYSTEM..
O Gaseous Waste Processing System Source Strengths 3
for Chiller. Gas Surae Tank (300 ft ). and Compressor Enerav Group (Mev/aamma)
Source Strenath (Mev/cm -sec)
- O 4
Gas Surge Tank Chiller and ComDressor 5 (a) 6 (a)
O.2 - 0.4 4.2 x 10 2.6 x 10 0.4 - 0.9 7.6 x 10" 3.7 x 10 4
5 0.9 - 1.35 1.4 x 10 1.0 x 10 1.35 - 1.8 4.2 x 10*
3.2 x 10 4
5 1.8 - 2.2 7.3 x 10 5.2 x 10 2.2 - 2.6 1.6 x 10 9.8 x 10' 0
2 4
2.6 - 3.0 9.1 x 10 1.1 x 10 2
3.0 - 4.0 5.4 x 10 1.6 x 10 3
4.0 - 5.0 1.0 x 10 6
TOTAL 7.8 x 10 5.0 x 10 (a) Includes 80 key Xe-133.
O O
O I
WAPWD-RP 12.2-38 JUNE,1985 2962e:1d
TABLE 12.2-7 (Sheet 2 of 2)
RADIATION SOURCES - GASEOUS WASTE PROCESSING SYSTEM Gaseous Waste Processing System Source Strengths for Charcoal Guard and Adsorber Beds 3
3 (for 75 ft Guard 8ed and 150 ft Adsorber Beds)
Enerav GrouD (Mev/gama)
Source Strength (Mev/Cm -sec)
Guard 8ed Adsorber Beds 6 (a) 6 (a) 0.2 - 0.4 5.0 x 10 3.7 x 10 4
4 0.4 - 0.9 6.3 x 10 3.4 x 10 3
3 0.9 - 1.35 9.3 x 10 5.6 x 10 1.35 - 1.8 2.8 x 10 1.7 x 10 4
4 1.8 - 2.2 4.9 x 10 3.0 x 10 4
2.2 - 2.6 1.0 x 10 6.3 x 10 2
2 2.6 - 3.0 5.2 x 10 3.1 x 10 I
3.0 - 4.0 1.9 x 10 9.8 x 10 6
6 TOTAL 5.3 x 10 3.8 x 10 Y
Gaseous Waste Processina System Source Strenaths For HEPA Filter Energy Group Source Strength 3
(Mev/aamma)
(Mev/cm -sec)
O 6 (a) 0.2 - 0.4 1.2 x 10 2
0.4 - 0.9 8.1 x 10 0
0.9 - 1.35 7.9 x 10 1.35 - 1.8 1.7 x 10 1.8 - 2.2 2.1 x 10 2.2 - 2.6 1.2 x 10 6
TOTAL 1.2 x 10 O
WAPWR-RP 12.2-39 JUNE,1985 2962e:1d
l TABLE 12.2-8 O
RADIATION SOURCES - SOLIO WASTE PROCESSING SYSTEM
- Solid Waste Source Strenaths at Time of Processina O
~
Spent Resin Evaporator Concentrates Energy Group Source Strength Source Strength 3
(Mev/ca ma)
(Mev/c;n -sec)
(Mev/aram-sec) 5 0.2 - 0.4 2.2 x 10 1.5 x 10 8
0.4 - 0.9 2.0 x 10 8.9 x 10 5
0.9 - 1.35 2.3 x 10 3.6 x 10 6
4 1.35 - 1.8 5.8 x 10 4.6 x 10 5
3 1.8 - 2.2 1.7 x 10 3.3 x 10 3
2.2 - 2.6 1.0 x 10 2.0 x 10 4
2.6 - 3.0 2.3 x 10 3
3.0 - 4.0 6.7 x 10 2
4.0 - 5.0 7.8 x 10 TOTAL 2.5 x 10 1.5 x 10 9-Solid Waste Source Strenaths Followina 30 Days Decay Spent Resin Evaporator Concentrates Energy Group Source Strength Source Strength -
3 l
(Mev/aama)
(Mev/cm -sec)
(Mev/aram-sec) 6 4
O.2 - 0.4 2.0 x 10 1.8 x 10 0.4 - 0.9 1.8 x 10 5.1 x 10 7
4 0.9 - 1.35 1.6 x 10 6.9 x 10 1.35 - 1.8 4.3 x 10 1.2 x 10*
O 8
5 TOTAL 2.0 x 10 6.1 x 10 O
MAPWR-RP 12.2-40 JUNE, 1985 2962e:1d
,-,-,,----g-
,,,,,_--,,..m
,,--.-------,---------vr-
--,----n-
-, e w---,---am--
w--
me---
---me
- - -o-e-o
--w e
TABLE 12.2-9 RADIATION SOURCES - SPENT FUEL PIT COOLING SYSTEM Spent Fuel Pit Demineralizer Source Strenaths (For 75 cubic feet of resin)
Energy Group Source Strength 3
(Mev/aama)
(Mev/cm -sec)
O 0.2 - 0.4 1.0 x 10" 5
0.4 - 0.9 6.5 x 10 1
5 0.9 - 1.35 3.0 x 10 3
1.35 - 1.8 5.9 x 10 Spent Fuel Pit Filter Source Strenaths Energy Group Source Strength 3
(Mev/aamma)
(Mev/cm -sec)
V 7
O.4 - 0.9 1.1 x 10 6
0.9 - 1.35 3.0 x 10 Spent Fuel Pit Filter Dimensions and Composition Source Dimensions Source Composition Filter (inches)
(volume Dercent)
Spent fuel pit Radius = 3.375 Air - 67%
i Length = 19 Water - 33%
i I
- O WAPWR-RP 12.2-41 JUNE, 1985 2962e
- 1d
,+-_--_,._,___---._.,_..-w
, - - _,. ~, _,.,, _ =,, -,.,.,.,. -.
.,m,,,.., - _ _
m..-._--,_,
_m.__._
i j
i TABLE 12.2-10 (Sheet 1 of 3)
O RADIATION SOURCES - STEAM GENERATOR BLOWDOWN PROCESSING SYSTEM Steam Generator 810wdown Demineralizer Source Strenaths (For 75 cubic feet of resin)
Energy Group Source Strength 3
(Mev/aama)
(Mev/cm,3,c) 5 0.2 - 0.4 2.9 x 10
~
0.4 - 0.9 3.9 x 10 6
0.9 - 1.35 2.2 x 10 1.35 - 1.8 9.6 x 10 3
j 1.8 - 2.2 1.7 x 10 2.2 - 2.6 5.4 x 10 O
Steam Generator Blowdown Spent Resin Storace Tank Source Strenaths Energy Group Source Strenath (Mev/aram-sec)
(Mev/aama)
Resin Liouid 5
I 0.2 - 0.4 2.9 x 10 2.9 x 10 3
0.4 - 0.9 3.9 x 10 3.9 x 10 6
2 O.9 - 1.35 2.2 x 10 2.2 x 10 5
I 1.35 - 1.8 9.6 x 10 9.6 x 10
-I 1.8 - 2.2 1.7 x 10 1.7 x 10 2
-2 2.2 - 2.6 5.4 x 10 5.4 x 10 O
O I
MAPWR-RP 12.2-42 JUNE,1985 2962e:1d
-....-,..-.,,._-.,.,,,.__,,,.,...._,.,__._n,,_,_,,,_,.
-.,,__.__,,,,.,________n_
_n,
...n,,,
TA8LE 12.2-10 (Sheet 2 of 3)
RADIATION SOURCES - STEAM GENERATOR BLOWDOWN PROCESSING S.YSTEM l
Steam Generator 810wdown Prefilter Source Strenaths Energy Group Source Strength (Mev/aamma)
(Mev/cm -sec)
I 0.2 - 0.4 9.9 x 10 3
0.4 - 0.9 1.6 x 10 2
0.9 - 1.35 6.5 x 10 I
1.35 - 1.8 9.6 x 10 1.8 - 2.2 1.3 x 10 0
2.2 - 2.6 5.8 x 10 0
2.6 - 3.0 2.8 x 10 l
Steam Generator 810wdown Outlet and Spent Resin Sluice Filter Specific Activity Energy Group Source Strength
~
(Mev/cansna)
(Mev/cm -sec) 5 0.2 - 0.4 2.9 x 10 7
0.4 - 0.9 3.9 x 10 6
0.9 - 1.35 2.2 x 10 5
1.35 - 1.8 9.6 x 10 1.8 - 2.2 1.7 x 10 2
2.2 - 2.6 5.4 x 10
'O i!o WAPWR-RP 12.2-43 LUNE, 1985 2962e:1d s
..-._--,..--,..__-.,.,,m-
_..m.._,_.__m,
_.,_,,._.r._
. = -. _ - -
l TABLE 12.2-10 (Sheet 3 of 3) 1 i
RADIATION SOURCES - STEAM GENERATOR BLOWDOWN PROCESSING' SYSTEM i
I 1
Dimensions and Composition of SGBPS Filters i
Source Dimensions Source Composition Filter (inches)
(volume percent)
Blowdown prefilter Radius = 5.0 Air - 66%
Length = 36 Water - 325 i
Stainless Steel - 2%
Blowdown outlet Radius = 3.375 Air - 67%
Spent resin sluice Length = 19 Water - 33%
i i
i 1
p.
l l
j t
6 l
WAPWR-RP 12.2-44 JUNE,1985 l
2962e:1d l
t I
O TABLE 12.2-11 RADIATIONSOURCES-RESIDUALHEATREMOVALSYSTEM}.
Enerav GrouD (Mev/Qama)
Source Strenath (Mev/am-sec) 4 Hours 8 Hours After Shutdown After Shutdown 4 (a) 4 (a) 0.2 - 0.4 7.6 x 10 6.5 x 10 4
4 0.4 - 0.9 5.8 x 10 2.8 x 10 4
4 0.9 - 1.35 2.4 x 10 1.1 x 10 3
3 1.35 - 1.8 7.0 x 10 2.3 x 10 3
3 1.8 - 2.2 5.8 x 10 1.3 x 10 3
3 2.2 - 2.6 6.1 x 10 1.3 x 10 2.6 - 3.0 4.5 x 10 9.6 x 10 2
I 3.0 - 4.0 1.8 x 10 3.7 x 10 I
0 4.0 - 5.0 3.0 x 10 6.5 x 10 (a) Includes 80 key xenon-133.
O O
O WAPWR-RP 12.2-45 JUNE, 1985 2962e:1d
O TABLE 12.2-12 CORE AVERAGE GAMMA RAY SOURCE STRENGTHS AT
~5 VARIOUS TIMES AFTER SHUTDOWN O
Source Strenath at Time After Shutdown (Mev/ watt-sec)
Energy Group (Mev/ cama) 12 Hours 24 Hours 100 Hours 1 Week 1 Month 0.2 - 0.4 1.8 x 10 1.6 x 10' B.4 x 10 6.0 x 10 1.5 x 10 9
8 0
0 10 9
0.4 - 0.9 1.1 x 10 9.6 x 10' 6.5 x 10' 5.6 x 10' 3.5 x 10 0.9 - 1.35 2.0 x 10' 1.3 x 10' 6.7 x 10 5.0 x 10 1.5 x 10 8
8 1.35 - 1.8 3.7 x 10' 3.3 x 10 2.7 x 10 2.3 x 10' 6.5 x 10 9
9 8
1.8 - 2.2 3.0 x 10 2.3 x 10 1.5 x 10 1.3 x 10 5.0 x 10 8
8 8
2.2 - 2.6 2.5 x 10 1.8 x 10 1.6 x 10 1.4 x 10 3.9 x 10 6
6 6
6 2.6 - 3.0 6.2 x 10 3.2 x 10 2.7 x 10 2.3 x 10 6.8 x 10 6
6 6
5 5
3.0 - 4.0 4.3 x 10 1.3 x 10 1.1 x 10 9.2 x 10 2.7 x 10 5
4 4.0 - 5.0 2.0 x 10 1.0 x 10 3 Months 6 Months 1 Year 5 Years 7
6 0.2 - 0.4 5.1 x 10 3.0 x 10 1.7 x 10 2.0 x 10 9
9 8
0 0.4 - 0.9 2.0 x 10 1.0 x 10 4.6 x 10 1.2 x 10 I
6 0.9 - 1.35 4.2 x 10 2.8 x 10 2.1 x 10 7.8 x 10 7
I 7
6 1.35 - 1.8 4.5 x 10 1.7 x 10 1.2 x 10 2.2 x 10 1.8 - 2.2 1.8 x 10 1.3 x 10 8.4 x 10 2.4 x 10 6
4 0
2.2 - 2.6 1.5 x 10 1.2 x 10 2
2.6 - 3.0 2.6 x 10 2.0 x 10 4
3.0 - 4.0 1.0 x 10 O
O WAPWR-RP 12.2-46 JUNE, 1985 2962e:1d
TA8LE 12.2-13 tO i
SPENT FUEL GAMMA RAY SOURCE STRENGTHS AT VARIOUS TIMES AFTER SHUTDOWN Source Strenath at Time After Shutdown (Mev/ watt-sec)
Energy Group (Mev/ gamma) 12 Hours 24 Hours 100 Hours 1 Week 1 Month 9
8 8
8 0.2 - 0.4 1.9 x 10' 1.6 x 10 8.6 x 10 6.2 x 10 1.6 x 10 10 10 9
0.4 - 0.9 1.2 x 10 1.0 x 10 6.8 x 10' 5.9 x 10 3.8 x 10' 8
8 8
0.9 - 1.35 2.2 x 10' 1.5 x 10' 8.5 x 10 6.6 x 10 2.2 x 10 9
9 8
8 1.35 - 1.8 3.6 x 10' 3.2 x 10 2.6 x 10 2.2 x 10 6.4 x 10 1.8 - 2.2 4.0 x 10 3.2 x 10 2.4 x 10 2.0 x 10 7.7 x 10 8
8 8
8 7
2.2 - 2.6 2.3 x 10 1.8 x 10 1.5 x 10 1.3 x 10 3.8 x 10 6
6 6
6 2.6 - 3.0 5.6 ' x 10 3.1 x 10 2.6 x 10 2.2 x 10 6.5 x 10 6
6 6
5 5
3.0 - 4.0 4.0 x 10 1.2 x 10 1.0 x 10 8.9 x 10 2.6 x 10 5
3 4.0 - 5.0 1.6 x 10 8.2 x 10 3 Ponths 6 Months 1 Year 5 Years 7
7 6
0.2 - 0.4 5.6 x 10 3.3 x 10 1.9 x 10 2.8 x 10 9
8 8
0.4 - 0.9 2.3 x 10 1.3 x 10' 7.1 x 10 2.1 x 10 0.9 - 1.35 6.4 x 10 4.5 x 10 3.5 x 10 1.4 x 10 6
1.35 - 1.8 5.6 x 10 2.7 x 10 2.0 x 10 3.8 x 10
~
l 1.8 - 2.2 2.2 x 10 1.4 x 10 9.1 x 10 2.6 x 10 2.2 - 2.6 1.5 x 10 1.1 x 10*
6 2.6 - 3.0 2.5 x 10 1.9 x 10 4
3.0 - 4.0 1.0 x 10 O
I O
WAPWR-RP 12.2-47 JUNE,1985 2962e:1d
I i
)
TA8LE 12.2-14
~
CORE AVERAGE AND SPENT FUEL NEUTRON SOURCE
~
STRENGTHS AT VARIOUS TIMES AFTER SHUTDOWN Core Average Spent Fuel Time After Shutdown (n/ watt-sec)
(n/ watt-sec) l F
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 29.5 71.4
)
L 24 hours 29.5 71.4 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> 29.3 71.4 1 week 29.3 71.2 1 month 28.3 69.5 3 months 26.7 65.7 6 months 24.5 61.4 1 year 22.0 55.9 5 years 17.4 44.8 6-83 to 93 percent of the neutron source strength is due to the spontaneous fission of curium-242 and curium-244.
The curium spontaneous fission neutron spectrum is quite similar to that of californium-252.
The californium-252 spontaneous fission neutron spectrum may be expressed by a Watt formula as follows:
O x(E) = 0.37 EXP(-0.88E) SINH (/2.0E) where E is the neutron energy and x(E) is normalized so that O
I x(E) dE a 1 o
i WAPWR-RP 12.2-48 JUNE, 1985 2962e:1d l
TABLE 12.2-15 (Sheet 1 of 2) i IRRA01ATED Sb-Be SECONDARY SOURCE ROD SOURCE STRENGTHS O
Gansna Rav 3
Source Strenath 'at Time Af ter Shutdown (Mev/cm _,,c)
(Nev/aansnal l_p.gy 1 Week 1 Month 6 Months 1 Year S Yeart i
10 10 10 10 9
7 0.2 - 0.4 3.0 x 10 2.9 x 10 2.5 x 10 1.1 x 10 3.7 x 10 2.2 x 10 13 12 12 II 10 8
0.4 - 0.9 1.1 x 10 7.0 x 10 4.6 x 10 8.1 x 10 9.7 x 10 1.8 x 10 l
0.9 - 1.35 6.7 x 10" 4.8 x 10" 3.4 x 10" 6.0 x 10 7.0 x 10' 10 12 12 12 U
1.35 - 1.8 7.6 x 10 7.1 x 10 5.5 x 10 9.7 x 10 1.2 x 10" t
U 10 1.8 - 2.2 9.8 x 10" 9.1 x 10" 7.0 x 10 1.2 x 10" 1.5 x 10 The secondary source cross-sectional area is 3.96 cm per rod.
The Sb-Be material density is 3.38 gm/cm.
o O
l WAPWR-RP 12.2-49 JUNE, 1985 l
2962e:1d
,n-,
,n.n.
_n,m, n,,,v--,
_--,.-.-,,-w,,_w.
TABLE 12.2-15 (Sheet 2 of 2)
{
IRRADIATED Sb-Be SECONDARY SOURCE ROD SOURCE STRENGTHS o
Neutron Neutron Sou ce Strength Time After Shutdown (n/cm -sec) 8 1 day 4.5 x 10 1 week 4.2 x 10 j
8 l
8 1 month 3.2 x 10 7
6 months 5.8 x 10 6
1 year 6.8 x 10 p
5 years l
l The secondary source rod cross-sectional area is 3.96 cm per rod.
The average neutron energy is 30 key.
3 The St,-Be material density is 3.38 gm/cm.
O l
l O
WAPWR-RP 12.2-50 JUNE,1985 2962e:1d
TABLE 12.2-16 O
IRRADIATED INCORE DETECTOR AND ORIVE CABLE MAXIMUM-WITH0RAWAL SOURCE STRENGTHS Energy Group Incore Detector Drive Cable 3
(Mev/aamma)
(Mev/cm -sec)
(Mev/cm -sec) 10 0.2 - 0.4 8.1 x 10 1.9 x 10' l
II ll 0.4 - 0.9 3.3 x 10 1.4 x 10 1
II 10 0.9 - 1.35 2.4 x 10 2.9 x 10 II 1.35 - 1.8 2.4 x 10 1.1 x 10' 10 II 1.8 - 2.2 6.2 x 10 1.3 x 10 10 10 2.2 - 2.6 6.6 x 10 4.2 x 10 10 2.6 - 3.0 3.4 x 10 4.5 x 10' 9-10 3.0 - 4.0 2.1 x 10 1.1 x 10' 10 4.0 - 5.0 1.1 x 10 i
~
9 5.0 - 6.0 3.0 x 10 The ef fective diameter and length of the incore detector are 0.305 and 3.56 cm respectively.
O 2
The ef fective cross-sectional area of the drive cable is 0.0302 cm.
l O
WAPWR-RP 12.2-51 JUNE, 1985 2962e:1d i
l
et TABLE 12.2-17 r
IRRADIATED INCORE DETECTOR DRIVE CABLE SOURCE STRENGTHS ' i'<
~
Source Strength at Time Af ter Shutdown (Mev/cm -seb' Energy Group (Mev/gauna) 8 Hours 1 Dav'
,/
1 Week 1 Month 6 Months 1 Year
.hYears i<
9 9
9 8
5 0.2 - 0.4 1.9 x 10 1.8 x 10,
1.6 x 10 9.1 x 10 2.4 x 10 4.2 x 10 10 10 10 9
0 0.4 - 0.9 2.9 x 10 1.5 x 10 1.4 x 10 1.3 x 10 7.9 x 10 4.9 x 10 1.9 x 10 r
10 10 10 10 10 10 9
2.4 x 10 1.7 x 10 1.5 x 10 9.1 x 10 3.9 - 1.35 2.8 x 10 2.8 x 10 2.7 x 10 I
I I
6 6
1.35 - 1.8 7.7 x 10 4.2 x 10 3.9 x 10 3.1 x 10 7.3 x 10 1.2 x 1C 7
l 10 (8
1.8 - 2.2 1.5 x 10 2.1 x 10 f.
8 6
2.2 - 2.6 5.0 x 10 6.8 x 10 f,
1 9
6 2.6 - 3.0 5.3 x 15 7.2 x 10 9
6 3.0 - 4.0 1.2 x 10 1.7 x 10 1
2 1he drive cable effective cross-sectional area is 0.0302 cm,
1:0 e
e
O O
OO
TABLE 12.2-18 IRRADIATED INCONEL 600 (0.10 WEIGHT PERCENT Co)
FLUX THIMBLE SOURCE STRENGTHS Source Strength at Time Af ter Shutdown (Mev/cm -sec)
Energy Group (Mev/ gamma) 12 Hours 1 Day 1 Week 1 Month 6 Months 1 Year 5 Years 9
9 9
9 I
5 0.2 - 0.4 7.2 x 10 7.1 x 10 6.1 x 10 3.4 x 10 8.1 x 10 8.0 x 10 ll ll ll Il 10 9
I 3.4 - 0.9 1.5 x 10 1.5 x 10 1.4 x 10 1.1 x 10 2.1 x 10 4.6 x 10 1.7 x 10 10 10 10 10 10 10 10
).9 - 1.35 4.6 x 10 4.6 x 10 4.5 x 10 4.5 x 10 4.2 x 10 3.9 x 10 2.3 x 10 10 9
9 9
8 7
1.35 - 1.8 2.0 x 10 1.6 x 10 1.4 x 10 1.2 x 10 2.7 x 10 4.5 x 10 8
5 1.8 - 2.2 1.7 x 10 6.8 x 10 l
4 2.2 - 2.6 5.6 x 10 2.2 x 10 7
4 2.6 - 3.0 6.0 x 10 2.4 x 10 6
3 3.0 - 4.0 1.4 x 10 5.5 x 10 2
The flux thimble cross-sectional area is 0.466 cm,
g,~9 O
O
O O
O
=~e O
s--
l TABLE 12.2-19 INTEGRATED GAMMA RAY AND BETA SOURCE STRENGTHS AT VARIOUS DMES FOLLOWING A MAXIMUM CREDIBLE ACCIDENT (TID-14844 Release Fractions) l l
Source Strength at Time After Release (Mev/ watt)
Energy Group l
l (Mev/camma) 0.5 Hour 1 Hour 2 Hours 8 Hours 1 Day 12 12 12 I3 I3 0.2 - 0.4 1.3 x 10 2.3 x 10 4.0 x 10 1.3 x 10 3.3 x 10 12 13 13 13 I3 0.4 - 0.9 8.4 x 10 1.5 x 10 2.4 x 10 5.3 x 10 8.6 x 10 12 12 13 13 13 0.9 - 1.35 3.7 x 10 6.4 x 10 1.1 x 10 2.5 x 10 3.8 x 10 12 12 13 13 13 1.35 - 1.8 3.6 x 10 6.3 x 10 1.0 x 10 2.1 x 10 3.0 x 10 12 12 12 13 I3 1.8 - 2.2 1.9 x 10 3.2 x 10 5.2 x 10 1.1 x 10 1.3 x 10 12 12 12 13 13 2.2 - 2.6 2.0 x 10 3.7 x 10 6.1 x 10 1.2 x 10 1.4 x 10 2.6 - 3.0 3.4 x 10" 5.5 x 10" 8.4 x 10" 1.3 x 10 1.4 x 10 12 12 3.0 - 4.0 3.6 x 10' 4.6 x 10" 5.8 x 10" 8.0 x 10" 8.3 x 10" 4.0 - 5.0 1.6 x 10" 1.6 x 10" 1.7 x 10" 2.0 x 10" 2.0 x 10" 10 10 10 10 10 5.0 - 6.0 1.1 x 10 1.1 x 10 1.1 x 10 1.1 x 10 1.1 x 10 13 13 13 I3 I4 Beta 1.3 x 10 2.2 x 10 3.6 x 10 8.9 x 10 1.5 x 10 1 Week 1 Month 6 Months 1 Year I
I4 I4 I4 6
0.2 - 0.4 1.3 x 10 2.3 x 10 2.6 x 10 2.6 x 10 I4 I4 I4 8
0.4 - 0.9 1.7 x 10 2.7 x 10 5.3 x 10 6.4 x 10 13 13 13 0
0.9 - 1.35 4.8 x 10 5.3 x 10 6.0 x 10 6.4 x 10 13 13 13 6
1.35 - 1.8 4.6 x 10 7.2 x 10 8.6 x 10 8.9 x 10 13 13 13 5
1.8 - 2.2 1.4 x 10 1.5 x 10 1.8 x 10 2.0 x 1G 13 I
I 13 2.2 2.6
- 1. 5 x 10 1.7 x 10 1.8 x 10 1.8 x 10 12 12 12 12 2.6 - 3.0 1 4 x 10 1.5 x 10 1.5 x 10 1.5 x 10 3.0 - 4.0 8.4 x 10' 8.5 x 10"
- 8. 5 x 10" 8.5 x 10" 4.0 - 5.0 2.0 x 10" 2.0 x 10" 2.0 x 10" 2.0 x 10" 10 10 10 10 5.0 - 6.0 1.1 x 10 1.1 x 10 1.1 x 10 1.1 x 10 I
15 Beta 3.9 x 10
- 6. 5 x 10" 1.1 x 10" 1.4 x 10 WAPWR-RP 12.2-54 JUNE, 1985 2962e:1d
l l
l TABLE 12.2-20 SOURCE STRENGTH IN THE RESIDUAL HEAT REMOVAL LOOP AT VARIOUS TIMES FOLLOWING AN EQUIVALENT FULL CORE MELTDOWN ACCIDENT t
Source Strength at Time After Release (Mev/am-set-watt)
Energy Group (Mev/camma)
O Hour 0.5 Hour 1 Hour 2 Hours 8 Hours
-I
~I
-I
~I
-I 0.2 - 0.4 2.6 x 10 1.8 x 10 1.6 x 10 1.5 x 10 1.3 x 10
-I 0.4 - 0.9 3.1 2.4 1.9 1.4 5.3 x 10
-I
-I 0.9 - 1.35 1.5 9.4 x 10 8.2 x 10 6.5 x 10 2.9 x 10
-I
-I
-I
-I 1.35 - 1.8 1.1 6.3 x 10 5.4 x 10 4.3 x 10 1.9 x 10 1.8 - 2.2 1.0 x 10' 6.5 x 10 5.2 x 10-3.8 x 10 1.4 x 10
~
~
-I
-2
-2
-2
-2 2.2 - 2.6 2.2 x 10 4.4 x 10 3.6 x 10 2.7 x 10 1.2 x 10 2.6 - 3.0 2.7 x 10' 4.5 x 10-2.1 x 10 5.7 x 10
~
~
-I
-2
-2
-3 3.0 - 4.0 1.5 x 10 3.1 x 10 1.5 x 10 4.1 x 10
-I
-4
-4 O
4.0 - 5.0 1.6 x 10 4.0 x 10 2.0 x 10
-3 5.0 - 6.0 1.2 x 10 1 Day 1 Week 1 Month 6 Months 1 Year V
~I
-2
-3
-4
-4 0.2 - 0.4 1.1 x 10 6.5 x 10 9.4 x 10 2.0 x 10 1.2 x 10
-3 0.4 - 0.9 2.8 x 10" 5.4 x 10 2.6 x 10-7.2 x 10 3.1 x 10'
~
-2
-3
-3
~4
-4 O.9 - 1.35 7.0 x 10 3.6 x 10 1.0 x 10 1.9 x 10 1.4',x 10 1.35 - 1.8 5.1 x 10 1.6 x 10 4.4 x 10-1.2 x 10 8.4 x 10 '
~
~
l
-3
-4
~4
-5 1.8 - 2.2 3.3 x 10 8.6 x 10 3.4 x 10 9.0 x 10 5.7 x 10-5 O
2.7 x 10~#
~
2.2 - 2.6 3.2 x 10 9.2 x 10 l
O O
WAPWR-RP 12.2-55
. LUNE, 1985 2962e:1d
12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 Facility Design Features l'.
Specific design f eature s for maintaining personnel exposure as low as reasonably achievable (ALARA) are discussed in this subsection.
The design
~
feature reconsnendations given in Regulatory Guide 8.8, Paragraph C.2, are utilized to minimize exposures to personnel.
12.3.1.1 Plant Design Description for ALARA 1
1 The equipment and plant design features employed to maintain radiation exposures ALARA are based upon the design considerations of Subsection 12.1.2 and are outlined in this subsection.
12.3.1.1.1 Equipment and Component Designs for ALARA This subsection describes the design features utilized for several general classes of equipment or components.
These classes of equipment are common to many of the plant systems; thus, the features employed for each system to maintain minimum exposures are similar and are discussed by equipment class i,n the following paragraphs.
12.3.1.1.1.1 Nuclear Power Block (NPB) Equipment A.
Reactor Vessel The reactor vessel design includes an integrated head package which combines the head lifting rig, control and gray rod drive mechanism (CRDM's/GRDM's), water displacer rod drive mechanisms (WDR's) seismic supports, lif t columns, reactor vessel missile shield, CRDM cooling O-system and power and instrumentation cabling into an effective, one-package reactor vessel head design.
Mounted directly on the reactor vessel head, the system helps to minimize the time, manpower and radiation exposure associated with head removal and replacement O
WAPWR-RP 12.3-1 JUNE, 1985 3061e:1d
4 during refueling.
Integral in the design is permanent shielding for reducing work area dose rates for the CRDM drive shafts.
The conventional top mounted instrumentation ports /conoseal thermocouple arrangement is replaced with a combination thermocouple /
incore detector system on the WAPWR.
This improvement eliminates the need to disassemble and reassemble the instrument port conoseals at each refueling, which has historically been a relatively high radiation exposure task.
The reactor vessel nozzle welds are designed to accomodate reste inspection with ultrasonic sensors.
The nozzle area is tapered along the reinf 3rced areas to ensure a smooth transition, and pipe branch locations are selected to ensure no interference from one branch to the next.
All weld-to-pipe interfaces require a smooth, high quality finish.
Insulation in the area of the reactor vessel nozzle welds is O
fabricated in sections with quick disconnect clasps to facilitate removal of the insulation for inspection of the welds.
+-
B.
Reactor Coolant Pumps The reactor coolant pump design includes assembled cartridge seals which reduce the time required for replacement.
The cartridge seal packages are expected to be capable of operating for 2 years without inspection or maintenance.
Also, the reactor coolant pump casing will be manufactured by a cast method without major welds that require inservice inspection (ISI);
thereby reducing personnel exposure associated with performing ISI.
C.
Steam Generators The Westinghouse APWR steam generator incorporates many design features to facilitate maintenance and inspection in reduced radiation WAPWR-RP 12.3-2 JUNE, 1985 3061e:1d
r fields.(
(a,c)
]The steam generator manways (entrance to channel head) are sized for easier entrance and exit of workers with-protective clothing, and to facilitate the installation and removal of tooling.
The specification of low cobalt tubing material for the WAPWR steam generator is an extremely important feature of the design - not only in terms of reduced exposure relative to the steam generator - but to i
the total plant radiation source term.
The previous limit on the amount of cobalt in steam generator tubing was 0.1 weight percent and has been considered to be a najor source of Co-60 activity in the i
plant.
This limit has been substantially reduced to weight (a,c) percent for the MAPWR design.
Other significant improvements to the steam generator design incluue a (a,c)
~
~
which should eliminate or substantially reduce the need for sludge lancing, and reduces tube and tube support degradation.
Additional changes to increase steam gener,ator reliability will also reduce occupational radiation exposure, (a,c) y 5-
._.m 2
D.
Reactor Coolant Pipe Connections To minimize crud buildup in branch lines, piping connections to the reactor coolant loops are located on or above the horizontal centerline of the pipe wherever possible.
The resistance temperature detector (RTD) bypass manifold and its associated piping and valves has been replaced with an N-16 Transit Time Flow / Power Meter.
This eliminates maintenance on the RTD system components and results in lower radiation fields about the RCS components.
WAPWR-RP 12.3-3 JUNE, 1985 3061e:1d i
4
4 i
i 12.3.1.1.1.2 Auxiliary Equipment A.
Filters i
Filters that accumulate radioactivity are supplied with the means to perform cartridge replacement with semi-remote tools.
Adequate space is provided to allow removing, cask loading, and transporting the cartridge to the1 solid radwaste area.
O I
Radioactive filters are located in a centralized location in the f
reactor external buildings, with a remote filter handling system for l
the removal of spent radioactive filter cartridges from their housings l
and for their transfer to the drunning station for packaging and shipment from the site for burial.
The process is accomplished in such a nenner that exposure to personnel and the possibility of inadvertent radioactive release to the environment is minimized.
Each filter is contained in a shielded compartment and provided with vent and drain valving, and compartment drainage capabilities.
A design criteria for the filter handling system is that it be simple with a minimum of components susceptible to malfunction.
B.
Demineralizers Demineralizers for highly radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to spent resin tanks prior to solidification and so that fresh resin can be loaded into the demineralizer remotely.
Underdrains are designed for full system pressure drop.
The demineralizers and piping are designed with provisions for flushing.
C.
Evaporators Evaporators are provided with chemical addition connections to allow the use of chemicals for descaling operations.
Space is provided to allow removal of heating tube bundles.
The highly radioactive MAPWR-RP 12.3-4 JUNE,1985 3061e:ld me -
l
evaporator components are separated from those that are less radioactive.
Instruments and controls are located in accessible low background radiation areas.
D.
Pumps O
Pumps and associated piping are arranged to provide adequate space for access to the pumps for servicing.
Small pumps are installed in a manner which allows easy removal if necessary.
All pumps in radioactive waste systems are provided with flanged connections for ease of removal.
E.
Tanks In general, horizontal and flat-bottom tanks are sloped downward to the tank drain.
Overflow lines are directed to the waste collection system to control any contamination within plant structures.
For tanks outside structures, which can potentially contain radioactive fluids, dikes are used to contain overflows.
F.
Heat Exchangers V
Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials to minimize leakage.
Impact.baf fles are provided, and tube side and shell side velocities are limited to minimize erosive effects.
Wherever possible, the radioactive fluid passes through the tube side of the heat exchanger.
O G.
Instruments Instrument devices are located in low radiation zones and away f rom radiation sources whenever practical.
Primary instrument devices, which for functional reasons are located in high radiation zones are designed for easy removal to a lower radiation zone for calibration.
!O WAPWR-RP 12.3-5 JUNE, 1985 3061e:1d
l Transmitters and readout devices are located in low radiation zones, such as corridors and the control room, for servicing.
Integral radiation check sources for response verification for airborne radiation monitors and safety-related area radiation monitors GT are provided.
H.
Valves O
To minimize personnel exposures from valve operations, motor-operated, air-operated, or other remotely actuated valves are used where justified by the activity levels and frequency of use.
Valves are located in valve galleries so that they are shielded separately from the major components.
Long runs of exposed piping are minimized in valve galleries.
In areas where manual valves are used on frequently operated process lines, either valve stem extenders or shielding is provided such that personnel need not enter a high radiation area for valve operation.
Valves for clean, non-radioactive systems are separated from radioactive sources and are located in readily accessibliareas.
P Recognizing that valve maintenance can be a major source of personnel radiation exposure, the WAPWR design reflects a reduced number of valves and piping in the auxiliary systems.
This is evidenced in the integrated safeguards system (ISS) design, whereby the independent trains require less interconnecting piping and valves.
O I.
Piping The piping in pipe chases is designed for the lifetime of the unit.
Wherever radioactive piping is routed through areas where routine maintenance is required, pipe chases are provided to reduce the radiation contribution from these pipes to levels appropriate for the inspection or maintenance requirements.
Butt welds are used to the O
WAPWR-RP 12.3-6 JUNE,1985 3061e:1d
fullest extent possible in radwaste piping utilized for transport of spent resins or slurries.
Piping containing radioactive material is routedtominimizeradiationexposuretotheunitpersonneli J.
Floor Drains O
Floor drains and properly sloped floors are provided for each room or cubicle containing serviceable components containing radioactive liquids.
If a ' radioactive drain line must pass through a plant area i
requiring personnel access, shielding is provided as necessary to ensure radiation levels consistent with the required access.
K.
Sample Station The sample station for routine sampling of process fluids is located as shown in Figure 12.3-3.
Proper shielding and ventilation are provided at the local sample stations to maintain low radiation fields in proximate areas and minimize personnel exposure during sampling.
O The sample room is located directly over the chemical drain tank to minimize runs of chemical drain lines.
- Also, the sample heat exchangers inside the sample room are locally shielded to reduce t-background radiation levels inside the room.
L.
Clean Services Whenever possible, clean services and equipment such as compressed air l
{
piping, clean water piping, ventilation ducts, and cable trays are not l
routed through radioactive pipeways.
4 12.3.1.1.2 Facility and Layout ALARA Considerations This subsection describes the design features utilized for standard type plant process and layout situations.
These features are employed in conjunction with the general equipment described in paragraph 12.3.1.1.1 and include the features discussed in the following paragraphs.
-WAPWR-RP 12.3-7 JUNE, 1985 30Gle:ld
~.
A.
Valve Galleries Valve galleries are provided with shielded entrances for personnel protection.
Floor drains are provided to control radioactive leakage.
To facilitate decontamination in valve galleries, concrete d
surfaces are covered with a smooth surface coating which will allow easy decontamination.
The valve gallery shield walls are designed to minimize personnel exposure during maintenance of components within or adjacent to the gallery and to protect personnel who remotely operate O
the valves.
B.
Piping Chases Pipes carrying radioactive materials are routed through controlled access areas properly zoned for that level of activity.
Each piping run is individually analyzed to determine the potential radioactivity level and surface dose rate.
Where it is necessary that radioactive piping be routed through corridors or other low radia *, ion zone areas, shielded pipeways are provided.
Whenever practicable, valves and instruments are not placed in radioactive pipeways.
Equipment compartments are used as pipeways only for those pipes associated with P
equipment in the compartment.,
When possible and practical, radioactive and nonradioactive piping are separated to minimize personnel exposure.
Should maintenance be required, provision is made to isolate and drain radioactive piping and associated equipment.
Potentially radioactive piping is located j
O in appropriately zoned and restricted areas.
Piping is designed to minimize low points and dead legs.
Drains are provided on piping where low points and dead legs cannot be O
eliminated.
In radioactive systems, the use of nonremovable backing rings in the piping points is prohibited.
Whenever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of the main pipe.
WAPWR-RP 12.3-8 JUNE, 1985 3061e:1d
Piping which carries resin slurries or evaporator bottoms is run vertically as much as possible.
Horizontal runs carrying spent resin are sloped toward the spent resin tanks.
Large radius bends are utilized instead of elbows. Where sloped lines or large radius bends are impractical, adequate flush and drain capability is provided to prevent flow blockage and minimize crud traps.
C.
To minimize radiation streaming through penetrations, as many penetrations as practicable are located with an offset between the source and the accessible areas.
If of fsets are not practicable, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel.
If these two methods are not used, alternate means are employed, such as baffle shield walls or grouting the area around the penetration.
D.
Contamination Control Access control and traf fic patterns are considered in the basic plant layout to minimize the spread of contamination.
Equipment vents and drains from highly radioactive systems are piped directly to the collection system instead of allowing any contaminated fluid to flow across to the floor drain.
All-welded piping systems are employed on contaminated systems to the maximum extent practicable to reduce system leakage and crud buildup at joints.
The valves in some radioactive systems are provided with leakoff connections piped directly to the collection system.
Those systems that become highly radioactive, such as the spent resin lines in the radwaste system, are l
provided with flush and drain connections, i
Decontamination of potentially contaminated areas and equipment within the plant is f acilitated by the application of decontaminable paints and suitable smooth-surface coatings to the concrete floors and O
WAPWR-RP 12.3-9 JUNE,1985 3061e:ld
walls.
Floor drains with properly sloping floors are provided in all potentially contaminated areas of the plant.
In addition, radioactive and potentially radioactive drains are separated from nonradioactive drains.
p The role of the ventilation systems in minimizing the spread of airborne contamination is discussed in Subsection 12.3.3 of this module.
l E.
Equipment Layout In those systems where process equipment is a major radiation source
- pumps, valves, and instruments are separated from the process component.
This allows servicing and maintenance of these items in reduced radiation zones.
In general, control panels are located in low radiation zones.
Major components such as tanks, deminerali7ers, and filters in radioactive systems are isolated in individual shielded compartments Gg insofar as practical.
Labyrinth entranceway shields or shielding doors are provided for each compartment from which radiation could t-stream or scatter to access areas and exceed the radiation zone dose limits for those areas.
For potentially high radiation components (such as ion exchangers or heat exchangers and tanks in the primary coolant system) completely enclosed shielded compartments with hatch openings or removable concrete block walls are used.
Provision,is made on some major plant components for removal of these components to lower radiation zones for maintenance.
O Exposure from routine in-plant inspection is controlled by locating, whenever
- possible, inspection points in properly shielded low-background radiation areas.
Radioactive and nonradioactive systems O
are separated as far as practicable to limit radiation exposure from routine inspection of nonradioactive systems.
For radioactive O
WAPWR-RP 12.3-10 JUNE, 1985 3061e:1d
systems, emphasis is placed on adequate space and ease of motion in a properly shielded inspection area.
Where longer times for routine inspection are required and permanent shielding is not feasible, sufficient space for portable shielding is provided. When-this is not practicable, written procedures are used which reduce the total time personnel are exposed to the radiation field.
Also, access to high radiation areas is under the direct supervision of the unit health physics personnel.
12.3.1.2 Radiation Zoning and Access Control Access to areas inside the plant structures and plant yard area is regulated i
and controlled by radiation zoning and access control under the direction of the plant health physics staff.
Each radiation zone defines the radiation level range to which the aggregate of all contributing sources must be attenuated by shielding.
During plant operation, personnel gain access to radiation controlled areas through an access control point.
All plant areas are categorized into radiation zones according to expected radiation levels and anticipated personnel occupancy with consideration given toward neintaining personnel exposures ALARA and within the standards of 10 CFR 20.
Each room, corridor, and pipeway of every plant building is evaluated for potential radiation sources during normal, shutdown, and emergency operalions; for maintenance occupancy requirements; for general access requirements; and for material exposure limits to determine appropriate zoning.
The radiation zone categories employed and their descriptions are given in Table 12.3-1.
The zoning for each plant area under normal operational and refueling outage conditions are shown in Figures 12.3-1 through 17.3-6.
Elevation views of the reactor building which further illustrate the component locations and layout are shown in Figures 12.3-7 and 12.3-8.
The radiation zones shown in the figures represent conservative estimates of the maximum general area dose rates at the various plant areas.
Dose rates at certain locations within the area may exceed the zone value due O
WAPWR-RP 12.3-11 JUNE, 1985 3061e:ld
i to component c rud traps or radiation streaming, but design features are O
incorporated to minimize such effects and the higher dose rates are expected to be highly localized and/or intermittent.
Actual in-plant zones and control of personnel access will be based upon surveys conducted by the p.lant health physics staff.
O Areas which may require occupancy to permit an operator to aid in the long term recovery from an accident are considered in the design.
Such areas include the control room, safety-related motor control centers and switchgear, accident sampling system room, radiochemistry laboratory, and remote O
post shutdown panels.
Such radiation protection design features are described in Section 12.3.2 of this module.
In the event that entry is desired into areas where excessive radiation exposures may occur, due consideration is given to the dose rates in the area, and appropriate time iimits for presence in the area are imposed.
Ingress or egress of plant operating personnel to controlled access areas is controlled by the plant health physics staff to ensure that radiation levels and exposures are within the limits prescribed in 10 CFR 20.
Any area having a radiation level that could cause a whole body exposure in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of 5 mrem, or in any 5 consecutive days in excess of 100 mrem, will be posted " Caution, Radiation Area." Radiation areas are provided with access alert barriers, e.g., chain, rope, door, etc.
Any area having a radiation level that could cause whole body exposure in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of 100 mrem will be posted " Caution, High Radiation Area."
High radiation areas (> 100 mrem /hr) are provided with locked or alarmed barriers.
During periods when access to a high radiation area is required, positive control is exercised To the extent practicable, the measured radiation O
over each individual entry.
level and the location of the source is posted at the entry to any radiation area or high rad'. tion area.
The posting of radiation signs, control of personnel access, and use of alarms O
and locks are in compliance with requirements of 10 CFR 20.203.
O WAPWR-RP 12.3-12 JUNE, 1985 3061e:ld
12.3.2 SHIELDING The bases for the nuclear radiation shielding and the shielding cohfigurations are discussed in this subsection.
12.3.2.1 Design Objectives l
The design objective of the plant radiation shielding, in conjunction with a program of controlled personnel access to and occupancy of radiation areas, is to reduce personnel and population exposures to levels that are within the dose standards of 10 CFR 20 and are as low as reasonably achievable (ALARA).
Shielding and equipment layout and design are considered in ensuring that exposures are maintained ALARA during anticipated personnel activities in areas of the plant containing radioactive materials, utilizing the design recommendations given in Regulatory Guide 8.8, paragraph C.2, where practicable.
Three plant conditions are considered in the nuclear radiation shielding design:
A.
Normal, full-power operation.
B.
Shutdown operation.
~
C.
Emergency operations (for required access to safety-related equipment).
The shielding design objec t ives for the plant during normal operation (including anticipated operational occurrences), for shutdown operations, and for emergency operations are listed below:
1 A.
To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and site boundary occupants are ALARA and within the limits of 10 CFR 20.
O WAPWR-RP 12.3-13 JUNE, 1985 3061c:1d
8.
To ensure sufficient persenel access and occupancy time to allow normal anticipated maintenance, inspection, and safety-related O
operations required for each plant equipment and instrumentation area.
C.
lo reduce potential equipment neutron activation and to mitigate the possibility of radiation damage to materials.
O D.
To provide sufficient shielding for the control room so that for design basis accidents (DBAs) the direct dose plus the inhalation dose O
will not exceed the limits of 10 CFR 50, Appendix A, General Design Criterion 19.
12.3.2.2 General Shielding Design Shielding is provided to reduce the radiation levels resulting from direct and scattered radiation to less than the upper limit of the radiation zone assigned to each area.
General locations of the plant areas and equipment discussed in this subsection are shown in the radiation zone diagrams of O
Section 12.3.1 of this module.
V 1
The material used for most of the plant shielding is ordinary co.ncrete with a 3
P-bulk density of approximately 145 lb/ft.
Whenever poured-in-place concrete has been replaced by concrete blocks, the design ensures protection on an 1
equivalent shielding basis as determined by the density of the concrete block selected.
Water is used as the primary shield material for areas above the spent fuel storage area and in the refueling cavity during refueling operations.
12.3.2.2.1 Reactor Containment Building Shielding Design During reactor operation, the reactor containment building protects personnel occupying adjacent plant structures and yard areas from radiation originating in the reactor vessel and primary loop components.
The concrete containment wall and the reactor vessel and steam generator compartment shield walls O
WAPWR-RP 12.3-14 JUNE,1985 3061e:1d
O sources inside containment.
reduce radiation levels outside the containment to less than 0.25 mrem /hr from The containment shield completely surrounds the nuclear steam supply system with a wall thickness which ranges from 1.5 meters to a minimum of 0.6 meters at the top of the dome.
For design basis accidents (DBAs), the containment shield and the control room shielding reduce the plant radiation intensities from fission products inside the containment to acceptable emergency levels, as defined by 10 CFR 50, A,
General Design Criterion 19, for the control room.
(See O
Appendix Subsection 12.3.2.2.7.)
Where personnel and equipment hatches or penetrations pass through the containment wall, additional shielding is provided to attenuate radiation to the required level defined by the outside radiation zone during normal operation and shutdown and to acceptable emergency levels as defined by 10 CFR 50, Appendix A, General Design Criterion 19, during DBAs.
12.3.2.2.2 Reactor Containment Building Interior Shielding Design During reactor operation, many areas inside the containment are Zone V and normally inaccessible.
However, shielding is provided to reduce dose rates to
- ~~
approximately 100 mrem /h or less in areas of the containment that potentiall'y require access at power.
These are the Zone IV or lower areas shown in Figures 12.3-1 to 12.3-6.
The. main sources of radiation are.the reactor vessel and the primary loop components, consisting of the steam generators, pressurizer, reactor coolant pumps, and associated piping.
The reactor vessel is shielded by the concrete primary shield and by the concrete secondary shield which also surrounds other primary loop components.
Air cooling is provided to prevent overheating, dehydration, and degradation of the structural and shielding properties of the O
primary shield.
!O WAPWR RP 17.3-15 JUNE, 1985 3061e:ld I-
f The primary shield is a large mass of reinforced concrete surrounding the reactor vessel and extending upward from the containment floor to the walls of the fuel transfer canal. The minimum concrete thickness of the pri_ mary shield is 2.5 meters. The primary shield meets the following objectives:
A.
In conjunction with the secondary shield, to reduce the radiation level from sources within the reactor vessel and reactor coolant system (RCS) to allow limited access to the containment during normal, full-power operation.
4 B.
After shutdown, to limit the radiation level from sources within the reactor vessel, to permit limited access to the reactor vessel, and to permit limited access to the RCS equipment.
C.
To limit neutron activation of component and structural materials.
The secondary shield is a reinforced concrete structure surrounding the RCS equipment, including piping, pumps, and steam generators.
This shield personnel from the direct gansna radiation resulting from reactor O
protects coolant activation products and fission products circulating in the reactor coolant.
In addition, the secondary shield supplements the primary shield by p-attenuating neutron and gamma radiation escaping f rom the primary shield.
The secondary shield is sized to allow limited access to the containment during full-power operation.
The minimum thickness of secondary shield walls is 1.2 meters.
e Components of the ~ letdown portion of the chemical and volume control system CVCS in the containment are located in shielded compartments that are O
n(orma)ly l
Zone V,
restricted access areas.
Shielding is provided for each piece of equipment in the letdown system consistent with its postulated activity (Section 12.2 of this module) and with the access and zoning 1
requirements of adjacent areas. This equipment includes the regenerative heat exchanger, the letdown and excess letdown heat exchangers, and the letdown 4
lines.
i WAPWR-RP 12.3-16 JUNE, 1985 4
3061e:1d i
- ~ _ _ _. _ -
i Af ter shutdown, the containment is accessible for limited. periods of time and all access is controlled.
Areas are surveyed to establish allowable working periods.
Dose rates are expected to range from 0.5 to 1000 mren/hr, depending on the location inside the containment.
These dose rates result from residual I
fission products and neutron activation products (coinponents and corrosion products) in the RCS.
Spent fuel is the primary source of radiation during refueling.
Because of the high activity of the fission products contained in the spent fuel elements, extensive shielding is provided for areas surrounding the refueling pool and the fuel transfer canal to ensure that radiation' levels remain below zone levels specified for adjacent areas.
Water shielding is provided over the spent fuel assemblies during fuel handling.
12.3.2.2.3 Reactor External Building Shielding During normal operations, the major components in the reactor external building with potentially high radioactivity are those in the CVCS, steam generator blowdown, boron recycle, liquid radwaste, gaseous radwaste, and spent resin handling systems.
Shielding is provided for each piece of equipment consistent with its postulated activity (Sections 11.1, 11.2, 11.3 of RESAR-SP/90 PDA Module 12, " Waste Management", and 12.2 of this module) and with the access and zoning requirements of adjacent areas.
Depending on the equipment in the area, the radiation zones vary from Zone 1 through V.
Corridors are generally shielded to allow Zone 11 access, and operator areas for valve compartments are generally Zone III.
Removable sections of block shield walls and concrete plugs are utilized as necessary for equipment maintenance and spent filter cartridge replacement.
Permanent or temporary shielding is used between equipment in compartments j
with more than one piece of equipment to permit access for maintenance. Where necessary, labyrinth entrances with provisions for adequate ingress and egress for equipment maintenance and inspection are provided and are designed to be consistent with the access and zoning requirements of adjacent areas.
O WAPWR-RP 12.3-11 JUNE, 1985 l
3061e:1d I
._m
All emergency core cooling system (ECCS), residual heat removal (RHR) and containment spray system components, which are located out'ida the containment O
and s
used to recirculate radioactive coolant during normal or post-accident operation, are housed in dedicated, separated component areas (SC_A).
One of t
the unique features / benefits of the integrated safeguards system (ISS) of the h%P'..'N is the four separate independent subsystems and these features // benefits are maintained in the layout and shield design.
Each SCA includes the piping pencLration area, valve area and pump compartment area associated with one ISS module, as shown in Figures 12.3-1 to 12.3-3.
Layout considerations, which have been addressed, include pump pull space, motor laydown area, work area, monuralls, removable shield walls, distenre to stairs and equipment hatches.
The equipment and shielding layout is expected to result in a minimum of whole body exposure during normal maintenance and inspection and a low potential for significant surface contamination and airborne levels in areas where access to ISS equipment is required, particularly in a post-accident recirculation mode This arrangement provides separation of the low and highly radioactive equipment, one personnel access point for each of the four independent and separate SCA's, and makes it possible to passively prevent a major loss of l
emergency water storage tank (EWST) water outside containment.
Further, the separation of trains reduces the number of components which could potentially be damaged as a result of flooding.
&~
12.3.2.2.4 Fuel Handling Building Shielding Design The concrete shield walls surrounding the spent fuel cask loading and storage l
area, end the shleid walls surrounding the fuel transfer and storage areas, are sufficiently thick to limit radiation levels outside the shield walls in i
all accessible areas to Zone II.
The building external walls are sufficient to shield external plant areas to Zone I.
All spent fuel removal and transfer operations are performed under borated water to provide radiation protection and maintain subcriticality.
Water depths of greater than 3 meters above a fuel assembly during fuel handling are maintained in the reactor cavity, the fuel transfer canal, and the spent fuel
! O WAPWR-RP 12.3-18 JUNE, 1985 3061e:1d l
pool.
This limits the dose at the water surface to less than 2.5 mrem /hr for an assembly in a vert ical position at the maximum elevation.
Normal water depth above the stored assemblies in the spent fuel pit is greater than 8 meters and the dose rate at the pool surface is significantly le'ss than 2.5 mrem /hr.
The minimum 1.5 meter thick concrete walls of the fuel transfer canal and spent fuel pool walls supplement the water shielding and limit the maximum radiation dose levels in working areas to less than 2.5 mrem /h.
The spent fuel pit cooling system (SFPCS) shielding is based on the activity discussed in Section 12.2 of this module, and the access and zoning requiremerts of adjacent areas.
Equipment in the SFPCS to be shielded includes the SFPCS heat exchangers, pumps, and piping.
12.3.2.2.5 Radwaste Buildings Shielding Design The radwaste building is not within the scope of the W APWR NPB and therefore
-the shielding design is the responsibility of the plant specific applicant.
However, the radwaste building shielding design should be consistent with the radwaste source strengths presented in Section 12.2 of this module.
12.3.2.2.6 Turbine Building Shielding Design V
Radiation shielding is not required for process equipment located in the turbine building.
12.3.2.2.7 Control Room Shielding Design
~
i The design basis loss-of-coolant accident (LOCA) dictates the shielding requierments for the control room.
Consideration is given to shielding provided by the containment structure.
Shielding combined with other i
.ngineered safety features is provided to permit access and occupancy of the control room following a postulated LOCA, so that radiation doses are limited to 5 rem whole body from contributing modes of exposure for the duration of the accident, in accordance with 10 CFR 50, Appendix A,
I WAPWR-RP 12.3-19 JUNE, 1985 l
3061e:ld
I i
O The design basis LOCA is described in Subsection 15.6.4 of RESAR SP/90 PDA j
Module 1,
" Primary Side Safeguards System".
The contribution f rom direct radiation f rom airborne fission products inside the containment to personnel doses inside the control room following a postulated LOCA is discussed in Subsection 15.6.4.4.6.3 of RESAR SP/90 PDA Module 1, " Primary Side Safeguards V
System".
The shielding of the control room ensures compliance with 10 CFR 50, Appendix A,
The control room location and shielding provisions are illustrated in Figure 12.3-5.
12 3.2.2.8 Miscellaneous Plant Areas and Plant Yard Areas Suf ficient shielding is provided for all plant buildings containing radiation sources so that radiation levels at the outside surfaces of the buildings are maintained below Zone I levels.
Plant yard areas that are frequently occupied by plant personnel are fully accessible during normal operation and shutdown.
These areas are surrounded by a security fence and closed off from areas accessible to the general public.
O 12.3.2.3 Shielding Calculational Methods The shielding thicknesses provided to ensure compliance with plant radiation P
zoning and to minimize plant personnel exposure are based on equipment activities under the plant operating conditions described in Chapter 11 of RESAR-SP/90 PDA Module 12, " Waste Management", and Section 12.2 of this The thickness of each shield wall surrounding radioactive equipment module.
is determined by approximating as closely as possible the actual geometry and p
physical condition of the source or sources.
d The geometric model assumed for shielding evaluation of
- tanks, heat exchangers, filters, ion exchangers, evaporators, and the containment is a finite cylindrical volume source.
For shielding evaluation of piping, the O
geometric model is a finite shielded cylinder.
In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source.
O 1
WAPWR-RP 12.3-20 JUNE,1985 3061e:ld
I}
I)
Industry-accepted computer codes
- ANISN-WIII, D01-IllW SCAP and IU MORSE
- are used for shielding analysis.
ANISN is a multi-group one-dimensional discrete-ordinates transport code that solves the one-dimensional Boltzmann transport equation for neutrons and ganna rays in slab, sphere, or cylinder geometry.
Using a finite-dif ference technique, ANISN allows general anisotropic scattering; i.e.,
an Lth order Legendre exg.m lon of the scattering cross-sections.
D0T-IllW is a two-dimensional discrete-ordinates transport code which effects a solution of the Boltzmann transport equation.
DOT-IIIW permits anisotropic scattering to be included and is suitable for both neutron and gamma ray deep penetration calculations in a wide variety of shielding problems.
Monte Carlo techniques as described below may be used for more complicated geometries such as penetrations.
SCAP is a single-scatter point-kernel general purpose code for estimating the penetration and scattering of gamma rays that originate in a volume-distributed source.
ANISN and 00T are used primarily for primary shield design.
SCAP and DOT are used primarily for configurations not conveniently modeled in one-dimensional geometries.
For final design, a three-dimensional model,is used to simulate radiation streaming from the reactor vessel surface to the containment using the MORSE Monte Carlo program.
The source terms used for the MORSE code are generated P
by the computer code DOT.
The source terms are divided into 47 neutron energy groups and 20 ganna energy groups.
Neutron and gamma ray cross-sections are prepared using the AMPX-11 processing system. The reference cross-section library is Vitamin-C.Ib)
The shielding thicknesses are selected to reduce the aggregate computed radiation level f rom all contrib.. og sources below the upper limit of the radiation zone specified for each plant area.
Shielding requirements in each plant area are evaluated at the point of maximum radiation dose through any wall.
Since the actual failed fuel fraction during normal plant operation is O'
expected to be less than 0.25%, the actual anticipated radiation level in most plant areas is less than this neximum dose and consequently less than the radiation zone upper limit.
O WAPWR-RP 12.3-21 JUNL, 1985 t
3061e:ld
i I
i Where shielded entryways to compartments containing high-radiation sources are necessary, labyrinths are designed using methods sunenarized in ORNL-RSIC-21 and the SCAP computer code.
The labyrinths are constructed so that the scattered dose rate, plus the transmitted dose rate through the shield. wall f rom all contributing sources, is below the upper limit of the radiation zone specified for each plant area.
a O
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p a
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WAPWR-RP 12.3-22 JUNE,1985 3061e:1d i
T 12.3.3 Ventilation 12.3.3.1 Design Objectives The plant ventilation systems, in addition to their piarary function of preventing extreme thermal environmental conditions for operating personnel and equipment, will provide effective protection for operating personnel against possible airborne radioactive contamination in areas where this may occur.
The systems will operate to ensure that the maximum airborne radioactivity level for normal operation, including anticipated operational occurrences, are within tae limits of 10CFR20, Appendix 8 Table 1, for areas within plant structures and on the plant site where construction workers and visitors are permitted.
The maximum levels correspond to design-basis reactor coolant inventory.
The average airborne radioactivity levels meet the requirements of 10CFR20 and 50 and in fact will be considerably smaller since average coolant inventories and actual equipment leakage will be small.
The systems will operate to ensure compliance with normal operation offsite release limits as discussed in Section 11.3 of RESAR SP/90 PDA Module 12
" Waste Managetrent".
The control rcom ventilation system will also operate to provide a suitable environment for equipment and continuous personnel occupancy in the control room under post-accident conditions in accordance with 10CFR50, Appenaix A, Criterion 19.
O The expected airborne radioactivity levels for normal operations and anticipated operational occurrences, in areas within plant structures, including each building in the reactor facility, and on the plant site where personnel, constructior workers, or site visitors are permitted, along with the assumptions and methods used to calculate these airborne radioactivity levels will be presented in the Applicant's safety Analysis Report.
A discussion of the resulting estimated doses will also be presented.
O WAPWR-RP 12.3-23 JUNE, 1985 3061e:1d
i 12.3.3.2 Design Guidelines
- O i
In order to accomplish the design objectives, certain general design guidelines are followed when possible and applicable.
~
1.
Air movement patterns are provided from areas of lesser contamination potential to areas of progressively greater contamination potential prior j
to final exhaust.
2.
Slightly negative pressures are maintained, where applicable, to prevent l
uncontrolled exfiltration of contamination.
Slightly positive pressure is maintained in the control room to prevent infiltration of potential i
contaminants.
3.
Valves and equipment are maintained as leaktight as possible in order to prevent leakage of radioactive water and subsequent air-borne contamination.
4.
Individual air supplies are provided for each building in order to keep I
potentially contaminated air flows separate f rom noncontaminated air.
V 5.
The fresh air supply to the control room is designed to be operable during loss of offsite power.
The air is filtered to prevent contamination of the control room.
The plant ventilation system is described in further detail in RESAR-SP/90 PDA Mcdule 13 " Auxiliary Systems".
O O
WAPWR-RP 12.3-24 JUNE, 1985 3061e:ld
12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation The radiation monitoring system consists of the following:
A.
Area radiation monitoring system (ARMS).
O.
B.
Process and effluent radiation monitoring system (PERMS).
C.
Sampling system.
D.
Post-accident monitoring systems (PAMS) radiation monitors.
The PERMS, sampling systems, and PAMS (airborne) radiation monitors are described in Section 11.5 of RESAR-SP/90 PDA Module 12, " Waste Management".
12.3.4.1 Area Radiation Monitoring The ARMS is provided to supplement the personnel and area radiation survey provisions of the plant health physics program described in Section 12.5 and to ensure compliance with the personnel radiation protection guidelines of O
10 CFR 20,10 CFR 50,10 CFR 70, and Regulatory Guides 1.97, 8.2, 8.8, and 8.12.
The design of the fuel pool racks precludes criticality under all postulated 9-normal and accident conditions.
Therefore, criticality monitors, as stated in 10 CFR 70.24 and Regulatory Guide 8.12, are not needed.
12.3.4.1.1 Design Objectives The design objectives of the ARMS during normal operating plant conditions and 4
anticipated operational occurrences are:
A.
To furnish records of radiation levels in specific areas of the plart.
O B.
To _ warn of uncontrolled or inadvertent movement of radioactive material in the plant.
i O WAPWR-RP 12.3-25 JUNE, 1985 3061e:1d 4
--.,-.m..
...____4.
C.
To provide local and remote indication of ambient ganna radiation and local and remote alarms at key points where substantial change in radiation levels might be of immediate importance to, personnel frequenting the area.
O D.
To annunciate and warn of possible equipment malfunctions and leaks in specific areas of the plant.
E.
To furnish information for making radiation surveys.
By meeting the above objectives, the area radiation monitoring system aidh health physics personnel in keeping radiation exposures a: low as reasonably achievable (ALARA).
The design objectives of the ARMS during postulated accidents are:
A.
To provide the capability to alarm and initiate a containment O
isolation phase A or containment ventilation isolation signal in the event of a loss-of -coolant accident (LOCA), fuel handling accident inside containment, or abnormally high radiation inside the containment (monitors A-2A&B and A-11 A&B).
p.
B.
To provide long-term post-accident monitoring of conditions at strategic locations.
(See Subsection 11.5.5 of RESAR-SP/90 PDA Module 12, " Waste Management").
~,
12.3.4.1.2 Criteria for Location of Area Monitors l
Considerations for area monitor locations are based on the following:
Frequency and length of personnel occupancy of a specific area.
O A.
B.
Potential for personnel to unknowingly receive high radiation doses.
C.
Potential for equipment malfunction.
O WAPWR RP 12.3-26 JUNE,1985 3061e:1d
l 0.
Access areas
- where, during nonnal plant operation (including O
refueling), radiation exposures could exceed the radiation limits due to system failure or personnel error.
j E.
Access areas where new and spent fuel is received and stored.
F.
Containment area for indicating the level of radioactivity and detecting the presence of fission products due to a reactor coolant pressure boundary (RCPB) leak, or fuel handling accident.
O G.
Normally or potentially radioactive areas.
12.3.4.1.3 General System Description The area radiation monitors are located at selected locations throughout the i
plant to detect, indicate, and store information through their associated data processing module on the radiation levels and, if necessary, annunciate O
abnonnal radiation conditions.
The ARMS monitors are an integral part of the PERMS, which is described in detail in Section 11.5 of RESAR-SP/90 PDA Module
- 12. " Waste Management".
P The ARMS consists of individual, locally mounted area monitors.
Each monitor is composed of the requis,ite number of channels, with a channel consisting of a radiation detector and check source.
The detectors for all area nonitors are either gamma-sensitive Geiger-Mueller counter tubes or ionizat. ion chambers.
If exposed to radiation in excess of full-scale indication, the area nonitors indicate that the full-scale reading has been exceeded and remain at the full-scale value.
If the radiation field f.ausing the overload condition is removed, the system returns to its normal operating condition unless the detector has failed.
An administrative procedure (positioning the check source) is initiated to ensure that radiation monitoring equipment has not been damaged.
All channels associated with a monitor are served by a local dedicated data processing module.
All channels are indicated and annunciated in the control room and indicated end alarmed near the detector O
WAPWR-RP 12.3-2'l JUNE,1985 3061e:ld I
location (normally at the data processing module).
Monitors A-2A&B and O
A-11 A&B, which are safety-related, Class 1E.
are also indicated and annunciated at the safety-related display console.
12.3.4.1.4 Data Processing Module and Display Console A description of these components is given in Section 11.5 of RESAR-SP/90 PDA Module 12, " Waste Management".
12.3.4.1.5 tocal Annunciation All area monitors have local annunciation consisting of an audible alarm rated at 80 48 at 10 f t and a warning light at the local readout.
l 12.3.4.1.6 Power Supplies Each channel is provided with an independent power supply, designed such that a failure in that channel does not affect any other channel.
Monitors that are identified as safety-related are redundant and are supplied power from the plant 120-V safety-related buses.
Power to the channels that monitor only normal operations is supplied from the regulated 120-V instrumentation bus
- ~
that is backed by the diesel generator.
12.3.4.1.7 Redundancy, Diversity, and Independence Monitors designated safety-related are part of the safety-related portion.of the PERMS and are designed for redundancy, diversity, and independence in O
accordance with Institute of Electrical and Electronic Engineers (IEEE) 344-1975, IEEE 336-1971, IEEE 279-1971, IEEE 308-1974, IEEE 323-1974, and IEEE 384-1974. All monitors which are Seismic Category I are also manufactured and rated to the above standards.
O O
WAPWR-RP 12.3-28 JUNE, 1985 3061e:1d
12.3.4.1.8 Area Monitor Description l O l
Table 12.3-2 gives the conditions of service for the area monitors.
A brief description of each area monitor's function is given below.
A.
Control Room Area Monitor A-1 To continuously indicate the radiation levels in the control room.
A high alarm signal warns control room personnel of a deteriorated radiological condition inside the control room.
B.
Containment Low-Range Area Monitors A-2A and A-2B To continuously indicate the radiation levels inside the containment building at the operating deck.
During refueling operations a high radiation alarm indicates a fuel drop accident. and isolates the containment ventilation system.
During power operations, a high radiation alarm indicates a possible LOCA and isolates the containment ventilation system.
l C.
Radiochemistry Laboratory Area Monitor A-3 V
l l
To continuously indicate the radiation levels in the radiochemistry laboratory.
A high radiation alarm signal warns the occupants of the radiochemistry laboratory of a deteriorated radiological condition..
D.
Fuel Handling Building Area Monitor A-5 lo continuously indicate the radiation levels inside the fuel handling building.
A high radiation alarm signal warns the occupants of the fuel handling building of a deteriorated radiological condit f un.
O l
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WAPWR-RP 12.3-29 JUNE, 1985 3061e:1d
E.
Sampling Room Area Monitor A-6 To continuously indicate the radiation levels in the sampijng room.
A high radiation alarm signal warns the occupants of the sampling room of a deteriorated radiological condition.
F.
Seal Table Instrur entation Room Area Monitor A-7 To continuously indicate the radiation levels in the seal table room and establish radiological habitability prior to entry.
A high radiation riaru signal warns occupants of the seal table room of a deteriorated radiological condition.
G.
Containment Access Hatch Area Monitor A-9 i
To continuously indicate the radiation levels in the containment access hatch and establish radiological habitability prior to entry.
H.
Containment High Range Area Monitors A-11 A and A-llB To indicate, along with A-2A and A-2B, the radiation levels inside the containment building at the operating deck following a design basis accident.
A high alarm signal initiates containment isolation phase A to mitigate the consequences of a design basis accident (primarily a LOCA).
12.3.4.1.9 Range and Alarm Setpoints O
The range, setpoints, and control function of the PERMS area monitors are given in Table 12.3-3.
The setpoints are initial and are subject to modification as plant operating experience is developed.
Radiation zones are described in Table 12.3-1.
~
O WAPWR-RP 12.3-30
. LUNE, 1985 3061e:1d a.
1 The control room monitor A-1 has a greater sensitivity than the other area monitors, since it is located in a Zone I radiation area; monitors A-2A&B and A-11 A&B cover a wide range of radiation levels.
During plant shutdown including refueling operations, the radiation level on and above the operating deck should be less than 5 mR/hr.
The high end of the range is dictated by the design basis accident, a LOCA.
Each area monitor has two alarm setpoints, intemediate and high.
(See Table 12.3-3.)
If a nonitor has a control function, i.e., A-2A&B and A-11 A&B, the control function is triggered coincidentally with the high alarm setpoint. An intermediate alarm gives a visual indication in the control room and near the detector that the radiation level has reached the intermediate setpoint.
A high alann gives both a visual and audible indication near the detector (along with a visual indication and annunciation in the control room) that the high alarm setpoint has been reached.
For testing, each area monitor has a check source assembly which is operated from the control console and uses a
sealed Sr-90 source.
Inservice inspection, calibration, and maintenance of the ARMS monitors is discussed in Subsection 11.5.2.5 of RESAR-SP/90 PDA Module 12, " Waste Management".
p.
O O
O WAPWR-RP 12.3-31 JUNE, 1985 3061e:1d
1 i
12.3.5 References O
1.
- Soltesz, R.
G.,
et al, Final Proaress Report. Nuclear Rocket Shieldina Methods. Modification. Updatina, and Input Data Prepa ration. Vol.
4.
"One-Dimensional Discrete Ordinates Transport Techniaue." WANL-PR(LL)034, August 1970.
2.
- Soltesz, R.
G.,
et al, Final Proaress Report. Nuclear Rocket Shieldina Methods. Modification. Updatina, and Input Data Preparation. Vol.
5.
b "Two-Dimensional Discrete Ordinates Transport Techniaue." WANL-PR(LL)C34,
'J August 1970.
3.
- Soltesz, R.
G.,
et al, Final Proaress Report. Nuclear Rocket Shieldina
!!ethods. Modification. Updatin?. and InDut Data Preparation. Vol.
6.
" Point Kernel Techniaues." WANL-PR(LL)034, August 1970.
4.
RISC Computer Code Collection CCC-203, MORSF -CG.
General Purpose Monte Carlo Multi-Group Neutron and Gansna Ray Transport Code with Combinatorial Geometry.
5.
ORNl RSIC PSR -63, "AMPX-II, Modular GDE System for Generating Coupled 9-Multigroup Neutron Gamma Ray Cross-Section Libraries f rom Data in End of Fornat".
6.
ORNL RSIC DLC-41,
" Vitamin-C, 171
- Neutron, 36 Gansna Ray Group Cross-Sections in AMPX Interface Formats for Fusion and LMFBR Neutronics'.
7.
- Selph, W.
E.,
" Neutron and Ganrna Ray Albedos," O_RNL-RSIC-21, Oak Ridge National Laboratory, February 1968.
O O
WAPWR RP 12.3-32 JUNE,1985 70ble:ld
TABLE 12.3-1 RADIATION '20NE DEFINITIONS Maximum Dose Rate Zone _
(arem/hr)
Description ( }
l' O1 0.25 Uncontrolled access, no limitations on occupancy 11 2.5 Controlled access, unlimited occupational access III 5
Controlled access, frequent routine access IV 100 Controlled access, infrequent non-routine access V
> 100 Controlled access, access highly limited or not accessible i
i 4
a.
jLncontrol'ed access:
Where entry and exit by plant employees and visitors
- ~
are not under the direct supervision of the plant health physics staf f.
These areas can be occupied by plant personnel or visitors on an unlimited time basis with a minimum probability of health hazard from radiation exposure.
Controlled access:
Where higher radiation levels and/or radioactive contamination, which have a greater probability of radiation health hazard to individuals, can be expected.
Only individuals directly involved in
{,
the operation of the ~ plant will, in general, be allowed to enter these areas.
Entry and exit are authorized and supervised by the plant health physics staff.
O WAPWR-RP 12.3-33 JUNE, 1985 3061e:1d
O C
O O
O 1
1A8LE 12.3-2 I
6
?
CON 0lil0NS OF SERVICE FOR AREA RADIA110N NON110RS
+.
i Operating Number of Temperature Relative Radiation Safety Location /
Area Nonitor Detectors
(*F1 Pressure Humidity (f)
Zone Classification Elevation (Neters1 A-1 control room 1
65-85
-1/8 in, to 50 (max) i NNS Control room at 100.0 i
+1/2 in. E l
A-2A.8 containment 2
60-120
-1.5 in. E to 17.1 to 50 V
SC-3/1E Containment at 100.0 i
low range
+3 psig i
}
A-3 radiochemistry 1
65-1C4 0 psig 40 to 60 til NNS Radlochemistry lab at f
laboratory 84.8 A-5 fuel handling 1
40-104
-1/4 in. to 20 to 95 11 NNS Fuel handling build-building 0 in. E ing at 100.0 1
A-6 sampling room 1
65-100 0 pstg 40 to 60 IV NNS Sample room at 84.8 A-1 seal table 1
60-120
-1.5 in. E to 11.1 to 50 IV NNS Seal table room at instrumentation room
+3 psig g3.2 i
A-g containment 1
60-120
-1.5 in. E to 11.1 to 50 IV NNS Inside containment
[
access hatch
+3 psig personnel airlock at 100.0 l
A-11 A.8 containment 2
60-120
-1.5 in. WG to 17.1 to 50 SC-3/IE Outside surface of high range *I
+3 psig shield wall at 100.0 I
l
- 1hese monitors are quallfled inr post-LOCA environment.
WAPWR-RP 12.3-34 JUNE, 1985 3061e:Id
C O
O O
^
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TA8tE 12.3-3 CANGE AND SEIP0lN15 FOR AREA RADIAi!ON MONITORS 4
Range Sensittwity Initial Setoolnt non' 4r (mP/hr)
(mR/hr)
Intermediate Hieh Control Function Accuracy 3
A-1 control room 10' to 10 10' O.10 mR/hr 0.25 mR/hr No 120 percent of acten1 I
radiation field
}
A-2A,8 containment low range 10'I to 10" 10'I 5.0 mR/h(a) 15.0 mR/h(a) Yes, isolates 120 percent of actual (both) 0.40 R/h(b) 1.0 R/h(b) containment radiation field ventilation system A-3 radiocheJlstry labordtory 10' to 10 10'I 2.0 mR/hr 2.5 mR/hr No 120 percent of actual 4
radiation field A-5 fuel handling building 10'I to 10" 10'I 1.0 mR/hr 2.5 mR/hr No 120 percent of actual radiation field i
A-6 sampling room 10'I to 10" 10'I 2.0 mR/hr 2.5 mR/hr No 120 percent of actual
{
radiation field i
A-1 seal table instrumentation 10'1 to 10 10'I 50 mR/hr 100 mR/hr No 120 percent of actual 4
f room radiation fleid A-g containment access hatch 10'I to 10" 10'I 10 mR/hr 15 mR/hr No 120 percent of actual radiation field 3
U 3
A-11A,3 containment high range 10 to 10 10 2.0 R/hr 3.0 R/hr Yes, initiates 120 percent of actual (both) containment radiation fleid Phase A j
isolation 1
'During refueling operations During power operation WAPWR-RP 12.3-35 JUNE, 1985 3061e:Id
I i
i f
I i'@
4-e
!!O 4
l-c f
f 4
4
)
i s
i FIGURES 12.3-1 THROUGH 12.3-8 (FOLD 0VTS) l..
PROPRIETARY 1
i i
l 4
1 1
A 1
l@
1 i
=
6 j
3 i
l
)
12.4 DOSE ASSESSMENT O
Radiation exposures to operating personnel will be within 10CFR20 limits.
Radiation protection design features described in Section 12.3 of this module, I
and the health physics program supplied in the plant specific applicant's safety analysis report will ensure that the occupational radiation exposures
'(ORE) to operating personnel during operation and anticipated operational concurrences will be as low as reasonably achievable (ALARA).
l
{
Radiation exposures in the plant are primarily due to direct radiation from components and equipment containing radioactive fluidsr-in~ addition, in some plant radiation areas there can be radiation exposure to personnel due to the presence of airborne radionuclides.
Inplant radiation exposyres during normal operation and anticipated operational occurrences are discussed in Subsection 12.4.1 and radiation outside the plant structures are addressed in Subsection j
12.4.2.
12.4.1 Exposures Within the Plant Structures
- O 12.4.1.1 Direct Radiation Dose Estimates p.
NUREG-0713 presents a compilation of occupational radiation exposures (ORE) received annually at domestic connercial nuclear power plants.
From the most II) recent report average collective doses were calculated for domestic plants with Westinghouse-supplied NSSSs.
Figure 12.4-1 illustrates the tr.end in ORE for these plants over the period 1969 through 1982.
It is noted that:
o Specific spikes or peaks in the curve have been generally due to major j
repair operations.
For example, the 1972 peak was due to steam I
generator tube sheet weld repair operations inside the channel heads of two plants.
The manual repair operations resulted in over 500 man-rem alone.
3 i
I l
l WAPWR RP 12.4-1 JUNE,1985 3116e:1d i
. - _. _. -. _,. ~. _. _ _ _ _ - _ _. _. _.. _..
o In general, dips or valleys in the curve were due to new plant start-ups.
For example, the 1973-1975 time period had twelve new plants come on-line.
The low doses at these plants during their first year of operation lowered the average plant dose.
~
o In 1982, the. highest and lowest collective dose was about 1600 and 100, respectively.
In addition, the average collective dose dropped to 650 man-rem.
Factors which may have contributed to this increase in plant collective doses include the followingI }:
o Increasing plant radiation fields o
Required or mandated modifications /back-fits l
o Premature failures of major components 4
o Use of inexperienced workers F-o Management attitude In the analysis of ORE data cumulative man-rem per cumulative MW,-Yr of electricity generated accounts for plant size, and provides a relative measure of costs (man-rem) versus benefits (power production).
Table 12.4-1 provides a cumulative summary (up through 1982) of collective doses for domestic plants with Westinghouse-supplied NSSSs.
Also included on Taole 12.4-1 is a measure of the effective operating time (MW,-Yr/MW ).
As it can be seen f rom Table 12.4-1, cumulative man-rem per MW,-Yr ranges from 0.30 to 2.68.
The best performing plants operate in the range of 0.3 to 0.4 man-rem per Mef,-Y r.
Major factors which have contributed to these excellent performance levels include low plant radiation fields, good layout and access provisions, and excellent operational practices and procedures.
If O
WAPWR-RP 12.4-2 JUNE, 1905 3116e:1d j
it is assumed that the WAPWR achieves the same performance levels as these "best plants," approximately 350 man-rem per year would be achieved [(0.3 man-rem /MW,-Yr)*(1300 MW,)*(0.9 capacity factor)].
However, various design improvementt have been incorporated into the MAPWR plant, and operational improvements can be expected to be incorporated into the plant.
In addition to these improvements, additional ALARA features will be considered for the final MAPWR design.
With these additional plant features, approximately 215 man-rem per year is believed to be a realistic design goal for the annual WAPWR cumulative plant exposure.
Radiation exposure estimates have been made for each of the following work categories:
o Reactor Operations and Surveillance o
P.outine Maintenance o
In-Service Inspection o
Special Maintenance o
Waste Processing o
Refueling P
Table 12.4-2 is a breakdown of the expected annual plant collective doses for each of these work categories.
The values are based on a detailed breakdown of the doses incurred within each category based on feedback obtained f rom operating plants.
Detailed dose predictive models have been developed which identify the various steps which comprise the operation, anticipated radiation levels in the work area, the required number of workers, and the time to perform each step.
A dis.ussion and further breakdown of the collective doses in the above work categories follows.
Reactor Operations and Surveillance O
To support plant operations, the performance of various systems and components must be monitored.
Examples of these activities include the following:
O WAPWR-RP 12.4-3 JUNE, 1985 3116e:1d
o Routine inspections of plant components and systems O
o Unidentified leak checks o
Operation of manual valves
}
o Instrument readings o
Routine health physics ;n trols and surveys o
Decontamination of equipment or plant work areas o
Calibration of electrical or mechanical equipment o
Chemistry sampling and analysis A special concern in this category are those operations performed in contain-ment during power operation.
At-power containment radiation fields are significantly higher than during plant shutdown.
The frequency and duration of at-power containment entries vary greatly from plant to plant and are known to be influenced by the following:
o Plant technical specifications o
Containment design o
Frequency and degree of unscheduled repairs o
Spec 'de utility practices P
Table 12.4-3 provides a list of operations which may require containmen't c: cess during power operation.
Based on limited feedback from U.S. utilities and taking credit for design changes and reliability improvements, no more than 100 worker-hours per year should be required to maintain the plant during
~
at-power operations.
Table 12.4-4 provides a breakdown of the collective doses for reactor operations and surveillance.
Frem Table 12.4-4, the larr est f raction of th?
s I
total dose occurs during routine patrols and inspections of plant systems and components.
Routine Maintenance in order to keep the plant operational, routine maintenance must be performed on all mechanical and electrical components.
Table 12.4-5 provides a MAPWR-RP 12.4-4 JUNE, '985 3116e:1d
~_
breakdown of the collective doses for majcr routine maintenance items.
From
~
Table 12.4-5, the following are noted:
o Maintenance on the reactor coolant and other plant pumps ~(RHR, RCDT, SFP cooling, CVCS charging, etc.) result in the largest f raction of the annual doses.
This is due to the fact that the maintenance j
requires a significant amount of " hands-on" work on the radioactive pump components.
o Valve maintenance contributes a significant f raction to the annual doses.
This can be explained by:
- 1) the large number of valves in a a
plant; 2) the buildup of activity on the inside surfaces of valves; i
and 3) the f requent operation of particular valves.
o Steam generator sludge lance and secondary side inspection results in i
relatively low collective doses.
To maintain the RCPs, six different maintenance / inspection operations are performed.
Table 12.4-6 provides a description and schedule of these opera-tions.
An eighteen month fuel cycle and a cartridge style seal have been P
assumed.
Table 12.4-7 presents a dose estimate for each of the maintenanc'e inspection operations.
Table 12.4-8 summarizes the doses incurred for all pump maintenance and inspection operations.
From Table 12.4-8, approximately 2.3 man-rem per year per pump is required.
Note that in-service inspection operations have been considered an integral part of the maintenance / inspection
~
operations (sequences B through F).
In-Service Inspection Section XI of the ASME Boiler and Pressure Vessel Code requires both a pre-service and periodic in-service inspection (ISI) be performed on all plant safety class components (3)
The Code also addresses component accessibility and inspectability and the extent of and methods employed for such examinations, i
)
i WAPWR-RP 12.4-5 JUNE,1985 3116e:1d 4
~~,--,-,,,----,---~,w,-
-n
.,n_n.,--..
,,...-,._,---..,--,-,-,,-__,---n------,n.--,,.,,
The MAPWR design has features which permit complete compliance with the Code requirements. Examples of these design features include the following:
o All reactor internals are completely removatle from _the reactor vessel, allowing access to the entire inside surface of the vessel for O
inspection.
The tools and storage space required to permit removal of the internals are also provided, o
The reactor vessel closure head is stored in a dry condition on the operating deck during refueling, allowing direct access for inspection, o
All reactor vessel studs, nuts, and washers are removed to dry storage during refueling, allowing inspection in parallel with refueling operations.
i i
o Removable plugs are provided in the - primary shield just above.the reactor vessel primary coolant nozzles for inspection of the welds joining the nozzles to the safe-ends and the welds joining the ON safe-ends to the primary coolant piping.
Readily-removable insulation is provider' ver these weld areas.
V
~
o Manways are provided in the steam generator channel head and in the moisture separator section for access for internal inspection.
o A manway is provided in the pressurizer top head for access for internal inspection.
l o
The insulation covering all component and piping welds and adjacent O-base metal is designed for ease of removal and replacement in areas where external inspection is planned, o
Openings are provided in the operating deck concrete shielding above the main coolant pumps to permit removal of the pump motor for internal inspection access to the pumps.
O MAPWR-RP 12.4-6 JUNE,1985 3116e:1d
_=
i o
The primary loop compartments are designed to allow personnel entry during refueling operations, with shielding provided between major components, to permit direct inspection access to the external portion of piping and components.
~
In addition to these features, sufficient space is provided around the various components to permit access by the examiners and their equipment.
Allowances for component disassembly and insulation removal / installation are provided.
Where feasible, permanent platforms or scaffolding provisions, service lines, ventilation systems, and handling features are provided to improve access.
Table 12.4-9 provides a detail breakdown of the collective doses for in-service inspection.
Although some inspections are performed over varying time intervals, as shown in Table 12.4-10, the dose estimates have been annualized over a ten-year inspection period.
Special Maintenance To support plant operation, repairs will be required on various systems and components.
Since much of this work is performed on a non-routine basis, it P
has been included as part of the special maintenance category.
Examples off such activities include the following:
o Installation of new plant systems and components (R.V.
level indication system, R.V. head vent, snubbers, etc.)
o Modification of existing equipment o
Repairs of components which have failed prematurely I
Due to the nature of this work, f requency of repairs are very dif ficult to estimate.
However, Table 12.4-11 provides a best estimate of the annual dose breakdown for special maintenance.
From Table 12.4-11, the major contributors to the total doses are plant modifications, steam generator inspection /
WAPWR-RP 12.4-7
. LUNE, 1985 3116e:1d
._.~
repairs, and pump and valve repairs.
In general, these repairs are all labor O
intensive and/or performed in a high radiation environment.
V At most operating plants, special maintenance on the steam generators have resulted in significant personnel doses.
Even tnough steam generator doses at new plants are expected to benefit from both design changes and improved secondary side chemistry, periodic primary to secondary tube leaks are assumed to occur.
As a result, the following steam generator primary side special maintenance is assumed to be performed:
O o
Mechanical tube plugging of three tubes per steam generator every
- outage, o
A three percent eddy current tube inspection of each steam generator every outage (first outage excluded).
Waste Processing During various plant operations, liquid, gaseous, and solid waste products are generated.
Examples of such waste include the following:
p.
o Filter cartridges o
Demineralizer resins o
Tank sludges o
Evaporator bottoms o
Dry active wastes Since these wastes are generally radioactive, appropriate disposal practices and procedures must be followed.
A breakdown of the annual waste processing collective doses is presented in Table 12.4-12.
Refueling For the WAPWR, an eighteen month interval is assumed.
The refueling process is labor intensive and detailed planning and coordination is essential in O
WAPWR-RP 12.4-8 JUNE, 1985 3116e:1d 1
~ ___________
order to maintain personnel doses ALARA.
In addition, the incorporation of advance technology into the refueling process will lead to reduced personnel
~~
doses.
Table 12.4-13 provides a dose estimate for refueling.
From Table 12.4-13, the majority of the annual doses occur during reactor disassembly and reassembly.
These estimates ate based on the incorporation of improved refueling equipment and procedures as noted in Section 12.3.1.
Since the WAPWR is designed for eighteen mo.ith fuel cycles the annual average refueling dose is approximately fourteen man-rem.
12.4.1.2 Airborne Radioactivity Dose Estimates Due to leakages of radioactive fluids into the auxiliary, containment, radwaste, f u
d1ing, and turbine buildings, plant personnel are exposed to radionuclides aased into the atmosphere of these buildings by the leaked fluids.
These atmospheric contaminants contribute to the total body, thyroid, and lung doses.
The peak airborne concentrations for most areas in the plant are within the P
limits specified in 10 CFR 20.
By use of appropriate respiratory equipmerit l
and/or limitation of occupancy time, personnel are allowed to enter areas where the airborne activity levels exceed 10 CFR 20 limits.
Refer to the plant specific applicant's safety analysis report for' a disi:ussion of the annual airborne radiation exposures.
12.4.2 Radiation Exposure Outside the Plant Structures Refer to the plant specific applicant's safety analysis report for a discussion of radiation exposures outside the plant structures.
O O
WAPWR-RP 12.4-9 JUNE,1985 3116e:1d
=_
12.4.3 References 1.
B.G. Brooks.
" Occupational Radiation Exposure at Comerical Nuclear Power Reactors 1982."
Washington, DC:
U.S.
Nuclear Regulatory.. Comission, NUREG-0713, Annual Report, December 1983.
O 2.
" Actions Being Taken to Help Reduce Occupational Radiation Exposure at Commercial Nuclear Power Plants." GA0/EMD-82-91, August 24, 1982.
3.
" Rules for In-Service Inspection of Nuclear Power Components."
New York, NY:
American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section XI, 1980.
i O
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' O WAPWR-RP 12.4-10 JUNE, 1985 3116e:1d i
4
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-n.__,,,nn,_-,.__,,n,
_,._-,.-n---_
TABLE 12.4-1 COLLECTIVE DOSE
SUMMARY
FOR DOMESTIC WESTINGHOUSE PLANTS 4
Station Cumulative Summary Stations MW, MW,-Yr/MW, Man-Rem /MW,-Yr l
Four-Loop Plants Maximum 1180 12.17 1.75 Minimum 175
.44
.32 Average 943 4.66
.84 Three-Loop Plants Mayimum 890 9.13 2.68 Minimum 436 1.83
.73 Average 731 5.31 1.60 Two-Loop Plants i~
Maximum 512 8.82 1.97 Minimum 470 6.84 0.30 Average 496 7.89 0.81 Overall Average 777 5.6 1.1 2
i
- O a
Source: NUREG-0713, Volume 4 4
hstartup through 1982 4
4 O
WAPWR-RP 12.4-11
.lVNE, 1985 j
3116e:1d
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_ w y,,e w p mm m--w-w m y - o
---,p
-ww.w e-
TABLE 12.4-2 COLLECTIVE DOSE BREAKDOWN
~
^
i Percent Annual Dose Catecory (5)
(man-rem)
Reactor Operations and Surveillance 19 40 Routine Maintenance 26 56 In-Service Inspection 13 28 Special Maintenance 31 66 Waste Processing 5
11 Refueling 6
14 p.
100 215 i
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WAPWR-RP 12.4-12 JUNE, 1985 l
3116e:1d 1
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TABLE 12.4-3 O
OPERATION 5WHICHMAYREQUIRECONTAINMENTAT-POWERACCES5 SURVEILLANCE Routine Patrols Health Physics Surveys Patrols to Identify System Leaks MECHANICAL REPAIRS Isolate System / Component Leaks Valve Adjustments / Repairs Leak Test Personnel Hatch Fan Cooler Repairs RCP Oil Additions Repair In-Core Detector Drive System Repair / Replace Radiation Mo11toring Detectors 7
ELECTRICAL REPAIRS l
Repair / Replace Transmitters f-In-Core Thermocouple System Fepairs at Junction Box Valve Operator Repairs t
O l
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I WAPWR-RP 12.4-13 JUNE,1985 3116e:1d i
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TABLE 12.4-4
~
DOSE ESTIMATE FOR REACTOR OPERATIONS AND SURVEILLANCE Annual Dose t
Work Description (man-rem) i l
OPERATION SUPERVISION Routine Patrols and Inspections 15 f
valve Line-Ups (Manual) 2 System Flushing and Testing 1
i.
I HEALTH PHYSICS Job Coverage 5
Routine Surveys 5
t DECONTAMINATION Equipment and Work Areas 7
i 0-CALIBRATION Transmitters, Survey Instruments, Radiation Monitors, etc.
3 CHEMISTRY Sampling 2
O 1
i' TOTAL COLLECTIVE DOSE:
40 O
i O
WAPWR-RP 12.4-14 JUNE, 1985 3116e:1d
TABLE 12.4-5
"~
DOSE ESTIMATE FOR ROUTINE MAINTENANCE Annual Dose Work Description (man-rem)
Valve Adjustment / Repacking 12.0
{
Pump Overhaul 12.0 RCP Seal Maintenance / Inspection (4) 9.2 SG Sludge Lance (4) 2.1 Domineralizer Resin Change-out 6.0 1
(
Filter Replacement 5.0 P
Calibrate / Repair Electrical Components 4.0 i
Miscellaneous Work 2.5
- SG Secondary Side Inspection (4) 3.7 r
TOTAL COLLECTIVE DOSE:
56.5 O
l l
WAPWR-RP 12.4-15 JUNE, 1985 l
3116e:1d l
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TABLE 12.4-6 bV REACTOR COOLANT PUMP MAINTENANCE / INSPECTION SCHEDULE ~
5 Reauired Secuencea Months Since Years Since Pump Pump Pump Pump Outage Start-up Start-up 1
2 3
4 1
18 1.5 A
B A
B O,
2 36 3.0 C
A B
A 3
54 4.5 E
C E
B 4
72 6.0 B
E C
E 4
5 90 7.5 A
B A
C 6
108 9.0 F
A D
A 7
126 10.5 A
D A
D B
144 12.0 B
A B
A 9
162 13.5 E
C E
B 10 180 15.0 B
E C
E 11 19B 16.5 A
B A
C 12 216 18.0 0
A 0
A 13 234 19.5 A
F A
D 14 252 21.0 C
A B
A 15 270 22.5 E
B E
B 4
16 288 24.0 8
E C
E 17 306 25.5 A
B A
C 18 324 27.0 0
A F
A P
l 19 342 28.5 A
D A
D 20 360 30.0 C
A B
A 21 378 31.5 E
C E
B 22 396 33.0 8
E B
E 23 414 34.5 A
B A
C 24 432 36.0 0
A D
A 25 450 37.5 A
D A
B 26 468 39.0 C
A B
A aihe following is a description of the work operations:
A - Pump and motor inspections; R - Motor inspection with minor ISI; C - Pump and motor inspections with minor ISI; D - Motor inspection with major ISI (flywheel);
l E - Pump and motor inspection with major ISI (rotor / stator);
e F - Pump and motor inspection with minor ISI.
O WAPWR-RP 12.4-16
.1UNE, 1985 3116e:1d
[
r l
TABLE 12.4-7 EXPOSURE ESTIMATE FOR RCP MA'NTENANCE/ INSPECTION i
1 Work Time
- Dose Rate **
Dose I
Work Description (Man-Hrs)
(Rem /Hr)
(Man-Rem)
A - Pump and Motor Inspections 156
.022 3.4 8 - Motor Inspection With Minor ISI 129
.016 2.1 C - Pump and Motor Inspections With 4
Minor ISI 125
.030 3.7 D - Motor Inspection With Major ISI (Flywheel) 174
.016 2.8 E - Pump and Motor Inspection With Major ISI (Rotor / Stator) 387
.0155 6.0 p.
F - Pump and Motor Inspection With Minor ISI 181
.020 3.7 f
- Represents work time in significant radiation fields.
)
- Effective dose rate over the range of work area dose rates.
i O LO WAPWR-PP 12.4-17 JUNE, 1985 3116e:1d
l TA8tE 12.4-8 REACTOR COOLANT PUMP DOSE
SUMMARY
l Required Dose / Inspection Total Dose' a
Seauence Inspections" (man-rem)
(man-rem)
A 36 3.4 122.4 8
23 2.1 48.3 C
14 3.7 51.8 l
D 12 2.8 33.6 1
i E
16 6.0 96.0 i
l l
F 3
3.7 11.1 V
1 TOTAL COLLECTIVE 00SE':
363.2 ANNUAL COLLECTIVE DOSE:
2.3/RCP
" Refer to Table 12.4-6 for inspection description and frequency.
bRefer to Tables 12.4-7.
'Value is for four pumps.
O O
WAPWR-RP 12.4 18 JUNE,.1985 3116e:1d
TABLE 12.4-9 DOSE ESTIMATE FOR IN-SERVICE INSPECTION 1
O Annual Dose Corr.conent (man-remL Valves Bodies and Boltings 10.0 SG Primary Side Inspections 4.1 Reactor Vessel and Head 3.8 Other Piping 3.5 Reactor Coolant Loop Piping and Supports 1.B SG Shell 1.7
+-
Heat Exchanger Shells 1.3 Pressurizer Shell 1.2 Pump Housings and Supports 0.4 l
Tank Shells and Supports 0.3 l
Filter Housings and Supports 0.1 1
TOTAL COLLECTIVE DOSE: 28 i
l O
WAPWR-RP 12.4-19 JUNE, 1985 3116e:1d
k TABLE 12.4-10 EXPOSURE ESTIMATES FOR VARIOUS INSPECTION ACTIVITIES -
4 Average Work Time Dose Rate Dose Activity (Man-Hrs)
(Rem /Hr)
Freauency (Man-Rem) i Steam Generator 13
.04 Each refueling 0.5/SG Secondary Side outage Inspection i
.07 Each refueling 5.5/SG Eddy Current outage Inspe: tion Reactor Vessel O
In-Service i
Inspection e
10-Year Inspections 1127
.016 Once Per 10 Years 18.5 d
40-Month Inspection 619
.016 Twice Per 10 Years 10.0 O
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HAPWR-RP 12.4-20 JUNE, 1985 3116e:1d
TA8LE 12.4-11 DOSE ESTIMATE FOR SPECIAL MAINTENANCE O
Annual Dose Work Description (man-rem)
Plant Upgrades / Modifications 15.0 O
Valve Repairs 11.0 SG Primary Side Inspection (3) 10.7 Pump Repairs 9.5 Electrical Repairs 4.0 RCP Repairs and Inspection (4) 4.0 P
SG Tube Plugging (4) 5.6 Repairs to Tanks, Heat Exchangers, 2.5 Piping, etc.
SG Secondary Side Repairs (4) 1.4 Pressurizer Repairs 1.2 CRDM Repairs 0.8 O
TOTAL COLLECTIVE DOSE:
66 O
WAPWR-RP 12.4-21 JUNE,1985 3116e:1d
~
l l
TABLE 12.4-12 I
I DOSE ESTIMATE FOR WASTE PROCESSING
-l l-Annual Dose i
Work Description (man-rem)
Radioactive Waste Handling 5
i i
System Adjustments / Repairs 4
f f
System Operation (Sampling, Valve 1
i i
Adjustments, Monitoring,etc.)
i i
f Laundry Operations 1
i i
TOTAL COLLECTIVE DOSE:
11 9~
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7 i
t 1
1 i
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l WAPWR-RP 12.4-22 JUNE,1985
)
3116e:1d
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L..
I TABLE 12.4-13 DOSE ESTIMATE FOR REFUELING
~l Worktime Dose Rate Dose Work Description (man-hrs)
(rem /hr)
(man-rem) i 1
Refueling Operations (16 MAN-REM) o Preparation 158
.005 0.8 o
Reactor Disassembly 208
.019 3.9 o
Fuel Shuffle 362
.006 2.2 o
Reactor Reassembly 272
.025 6.8 o
Clean-Up 258
.009 2.3 Miscellaneous SFP Building Work 1000
.002 2.0 Equipment Checks and Repairs 1000
.002 2.0
}
9-TOTAL COLLECTIVE DOSE:
20 i
i ANNUAL DOSE:
14 ilO i
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12.5 HEALTH PHYSICS PROGRAM
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Refer to the plant specific applicant's safety analysis report for a i
discussion of the applicant's health physics program.
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