ML20151N432
| ML20151N432 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 07/31/1988 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19292J204 | List: |
| References | |
| RESAR-SP-90-01, RESAR-SP-90-1, RESAR-SP-90-A02, RESAR-SP-90-A2, NUDOCS 8808080215 | |
| Download: ML20151N432 (26) | |
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l' AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 11
- RADIATION PROTECTION i
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' O WEC P.O. Bsx 355 Pittsburgh, PA 15230 0 WAPWR-RP 8004e:1d AMENDMENT 2 JULY, 1988 8808080215 PDR eso727 .
ADOCK 05000601 l IN PDC - i
AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 11 RADIATION PROTECTION INSTRUCTION SHEET Place page 430-1 (Question / Response) after Amendment 1, in the Questions / Answers section to Module 11.
The following pages are submitted as part of Amendment 2 to Module 11. The text revisions include editorial changes and corrections to original text as a result of Westinghouse review of Staff draft SER dated June 10, 1988.
Replace current page vi with revised page vi.
Replace current page 12.2-26 with revised page 12.2-26.
Replace current page 12.3-19 with revised page 12.3-19.
Replace current page 12.3-?3 with revised page 12.3-23.
Replace current page 12.3-26 with revised page 12.3-26.
Replace current page 12.3-30 with revised page 12.3-30.
Replace current page 12.3-31 with revised page 12.3-31.
Replace current page 12.3-35 with revised page 12.3-35.
O O WAPWR-RP AMENDMENT 2 B004e:1d JULY, 1988
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AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 11 RADIATION PROTECTION INSTRUCTION SHEET - (cont'd)
The following pages were formally transmitted in Amendment 1 to Mcdule 11 (NS-NRC-86-3110, dated March 14,1986) in support of Westinghouse response to staff questions. They are now being submitted as Module 11 text revisions.
Replace current page 12.2-1 with revised page 12.2-1.
Replace current page 12.2-2 with revised page 12.2-2.
Replace current page 12.2-13 with revised page 12.2-13.
Replace current page 12.2-14 with revised page 12.2-14.
Replace current page 12.2-16 with revised page 12.2-16.
Replace current page 12.2-17 with revised page 12.2-17.
Replace current page 12.2-54 with revised page 12.2-54. j 1
Replace current page 12.3-5 with revised page 12.3-5.
O Replace current page 12.3-12 with revised page 12.3-12.
R*;iace current page 12.4-2 with revised page 12.4-2.
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.o WAPWR-RP AMENDMENT 2 5004e:1d JULY, 1988
REQUEST FOR ADDITIONAL INFORMATION WEST!NGH0CSE ADVANCED PRESSURIZED WATER REACTOR (RESAR SP-90)
DOCKET NO. 50-601 The following Question /Respsnse was formally transmitted in Addendum 4 to RESAR-SP/90 PDA in Westinghouse letter NS-NRC-88-3341, dated May 13, 1988.
430.18 Provide a discussion which indicates how TMI Action Fian Item "Additional Accident Monitoring Instrumentation,"
O II.F.1, recuirements have been met.
RESPONSE
The radiation and airborne radioactivity monitors for accident conditions, required to meet TMI action plan item II.F.1, are included in Subsection 11.5.5 - Post Accident Radiation Monitor-ing (PAMS) of RESAR-SP/90 PDA Module 12, "Waste Management," and Subse: tion 12.3.4 -
Area Radiation and Airborna Radioactivity Monitoring Instrumentation of RESAR-SP/90 PDA Module 11 "Radiation Protection."
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O WAPWR-RF 430-1 AMENDMENT 2 B004e:1d JULY, 1988 1
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l TABLE OF CONTENTS (Continued)
LIST OF TABLES Number Title FaSe 1.6-1 Material Incorporated by Reference 1.6-2 1.8-1 Standard Review Plan Deviations 1.8-2 12.2-1 Radiation Sources: Reactor Coolant Nitrogen-16 12.2-17 Activity 12.2-2 Radiation Sources: Pressurizer 12.2-18 12.2-3 Isotopic Composition and Specific Activity of 12.2-20 Typical Out-of-Core Crud Deposits 12.2-4 Radiation Sources: Chemical .nd Volume Centrol System 12.2-21 12.2-5 Radiation Sources: Beton Recycle System 12.2-27 12.2-6 Radiation Sources: Liquid Waste Processing System 12.2-32 12.2-7 Radiation Sources: Gaseous Waste Processing System 12.2-38 12.2-8 Radiation Sources: Solid Waste Processing System 12.2-40 12.2-9 Radiation Sources: Spent Fuel Pit Cooling System 12.2-41 12.2-10 Radiation 3ources: Stemo Sanarator Biowdown 12.2-42 Processing System 12.2-11 Radiation Sources: Residual Heat Removal System 12.2-45 l 2 12.2-12 Core Average Game Ray Source Strengths 12.2-46 12.2-13 Spent Fuel Gamma Ray Source Strengths 12.2-47 12.2-14 Core Average and Spent Fusi Neutrc*n Source Strengths 12.2-48 12.2-15 Irrsdiated Sb-Be Secondary Source Rod Source 12.2-49 l
Strengths 12.2-16 Irradiated Incore Detector and Drive Cable Maximum 12.2-51
, Withdrawal Source Strengths 12.2-17 Irradiated Incore Detector Drive Cable Source 12.2-52 Strengths 12.2-18 Irradiated Inconel 6D0 Flux Thimble Source Strengths 12.2-53 l l
O WAPWR-RP vi. AMENDMENT 2 I B004e:1d JULY, 1988
e TABLE 12.2-4 (Sheet 6 of 6)
O RADIATION SOURCES - CHEMICAL AND VOLUME CONTROL SYSTEM O Notes:
(a) The letdown coolant volume is plant layout dependent.
(b) Includes 80 key xenon-133.
(c) These sources correspond to a nominal operating level in the tank of 3
] ft in the vapor phase and ( 3
) ft in the liquid phase.
a,c 2
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(d) Homogeneous sources with the following dimensions and compositions:
Source Dimensions Inches Source Composition Filter Radius Length (Volume Percent)
Reactor coolant 3.375 19 67% air, 33% water Seal water return 3.375 19 67% air, 33% water Seal water injection 1.375 21 11% air, 89% water Boric acid 3.375 19 67% air 33% water Boric acid polishing 3.375 19 67% air, 33% water i
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' O WAPWR-RP 12.2-26 AMENDMENT 2 5004e:1d JULY, 1988 i
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e pool. This limits the dose at the water surface to less than 2.5 mrem /hr for i an assembly in a vertical position at the maximum elevation. Normal water depth above the stored assemblies in the spent fuel pit is greater thsn 8 meters and the dose rate at the pool surface is significantly less 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 mrom/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 requirements of adjacent areas. Equipment in the SFPCS, which is described in 2 Subsection 9.1.3 of RESAR-SP/90 PDA Module 13, "Auxiliary Systems," to be )
shielded includes the SFPCS heat exchangers, pumps, and piping. l I
1 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 Radiation shielding is not required for process equipment located in the turbine building.
12.3.2.2.7 Control Room Shielding Design O The design basis . loss-of-coolant accident (LOCA) dictates the shielding re-cuirements for the control room. Consideration is given to shielding provided by the containment structure. Shielding combined with other engineered safety features is provided to permit access and occupancy of the control room fol-lowing a postulated LOCA, so that radiation doses are limited to 5 rem whole body from contributing modes of exposure for the duration of ':he accident, in accordance with 10 CFR 50, Appendix A, General Design Criterion 19.
O WAPWR-RP 12.3-19 AMENDMENT 2 5004e: 1d JULY, 19CS
12.3.3 Ventilation !
O 12.3.3.1 Design Objectives
'The plant ventilation systems, in addition to their primary function of 2 preventing extreme thermal environmental conditions for cperating 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 the limits of 10CFR20, Appendix B, 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 I 10CFR20 and 50 and in fact will be considerably smaller since average coolant inventorie. 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 Management".
The control room ventilation system will also operate to provide a suitable environment for equipment and continuous personnel occupancy in the control j room under post-accident conditions in accordance with 10CFR50, Appendix 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 fecility, and on the plant site wtere personnel, construction wo.kers, or site v isi;; ors 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.
WAPWR-RP 12.3-23 AMENDMENT 2 B004e:1d Jul.Y, 1988
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C. To provide local and remote indication of ambient gamma 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.
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.
O By meeting the above objectives, the area radiation monitoring system aids health physics personnel in keeping radiation exposures as low as reasonably achievable (ALARA).
The design objectives of the ARMS during postulated accidents are:
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A. To provide the capability to alarm and initiate a containment l 2 l ventilation isolation signal in the event of a loss-of-coolant l
accident (LOCA), fuel handling accident inside containment, or abnormally high radiation inside the containment (monitors A-2A&B). 2 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"). ,
l 12.3.4.1.2 Criteria for Location of Area Monitors Considerations for area monitor locations are based on the following:
A. Frequency and length of personnel occupancy of a specific area.
B. Potential for personnel to unknowingly receive high radiation doses.
C. Potential for eqaipment malfunction.
WAPWR-RP 12.3-26 AMENDMENT 2 B004e:1d JULY, 1988 a
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E. Sampling Room Area Monitor A-6 To continuously indicate the radiation levels in the sampling room. A high radiation alarm signal warns the occupants of the sampling room of a deteriorated radiological condition.
F. Seal Table Instrumentation 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 alarm signal warns occupants of the seal table room of a deteriorated radiological condition.
G. Containment Access Hatch Area Monitor A-9 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-11A and A-11B To indicate, along with A-2A and A-2B, the radiation levels inside the containment building at the operat ng deck following a design basis accident. These monitors are in compliance with NUREG-0737, Item 1&2 11.F.1.3.
12.3.4.1.9 Range and Alarm Setpoints 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 modificatic' as plant operatincj experience is developed.
Radiation zones are described in Table 12.3-1.
O WAPWR-RP 12.3-30 AMENDMENT 2 B004e:1d JULY, 1988
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d 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-11A&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, intermediate and high. (See Table 12.3-3.) If a monitor has a control function, i.e., A-2A&B, the control 2 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 alarm gives both a visual a..d 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".
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O WAPWR-RP 12.3-31 AMENDHENT 2 5004e:1d JULY, 1988
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l TABLE 12.3-3 RANGE AND SETPOINTS FOR AREA RADIATION MONITORS Range Sensitivity Initial Setpoint Monitor (mR/hr), (mR/hr) Intermediate High Control Function Accuracy A-1 control ro se 10~ to 10 10~ O.10 mR/he 0.25 mR/hr No +20 percent of actual 1
radiation field
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f A-IA,8 containment low r<nge 10 ' to 10# 10 ' 5.0 mR/h(a) 15.0 mR/h(a) Yes, isolates +20 percent of actual 1
i (both) 0.40 R/h(b) 1.0 R/h(b) containment radiation field j ventilation I system 1
! A-3 radiochemistry tabncatory 10 to 10 10 2.0 mR/hr 2.5 mR/hr No +20 percent of actual l
j radiation field
~
A-5 fuel handling building 10 to 10# 10 ' 1.0 mR/hr 2.5 mR/hr No *2O percent of actual radiation field A-6 saan11ng room 10 to 10* 10 2.0 mR/4 2.5 mR/W No +20 percent of actual radiation field A-7 seal table instrue ntation 10 to 10# 10
~I 50 mR/hr 100 mR/hr No +20 percent of actual j room radiation field A-9 containment access hatch 10~ to 10 10 10 mR/hr 15 mR/hr No 120 percent of actual I radiation field
,}~
A-11A,8 containment high range 10 to 10 10 3.0 R/hr 100.0 R/hr No +20 percent of actual g
1 (both) radiation field I
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- During refue11pg or* rations
' b During power operation i
W 12.3-35 AMENOMENT 2
, _APWR-RP 8004e:1d sFJtY, 1988 l
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12.2 RADIATION SOURCES O This section discusses and identifies the sources of radiation that form the basis for shield design calculations and t.ie sources of airborne radioactivity used for the design of personnel protection measures and dose assessment.
O 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 I 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 thermal 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 based on a realistic (or expected) model are given in Section 11.1 of RESAR-SP/90 PDA Module 12, "Waste 1 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. I O 2. Gamma ray energy fluxes at the inside surface of the primary shield concrete at the core midplane.
O WAPWR-RP 12.2-1 AMENDMENT 1 {
5004e:1d FEBRUARY, 1986 l l
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- 3. Gamma ray dose rates at the inside surface of the primary shield concrete.
- 4. Detailed angular distributions of radiation leakage (neutron and gamma ray) from the reactor pressure vessel for streaming analyses. l O The nitrogen-16 activity of the coolant (produced from oxygen activation) is the controlling radiation source in the design of the secondary shield and is j 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 tianagement". The fission and corrosion product source strengths in the reactor coolant pressuri::er liquid and vapor 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-affinity-material areas, and possibly thin-boundary-layer regions.
Systems which process or contain reactor cooiant also contain radiation sources during full power operation. These systems include the chamical and 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 raciation sources for the CVCS, specifically delineating the sources fer:
O 1. CVCS letdown stream.
- 2. Mixed bed demineralizers.
- 3. Cation bed demineralizer.
O WAPWR-RP 12.2-2 AMENDMENT 1 !
B004e:1d FEBRUARY, 1986 i 1
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irradiation period of 400 days. Irradiated incore flux thimble gamma 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 The radiation sources of importance for the design basis accident are the containment source and the residual heat removal system source.
1 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 l NUREG-0737. The intcgrated gamma ray and beta particle source strengths for O various time periods, of fission products released from the fuel to the 1 containment following the postulated accident, are given in Table 12.2-19.
The post-accider,t recirculation system and shielding should be designed to 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 loop.
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 icep and associated equipment are given in Table 12.2-20.
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] Shielding design for vital areas is based on radiation sources specified in NUREG-0737,Section II.B.2, and the design dose rates for personnel in vital i l areas requiring continuous or infrequent access, will be such that the i
guidelines of GDC-19 will not be exceeded during the course of the accident.
l WAPWR-RP 12.2-13 AMENDMENT 1 l 5004e:1d FEBRUARY, 1986 l l
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Isotopic fission product sources from the maximum credible accident, based on the assumptions in TID-14844, are given in Chapter 15 of RESAR-SP/90 PDA 1 Module 4, "Reactor Coolant System".
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'12.2.2 Airborne Radioactive Material Sources l O
Sources of airborne radioactive material in equipment cubicles, corridors, or operating areas normally occupied by operating personnel from systems and components described in RESAR-SP/90 may be obtained from the reactor coolant activities given in Section 11.1 of RESAR-SP/90 PDA Module 12 "Waste Management".
Sources resulting from the removal of the reactor vessel head and the movement of spent fuel are dependent on a number of operating characteristics (e.g.,
coolant chemistry, fuel performance, etc.) and operating procedures followed during and after 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 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 surface to lets than 2.5 millirem per hour.
12.2.2.1 Model for Calculating Airborne Concentrations For those regions which are characterized by a constant leakrate of the radioactive source at constant source strength and a constant exhaust rate of 4
WAPWR-RP 12.2-14 AMENDMENT 1 5004e:1d FEBRUARY, 1986
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. l airborne concentration of the i th radioisotope at ' time t Cg (t) =
in vCi/cm3 in the applicable region From the above equation, it is evident that the peak or equilibrium concentration, C E i, f the i th radioisotope in the applicable region will be given by the f 11owing expression:
CEqi = (LR)$ Ag (PF)$/M Ti With high exhaust rates, this peak concentration will be reached within a few hours.
12.
2.3 REFERENCES
- 1. Lutz, R. J. , and Chubb, W., "Iodine Spikirig -
Cause and Effect," ANS Transactions, Vol. 28, pg. 649, June, 1978.
- 2. DiNunno, J. J., et al, "Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, March,1962.
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O WAPWR-RP 12.2-16 AMENDMENT 1
- 5004e
- 1d FEBRUARY, 1986 i
TABLE 12.2-1 O RADIATION SOURCES ,
REACTOR COOLANT NITROGEN-16 ACTIVITY Loop Nitrogen-16 Transit Activity Position in loop Time (sec) (microcuries/gm)
Leaving core 0.0 163 Leaving reactor vessel 3.3 118 Entering steam generator 3.8 112 Leaving steam generator 8.9 68 Entering reactor coolant pump 9.6 64 Entering reactor vessel 10.7 57 4 Entering core 12.3 53 1 Leaving Core 13.2 163 O
Nitrogen-16 Energy Emission Energy Intensity (Mev/ gamma) (Percent) 1.75 0.13 2.74 0.76 6.13 69.0 7.12 5.0 O
1 O WAPWR-RP 12.2-17 AMENDMENT 1 B004e:1d FEBRUARY, 1986
TABLE 12.2-19 INTEGRATED GAMMA RAY AND BETA SOURCE STRENGTHS AT VARIOUS TIMES FOLLOWING A MAXIMUM CREDIBLE ACCir!T (TID-14844 Release Fractions)
O Source Strength at Timo,After Release (Mev/ watt)
Energy Group (Mev/ gamma) 0.5 Hour
~
1 Hour 2 Hours 8 Hours 1 Day O 0.2 - 0.4 1.3 x 10 12 2.3 x 10 12 4.0 x 10 12 1,3 x 3g 13 3.3 x 10 13 13 0.4 - 0.9 8.4 x 10 12 1.5 x 10 13 2.4 x 10 13 5.3 x 10 8.6 x 10 13 0.9 - 1.35 3.7 x 10 12 6.4 x 10 12 1.1 x 10 13 2.5 x 10 13 3.8 x 10 13 1.35 - 1.8 3.6 x 10 12 6.3 x 10 12 1.0 x 10 13 2.1 x 10 13 3.0 x 10 13 1.8 - 2.2 1.9 x 10 12 3.2 x 10 12 5.2 x 10 12 1.1 x 10 13 1.3 x 10 13 2.2 - 2.6 2.0 x 10 12 3.7 x 10 12 6.1 x 10 12 1.2 x 10 13 1.4 x 10 13 2.6 - 3.0 3.4 x 10 11 5.5 x 10 11 8.4 x 10 11 1.3 x 10 12 1.4 x 10 12 3.0 - 4.0 3.6 x 10 11 4.6 x 10 11 5.8 x 10 11 8.0 x 10 11 8.3 x 10 11 O 4.0-5.0 5.0 - 6.0 1.6x10f1 1.1 x 10 20 1.6 x 10 11 1.1 x 10 10 1.7 x 10 11 1.1 x 10 10 2.0 x 10 11 1.1 x 10 10 2.0 x 10 11 1.1 x 10 10 Beta 1.3 x 10 13 2.2 x 10 13 3.6 x 10 13 8.9 x 10 13 1.5 x 10 14 1 Week 1 Month 6 Months 1 Year 0.2 - 0.4 1.3 x 10 14 2.3 x 10 14 2.6 x 10 14 2.6 x 10 14 0.4 - 0.9 1.7 x 10 14 2.7 x 10 14 5.3 x 10 14 6.4 x 10 14 1 0.9 - 1.35 4.8 x 10 13 5.3 x 10 13 6.0 x 10 13 6.4 x 10 13 1.35 - 1.8 4.6 x 10 13 7.2 x 10 13 8.6 x 10 13 8.9 x 10 13 1.8 - 2.2 1.4 x 10 13 1.5 x 10 13 1.8 x 10 13 2.0 x 10 13 2.2 - 2.6 1.5 x 10 13 1.7 x 10 13 1.8 x 10 13 1.8 x 10 13 2.6 - 3.0 1.4 x 10 12 1.5 x 10 12 1.5 x 10 12 1.5 x 10 12 3.0 - 4.0 8.4 x 10 11 8.5 x 10 11 8.5 x 10 11 S.5 x 10 11 4.0 - 5.0 2.0 x 10 11 2.0 x 10 11 2.0 x 10 11 2.0 x 10 U 5.0 - 6.0 1.1 x 10 10 1.1 x 10 10 1.1 x 10 10 1.1 x 10 10 Beta 3.9 x 10 14 6.5 x 10 14 1.1 x 10 15 1.4 x 10 15 WAPWR-RP 12.2-54 AMENDMENT 1 3004e:1d FEBRUARY, 1986
evaporator components are separated from those that are less radioactive. Instruments and controls are located in accessible low background radiation areas.
-D. Pumps for pumps containing high level radiation sources, means are provided 3 j for pump drainage prior to servicing. 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 cor. trol any contamination within plant structures. For tanks outside structures, which can potentially contain radicar.tive fluids, dikes are used to contain overflows.
I F. Heat Exchangers i I
Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials to minimize leakage. l Impact baffles are provided, and tube side and shell side velocities l
- are limited to minimize erosive effects. Wherever possible, the radioactive fluid passes through the tube side of the heat exchanger.
G. Instruments Instrt.m nt devices are located in low radiation zones and away from 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 AMENDHENT 1 5004e:1d FEBRUARY, 1986
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1 l
. to component crud traps or radiation streaming, but design features are incorporated to minimize such effects and the higher dose rates are expected O- to be highly localized and/or intermittent. Actual in plant zones and control of personnel access will be based upon survejs conducted by the plant health physics staff.
Areas which may require occupancy to permit an operator to aid in the long term recovery from an accident are considersd in the design. Such areas include the control room, Technical Support Center (TSC), safety-related motor l 1 control centers and switchgear, post accident sampling system room, radio-V chemistry laboratory, and remote 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 limits 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 O a radiation level that could cause a whole body exposure i r, 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 rariiation area is required, positive control is exercised over each individual entry. To the extent practicable, the measured radiation level and the location of the source is posted at the entry to any radiation area or hi 0h radiatic% area.
The posting of radiation signs, control of persennel access, and use of alarms and locks are in compliance with requirements of 10 CFR 20.203.
I l
WAPWR-RP 12.3-12 AMENDMENT 1 B004e:1d FEBRUARY, 1986 I l
l 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 j year of operation lowered the average plant dose.
o In 1982, the highest and lowest collective dose was about 1600 and 100 g man-rem, respectively. In addition, the average collective cose dropped to 650 man-rem.
Factors which may have contributed to this increase in plant collective doses include the following(2);
o Increasing plant radiation fields o Required or mandated modifications /back-fits o Premature failures of major components 1 l
o Use of inexperienced workers o Management attitude In the analysis of ORE data cumulative man-rem per cumulative MW,-Yr of electricity generated accocts for plant size, and provides a relative measure of costs (man-rem) versus benefits (power production). Table 12.4-1 provides j a cumulative summary (up through 1982) of collective doses for domestic plants l
with Westinghouse-supplied NSSSs. Also included on Table 12.4-1 is a measure of the effective operating time (MW,-Yr/MW,).
l As it can be seen from Table 12.4-1, cumulative man-rem per HW,-Yr ranges from 0.30 to 2.68. The best performing pla ts operate in the rarige of 0.3 to 0.4 man-ram per MW, Yr. 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 1
O WAPWR-RP 12.4-2 AMENDMENT 1 5004e:1d FEBRUARY, 1986 l
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