ML17212A082

Final Safety Analysis Report, Rev. 30, Chapter 12, Radiation Protection

ML17212A082
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Site:	Millstone
Issue date:	06/29/2017
From:	Dominion Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation
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MPS-3 FSARMillstone Power Station Unit 3 Safety Analysis Report Chapter 12 MPS-3 FSAR 12-i Rev. 30CHAPTER 12 - RADIATION PROTECTION Table of ContentsSection TitlePage

12.0INTRODUCTION

....................................................................................12.0-1 12.1ENSURING THAT OCCU PATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA).......................................................................12.1-112.2RADIATION SOURCES.........................................................................12.2-112.2.1CONTAINED SOURCES........................................................................12.2-112.2.1.1Sources for Design Basis Loss-of-Coolant Accident...............................12.2-212.2.2AIRBORNE RADIOACTIVE SOURCES...............................................12.2-212.

2.3REFERENCES

FOR SECTION 12.2.......................................................12.2-412.3RADIATION PROTECTION DESIGN FEATURES.............................12.3-112.3.1SHIELDING.............................................................................................12.3-112.3.1.1Primary Shielding.....................................................................................12.3-312.3.1.2Secondary Shielding.................................................................................12.3-3 12.3.1.3Accident Shielding....................................................................................12.3-512.3.1.3.1Containment and Control Room Design...................................................12.3-512.3.1.3.2Post-Accident Access to Vital Areas........................................................12.3-612.3.2FACILITY DESIGN FEATURES.........................................................12.3-1212.3.2.1Location and Design of Equipment to Minimize Service Time........................................................................................................12.3-1212.3.2.2Location of Instruments Requiring In Situ Calibration..........................12.3-1312.3.2.3Location of Equipment Requiring Servicing in Lowest Practicable Radiation Fiel d (or Movable to Lowest Practicable Radiation Field)....................................................................12.3-1312.3.2.4Valve Location and Selection.................................................................12.3-1412.3.2.5Penetrations of Shielding and Containment Walls by Ducts and Other Openings................................................................................12.3-1412.3.2.6Radiation Sources and Occupied Areas..................................................12.3-1412.3.2.7Minimizing Spread of Contam ination and Facilitation of Decontamination Following Spills.........................................................12.3-1512.3.2.8Piping to Minimize Buildup of Contamination......................................12.3-1512.3.2.9Flushing or Remote Chemical Cleaning of Contaminated Systems...................................................................................................12.3-1512.3.2.10Ventilation Design..................................................................................12.3-1612.3.2.11Radiation and Airborne Contamination Monitoring...............................12.3-1612.3.2.12Temporary Shielding..............................................................................12.3-16 MPS-3 FSARCHAPTER 12 -RADIATION PROTECTION Table of Contents (Continued)

Section Title Page 12-ii Rev. 3012.3.2.13Solid Waste Shielding.............................................................................12.3-1612.3.2.14Remote Handling Equipment..................................................................12.3-1612.3.2.15Maximum Expected Failures of Fuel Element Cladding and Steam Generator.....................................................................................12.3-1612.3.2.16Sampling Stations...................................................................................12.3-1712.3.2.17Cobalt Impurity Specifications...............................................................12.3-1712.3.2.18Reactor Cavity Filtration System............................................................12.3-1712.3.3VENTILATION......................................................................................12.3-1712.3.3.1Design Objectives...................................................................................12.3-17 12.3.3.2Design Description.................................................................................12.3-1812.3.3.3Personnel Protection Features.................................................................12.3-1912.3.3.4Radiological Evaluation..........................................................................12.3-20 12.3.4AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING.......................................................................................12.3-2012.3.4.1Purpose....................................................................................................12.3-20 12.3.4.2System Design........................................................................................12.3-2012.3.4.3Class 1E Area Monitors..........................................................................12.3-2112.3.4.4Non-Class 1E Area Monitors..................................................................12.3-22 12.3.4.5Airborne Radioactivity Monitoring........................................................12.3-2212.

3.5REFERENCES

FOR SECTION 12.3.....................................................12.3-2212.4DOSE ASSESSMENT.............................................................................12.4-112.5HEALTH PHYSICS PROGRAM............................................................12.5-1 12.5.1ORGANIZATION....................................................................................12.5-112.5.2EQUIPMENT, INSTRUMENTATION, FACILITIES............................12.5-2 12.5.3PROCEDURES........................................................................................12.5-612.5.4REFERENCE FOR SECTION 12.5.......................................................12.5-11 MPS-3 FSAR 12-iii Rev. 30CHAPTER 12 - RADIATION PROTECTION List of Tables Number Title12.2-1Parameters Used in Calculation of Design Radiation Source Inventories (HISTORICAL)12.2-2Radioactive Sources in Cont ainment Building (HISTORICAL)12.2-3Radioactive Sources in the Auxiliary Building (HISTORICAL)12.2-4Radioactive Sources in the Waste Disposal Building (HISTORICAL)12.2-5Radioactive Sources in the Fuel Building (HISTORICAL) 12.2-5AOther Radioactive Sources (HISTORICAL) 12.2-6Inventory of an Average Fuel Assembly after 650 Days of Operation at 3,636 MWt at Shutdown and 100 Hours after Shutdown (µCi) (HISTORICAL)12.2-7Source Intensity in the Most Radioact ive Fuel Assembly

• after 650 Days of Operation at 3636 Mwt (HISTORICAL)12.2-8Radionuclide Concentrations in the Spent Fuel Pool from Refueling 100 Hours after Shutdown* (HISTORICAL)12.2-9Radiation Sources

• Reactor Coolan t Nitrogen-16 Activi ty (HISTORICAL)12.2-10Assumptions Used in the Calculation of Airborne Concentrations (HISTORICAL)12.2-11Airborne Concentrations Inside Ma jor Buildings (mci

/cc) (HISTORICAL)12.3-1Radiation Zones 12.3-2Radiation Monitoring System -

Area Radiation Detector Location12.3-3Operator Activity Loca tions and Time Durations12.3-4Activity Initiation Time 12.5-1Deleted by FSARCR 04-MP3-040 12.5-2Deleted by FSARCR 04-MP3-040 MPS-3 FSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

12-iv Rev. 30CHAPTER 12 - RADIATION PROTECTION List of Figures Number Title12.2-1Arrangement - Operating Personnel Access and Egress12.2-2Arrangement - Operating Personnel Access and Egress12.2-3Arrangement - Operating Personnel Access and Egress12.2-4I 131 Concentration Containment12.3-1Design Basis Radiation Zones fo r Shielding (Normal Operations)12.3-2Design Basis Radiation Zones fo r Shielding (Normal Operations)12.3-3Design Basis Radiation Zones fo r Shielding (Normal Operations)12.3-4Design Basis Radiation Zones fo r Shielding (Normal Operations)12.3-5This figure moved to Section 11.5 (Figure 11.5-2) 12.3-6Design Basis Radiation Zones for Shielding (Shutdown/Refueling)12.3-7Design Basis Radiation Zones for Shielding (Shutdown/Refueling)12.3-8Design Basis Radiation Zones for Shielding (Shutdown/Refueling)12.3-9Design Basis Radiation Zones for Shielding (Shutdown/Refueling)12.3-10Routes to Post-Accident Vital Areas 12.3-11Fuel Transfer Tube Shielding 12.3-12Upper Reactor Cavity Neutron Shield 12.5-1Figure has been deleted MPS3 UFSAR12.0-1Rev. 30CHAPTER 12 - RADIATION PROTECTION

12.0INTRODUCTION

Prior to the licensing and operation of a nuclear power reactor, the applicant must include, in Chapter 12 of the FSAR, an estimate of the radiation dose expected to be received by station personnel. This includes an estimate for both w hole body dose from direct radiation and internal dose from airborne activity. This is provided to ensure the pr oposed station design related to occupational radiation exposure control (e.g., shield ing and airborne activity control) will be sufficient to ensure compliance with 10 CFR

20. The assessments presen ted in Chapter 12 are based on nominal assumptions and generic models and criteria that were appropriate at the time the original FSAR was written. They represent estimates chosen for the purpose of projecting occupational dose consequences. They do not represent design or operational requirements. It was fully expected that actual operational data woul d not match the chosen assumptions and criteria presented in Chapter 12, but in general, the es timates were expected to be conservative.

Once the plant is operational, compliance with occupational exposure limi ts and controls is ensured and controlled by compliance with the Technical Specificati ons and 10 CFR 20. These documents require a Radiation Protection Prog ram and an ALARA Program. These are dynamic programs that change to meet changing regulati ons, industry initiatives and state-of-the-art practices. These programs provide detailed cont rols on occupational exposure. The Radiation Protection Program ensures that the requireme nts of 10 CFR 20 are met. The ALARA Program ensures that controls are imposed and assessments are performed to reduce occupational exposure to levels that meet current standards and are not the original estimates of Chapter 12. These programs are routinely audited by licensee and NRC staff for compliance and effectiveness.

Annual occupational exposure reports are provided to the NRC to pr ovide a real ti me measure of the effectiveness of occupa tional exposure controls.

Therefore, compliance with oc cupational exposure regulations is controlled by the Radiation Protection and ALARA programs, which are desc ribed in Chapter 12 of the FSAR. Current program measures are contained in Chapter 12, however, the design parameters or quantities provided here are not updated from original valu es. The original bases for station design and radiation control are relegated to historical perspective, as thes e bases were never intended to describe conditions of operation. More accurat e information on radiological quantities or conditions should be determined by referencing curr ent radiation protection data available in the radiation protection department.

MPS3 UFSAR12.1-1Rev. 3012.1ENSURING THAT OCCUPATIONAL RADIATION EXPO SURES ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA)

It is Millstone policy to implement a program that meets the intent of 10 CFR 20 and ensure that occupational radiation exposures at its nuclear facilities ar e kept "as low as reasonably achievable" (ALARA).The ALARA Program criteria shall be in accordance with 10 CFR 20, Regulatory Guide 8.8, Rev.

4 and Regulatory Guide 8.10 (Rev. 1-R); and, it should meet the intent of INPO 91-014, Rev 1.

The program shall ensure that:

Annual and lifetime doses to i ndividuals are ALARA. External and internal exposure is optimized by keeping TEDE ALARA.

Annual collective doses, (p erson rem) are ALARA.

Individual doses within work groups ar e balanced to be consistent with:a.experience b.manpower availabilityc.existing agreements Annual and three year goals are devel oped for collective doses at each unit.

Outage goals are developed thirty to sixt y days prior to the start of an outage.

An ALARA job review process ex ists for jobs with the pote ntial for significant exposure.

ALARA economic evaluations are performed in support of backfits, modifications, decommissioning, etc.

Personnel are aware of ALARA program ph ilosophy and trained in ALARA concepts.Millstone management provides the necessary policy, resources and commitment for ALARA program.

An ALARA feedback system exists for worker s to identify ALARA concerns or suggest ALARA improvements.Corrective actions are considered when the attainment of specific ALARA goals are jeopardized.

MPS3 UFSAR12.2-1Rev. 3012.2RADIATION SOURCES12.2.1CONTAINED SOURCESThe radioactivity values provided in this section are the design basis values used for the design of plant shielding. As such they are considered hi storical and not subject to future updating. This information is retained to avoid loss of origin al licensing bases. As discussed in Chapter 12.0, compliance with occupational exposure limits and controls is ensured and controlled by compliance with the Technical Specifications and 10 CFR 20 wh ich was implemented at MPS-3 via the Radiation Protection Pr ogram and the ALARA Program.The source of radioactivity contained in the streams of the various radioactive waste management systems are the nuclides generated in the reactor core and activation of nuclides in the reactor coolant system and the air surr ounding the reactor vessel. These s ources are described in Chapter 11. Table 12.2-1 presents the principal parameters which are used to establish design radiation source inventories. The design basis for the shield ing source terms for fission products in this section is cladding defects in fuel rods producing 1 percent of the core thermal power. The design basis for activation and corrosion product activitie s are derived from meas urements at operating plants and are independent of fuel defect level.

The radionuclide activity levels in the reactor coolant at the design basis level are given in Section 11.1. The models and assumptions used in determining these sources are also given in Section 11.1.

The reactor core source description is similar in that given in Topical Report RP-8A (Stone & Webster Engineering Corporation), Section 4.1.1, with appropriate adjustment to account for power level difference.The activity of a spent fuel assembly is calc ulated using appropriate fission yields, decay constants, and thermal neutron cross sections. Isotopic inventories are based on full power operation for 650 days. The inventory of an aver age fuel assembly at shutdown and 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown is given in Table 12.2-6. The source strengths in MeV/sec for the most radioactive fuel assembly at several decay times (assuming a radial peaking factor of 1.65) is given in Table 12.2-7. The isotopic activities (expected and de sign) in the fuel pool water are given in Table 12.2-8 based on expected and design primary coolant activities homogeneously mixed with refueling cavity water and spent fuel pool water from refueling operations 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown.The location and geometry of significant sour ces of radiation in the containment building, auxiliary building, fuel buildi ng, waste disposal building, condensate polishing building, ESF building, and the tank yard are presented in Tables 12.2-2 through 12.2-5A and on Figures 12.2-1 through 12.2-3. The method used to arrive at the source terms presented in Tables 12.2-2 through 12.2-5A considers the operating parameters described in Table 11.1-3. The evaporator bottoms activity is based on liqui d particulate concentrat ions at the start of plant operations, thereby minimizing the decay time to produce a batch of concentrate. Normal operating flow rates are also used to develop demineralizer nuclide concentrations consistent with the de contamination factors used to determine radioactive concentrations of waste liquid streams. The inventory of radioactive nuclides on filters downstream of the demineralizer assumes a fraction of the resin fines are MPS3 UFSAR12.2-2Rev. 30 transported via the liquid flow and deposited on the filter cartridge. Significa nt sources are listed by name in the tables and are numbered to co rrespond to numbered lo cations on the figures.

Radiation source terms are also presented in these tables in terms of seven discrete energy levels (MeV) used in plant shielding design.The eighth energy level corresponds to the N-16 activity in the reactor coolant for various components, listed in Table 12.2-2. N-16 activity is the controlling source in the design of the secondary shield and is tabulated in Table 12.2-9, in Ci per gram of coolant, as a function of transport time through the reactor coolant loop.The method used to calculate the source strength of N-16 in each component is described in detail in the Stone & Webster Topical Report RP-8A (Stone & Webster Engineering Corporation). The required parameters for this calcu lation are transit time to the component, transit time through the component. N-16 is not a factor in the radiation sources for th e systems and components located outside the containment due to its short (7.11 s econds) half life and the greater than one minute transport time before the letdown flow exits the containment.12.2.1.1Sources for Design Basis Loss-of-Coolant Accident

The radiation sources of importance for the design basis accident are the sources within the containment and sources transported via the emergency safegua rds features (ESF) cooling system.The fission product radiation sources considered to be released from the fuel to the containment following a maximum credible accident are base d on the assumptions given in Regulatory Guide 1.4 and NUREG-0737. This s ource term was used to evaluate the original plant shielding design, and is not used in Chapter 15 analyses.The sources in the ESF system are based on the nongaseous activity, i.e., 50 percent halogens and 1 percent remainder, being retained in the coolant water. Noble gases formed by the decay of halogens in the sump water are assumed to be rele ased to the containment and not retained in the water. Credit has been taken for dilution by the reactor coolant system volume plus the contents of the refueling water storage tank and other ESF system component volumes.

Isotopic fission product sources in the Fuel are given in Section 11.1.12.2.2AIRBORNE RADIOACTIVE SOURCES The principal sources of airborne radionuclide s are the reactor coolant system and the air surrounding the reactor vessel. Reactor coolant leak ing into plant buildings results in the release of airborne contamination. The radioactivity sour ces which contribute to the radioactive airborne releases from the plant waste management system and the plant ventilati on system are described in Chapter 11.

Concentrations of airbor ne activity for the expected and design conditions in the containment structure, turbine building, and fuel building are listed in Table 12.2-11. The bases used to derive MPS3 UFSAR12.2-3Rev. 30 these concentrations are presented in Table 12.2-10 and in NU REG-0017. Airborne radioactivity concentrations in aisleways and manned spaces in the auxiliary building and at other locations is considered negligible.Airborne levels in general access areas of the auxiliary, turbine, and fuel buildings are expected to be lower than equipment cubicles of these build ings, since ventilation flow paths are normally directed from areas with less potential contamination to areas with greater potential for contamination.Containment StructureThe containment structure is not normally occupied during power operation.

Radiation protection procedures control access. Two r ecirculating charcoal filters ca n be operated to ensure that airborne iodine in the containment is as low as is reasonably achievable for work in that area (Section 9.4.7).

Figure 12.2-4 presents expected i odine-131 concentrations in the containment structure after the two charcoal filters have been placed in operation.After shutdown, the containment purge air system can be used to reduce the airborne activity within the containment structure. The filtered purge is rated at 30,000 cfm.Radioactivity associated with primary coolant leakage is mixed in the containment atmosphere by the containment atmosphere recirculation sy stem (Section 9.4.7.1). Re moval occurs through radioactive decay. At appropriate times, the ra dioactive inventory may be reduced by means of recirculating the containment air through the charcoal filters or containment purging.

The containment atmosphere tritiu m assumes the same relative concentration (Ci of tritium per gram of water) as exists in the reactor coolant leakage.Turbine BuildingRadioactivity associated with steam leakage is assumed to be uniformly distributed by the turbine building ventilation sy stem (Section 9.4.4).

Removal occurs through decay and ventilation exhaust. The tritiu m concentration in the turbine building is calculated assuming that all the steam leakage into the turbine building remains gaseous.Auxiliary Building Airborne radioactivity associated with primary coolant leakage is assumed to be limited to process equipment cubicles and is remove d by auxiliary building ventilation system (Section 9.4.3). Removal occurs through decay and ve ntilation exhaust. The ve ntilation system is configured to preclude mixing of the atmosphe re in process equipment cubicles with the atmosphere in the general access areas as described above.

MPS3 UFSAR12.2-4Rev. 30The tritium concentration in th e auxiliary building atmosphere is conservatively calculated assuming all the primary coolant leakage into the auxiliary building evaporates when in fact the leakage is collected in sumps and drains a nd is not generally available for evaporation.Fuel Building The fuel building ventilation syst em is described in Section 9.4.2.

The tritium concentrations in the fuel building atmosphere assumes that the atmosphere above spent fuel pool has the same relative tritium concentration, Ci of tritium per gram of water, as the fuel water. Airborne concentrati ons are presented in Table 12.2-11.12.

2.3REFERENCES

FOR SECTION 12.212.2-1NUREG-0017 United States Nuclear Regulator y Commission. Calculati on of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE CODE). 1976 Office of Standards Development.12.2-2Regulatory Guide 1.4.12.2-3Stone & Webster Engineering Corporati on (SWEC) 1975. Radiation Shielding Design and Analysis Approach for Light Water Reactor Power Plants RP 8A. Topical Report.

Cambridge, Mass.12.2-4Westinghouse Letter NEU-3492, Dated July 31, 1980.

MPS-3 FSARPage 1 of 1Rev. 30TABLE 12.2-1PARAMETERS USED IN CAL CULATION OF DESIGN RADIATION SOURCE INVENTORIES (HISTORICAL)

Parameter 1. Power level (MWt) 3,636 2. Failed Fuel Fraction 0.01

3. Primary-to-Secondary Leak Rate (lb/day) 1,370

4. Reactor Operating Time (days) 650 5. Escape Rate Coefficients (sec

-1): 1. Noble Gases 6.5 x 10

-82. Br, Rb, I, and Cs nuclides 1.3 x 10

-83. Te nuclides 1.0 x 10

-94. Mo nuclides 2.0 x 10

-95. Sr and Ba nuclides 1.0 x 10-116. Y, La, Ce, Pr nuclides 1.6 x 10

-126. Purification Letdown Flow Rate (gpm) 75

7. Degasification Charcoal Delay Bed Holdup Time (days): 1. Kr 6.1

2. Xe 142.1 Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.

MPS-3 FSARPage 1 of 2Rev. 30TABLE 12.2-2RADIOACTIVE SOURCES IN CONTAINMENT BUILDING (HISTORICAL) Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source 0.400.801.301.702.202.503.506.10Height (cm)Outside Diameter (cm)Volume (cc) 1Reactor Core 1.05+12 (1)4.63+124.68+123.83+127.98+111.27+123.29+122.23+0713034391.97+082Regenerative Heat Exchanger Shell1.73+052.91+051.26+051.22+059.74+041.36+053.03+042.01+07427.030.483.12+05Tube7.23+036.11+042.67+043.99+041.92+031.20+049.42+03-3Steam GeneratorShell1.44+052.43+051.05+051.02+058.14+041.14+052.53+041.94+073352.11+08Tube2.97+011.38+023.19+011.62+011.79+009.84+011.21+00-19704274Pressurizer Liquid2.83+041.62+058.61+046.30+041.13+041.01+041.85+044.93+0514992133.06+07Gas8.79+046.22+044.92+033.52+049.16+041.30+054.20+03-2.04+075Containment Drains Transfer Tank1.98+053.33+051.43+051.40+051.11+051.55+053.46+04-259145.03.75+066Reactor Coolant Piping1.34+052.26+059.74+049.49+047.57+041.06+052.35+04(See Table12.2-9)---9Excess Letdown Heat Exchanger1.81+053.05+051.31+051.28+051.02+051.43+053.17+041.32+0736630.482.67+0510Pressurizer Relief Tank3.87+042.74+042.16+031.55+044.03+045.72+041.85+03-8262905.10+07 MPS-3 FSARPage 2 of 2Rev. 30 (1) 1.05+12 = 1.05 x 10 12 Historical, not subject to future updating. This table ha s been retained to preser ve original design basis.11Pressurizer Relief Tank Transfer Pump3.87+042.74+042.16+031.55+044.03+045.72+041.85+03-29.238.13.02+0412Containment Drains Transfer Pump1.98+053.33+051.43+051.40+051.11+051.55+053.46+04-29.238.13.02+0413Reactor Coolant Pump1.51+052.54+051.09+051.07+058.52+041.19+052.64+041.25+07427.0122.05.0+06TABLE 12.2-2RADIOACTIVE SOURCES IN CONTAINMENT BUILDING (HISTORICAL) Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source 0.400.801.301.702.202.503.506.10Height (cm)Outside Diameter (cm)Volume (cc)

MPS-3 FSARPage 1 of 3Rev. 30TABLE 12.2-3RADIOACTIVE SOURCES IN THE AUXILIARY BUILDING (HISTORICAL)Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source0.400.801.301.702.202.503.50 Height (cm)Outside Diameter (cm)Volume (cc)14Cesium Removal Ion Exchanger 1.87+07 (1)5.02+082.67+071.48+063.34+051.46+056.02+04206.4106.71.71+0615Boron Evaporator Reboiler4.68+051.35+061.16+052.73+048.74+031.70+022.09+01441.953.39.21+0516Boron Evaporator4.68+051.35+061.16+052.73+048.74+031.70+022.09+01983.02292.14+0717Boron Evaporator Reboiler Pump4.68+051.35+061.16+052.73+048.74+031.70+022.09+01104.143.25.29+0418Boron Evaporator Bottoms Pump4.68+051.35+061.16+052.73+048.74+031.70+022.09+0135.618.43.82+0319Boron Recovery Filters9.00+062.45+086.06+079.57+051.60+056.97+042.88+0483.317.72.08+0420Boron Evaporator Bottoms Filter4.68+051.35+061.16+052.73+048.74+031.70+022.09+0178.417.71.95+0421Reactor Coolant Filter1.72+084.11+086.33+071.32+073.08+062.88+051.74+0548.216.81.07+0422Sealwater Injection Filter4.02+054.72+074.68+082.44+06------55.87.02.14+0323Sealwater Heat Exchanger1.09+045.96+042.62+044.45+041.48+031.28+041.14+04419.135.54.17+0524Thermal Regeneration Demineralizer5.05+064.28+062.73+051.37+053.94+043.87+01 1.30+02253.6122.02.10+0625Cation Bed Demineralizer2.38+054.34+081.39+070.00+000.00+000.00+000.00+00274.966.09.42+0526Mixed Bed Demineralizer2.25+085.37+088.27+071.73+074.02+063.76+052.27+05270.881.31.40+0627Letdown Heat Exchanger1.67+052.75+051.27+051.32+051.05+051.43+054.14+04541.053.31.21+0628Letdown Reheat Heat Exchanger1.67+052.75+051.27+051.32+051.05+051.44+054.15+0422122.99.10+04 MPS-3 FSARPage 2 of 3Rev. 3029Letdown Chiller Heat Exchanger1.37+051.24+053.07+048.16+049.96+041.51+051.59+04518.253.31.16+0630Moderating Heat Exchanger1.37+051.24+053.07+048.16+049.96+041.51+051.59+04556.345.79.12+0531Volume Control Tank Liquid8.39+038.87+043.24+044.30+043.09+031.30+041.25+04125.5228.64.53+06Volume Control Tank Gases6.67+044.24+043.00+032.16+046.55+049.67+041.29+03188.3228.66.80+0632Degasifier Recirculation Pump1.13+049.46+042.64+043.87+041.40+031.16+048.19+0335.127.01.57+0433Sealwater Return Filter7.51+048.82+068.74+074.56+05------48.316.81.07+0434Letdown Filter7.56+058.88+078.80+084.59+06------48.316.81.07+0435Boric Acid Tanks4.68+051.35+061.16+052.73+048.74+031.70+022.09+015794881.08+0836Boric Acid Filters4.68+051.35+061.16+052.73+048.74+031.70+022.09+0148.316.81.07+0437Boric Acid Transfer Pumps4.68+051.35+061.16+052.73+048.74+031.70+022.09+0134.328.62.19+0438Primary Drains Transfer Tanks1.98+053.33+051.48+051.40+051.11+051.55+053.46+04356.9152.46.51+0639Primary Drains Transfer Pump1.98+053.33+051.48+051.40+051.11+051.55+053.46+0429.830.52.17+0440Process Gas Charcoal Bed Adsorbers2.56+072.67+051.46+051.90+054.73+057.48+057.91+04492.7182.81.29+0741Degasifier Recovery Heat Exchanger1.55+051.45+053.32+048.30+041.02+051.58+051.28+04709.038.17.11+0542Degasifier Trim Cooler1.55+051.45+053.32+048.30+041.02+051.58+051.28+04652.827.33.86+0543Degasifier Liquid1.13+049.46+042.64+043.87+041.40+031.16+048.19+03538.8167.69.78+06TABLE 12.2-3RADIOACTIVE SOURCES IN THE AUXILIARY BUILDING (HISTORICAL)Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source0.400.801.301.702.202.503.50 Height (cm)Outside Diameter (cm)Volume (cc)

MPS-3 FSARPage 3 of 3Rev. 30 (1) 1.87+07 = 1.87x10 7 Historical, not subject to future updating. This table ha s been retained to preser ve original design basis.Degasifier Vapor1.51+037.78+024.89+013.91+021.05+031.50+033.68+01152.4167.61.12+0644Degasifier Feed Preheater1.51+051.40+053.22+048.05+049.88+041.53+051.24+04373.430.482.72+0545Fuel Pool Post Filter2.01+053.77+054.31+045.77+032.50+031.18+033.03-0211445.71.87+0546Fuel Pool Demineralizer2.01+073.75+074.27+065.76+052.49+051.18+053.03+00259.160.97.34+0547Process Gas Receiver0.0+008.55+020.0+000.0+000.0+000.0+000.0+0018361.05.34+05 48Charging Pumps4.76+051.44+061.48+057.03+041.18+041.32+041.25+0414591.49.50+0549Boron Evaporator Bottoms Cooler4.68+051.35+061.16+052.73+048.74+031.70+022.09+01336.521.91.26+0550Boron Evaporator Distillate Cooler4.89-011.35+003.43-011.78-013.40-023.75-023.40-03340.321.91.28+0553Degasifier Condenser1.23+066.32+053.97+043.17+058.54+051.21+062.99+04152.476.26.95+0454Effluent Filters7.58-016.79+003.02+001.79+003.74-019.46-041.56-04147.332.41.21+05TABLE 12.2-3RADIOACTIVE SOURCES IN THE AUXILIARY BUILDING (HISTORICAL)Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source0.400.801.301.702.202.503.50 Height (cm)Outside Diameter (cm)Volume (cc)

MPS-3 FSARPage 1 of 2Rev. 30TABLE 12.2-4RADIOACTIVE SOURCES IN THE WASTE DISPOSAL BUILDING (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source 0.400.80 1.301.702.20 2.50 3.50Height (cm)Outside Diameter (cm)Volume (cc)51Boron Demineralizer 1.31+03 (1)3.26+049.59+025.12+011.38+019.80-012.31-02206.4106.71.7+0652Boron Demineralizer Filter5.97+021.48+044.36+022.33+016.27+004.46-011.05-0287.317.82.17+0455Waste Evaporator3.80+041.31+051.38+041.80+035.57+022.18+026.70-03973.5182.92.10+0756Waste Evaporator Reboiler3.80+041.31+051.38+041.80+035.57+022.18+026.70-03621.355.91.52+06 57Waste Evaporator Reboiler Pump3.80+041.31+051.38+041.80+035.57+022.18+026.70-03109.243.26.43+0458Boron Evaporator Feed Pumps1.61+021.34+046.38+027.98+001.37+001.69+001.98+0043.218.45.06+0359Waste Evaporator Feed Pumps1.61+035.51+036.70+021.46+023.59+011.21+012.48+0035.618.43.82+0360Waste Evaporator Bottoms Pump3.80+041.31+051.38+041.80+035.57+022.18+026.70-0335.618.43.82+0361Spent Resin Hold Tank2.34+078.36+079.07+061.75+063.88+056.64+044.39+04360.0182.98.90+0662Spent Resin Transfer Pump Filter2.25+065.37+068.27+051.73+054.02+043.76+032.27+0383.817.82.08+0463Spent Resin Transfer Pump2.25+065.37+068.27+051.73+054.02+043.76+032.27+0330.530.52.22+0464High-Level Waste Drain Tank1.61+035.51+036.70+021.46+023.59+011.21+012.48+001295.4320.01.01+0865Low-Level Waste Drain Tank7.64-017.11+003.13+001.85+003.74-019.46-041.56-04340.4274.31.80+0766Low Level Waste Drain Pump7.64-017.11+003.13+001.85+003.74-019.46-041.56-0435.618.43.82+0367Waste Evaporator Bottoms3.80+041.31+041.38+041.80+035.57+022.18+026.70-03335.321.91.27+0568Waste Distillate Cooler3.80+001.31+011.38+001.80-015.57-022.18-026.70-07365.827.32.17+05 MPS-3 FSARPage 2 of 2Rev. 30 (1) 1.31+03 = 1.31 x 10 3 Historical, not subject to future updating. This table ha s been retained to preser ve original design basis.72Waste Demineralizer1.25+031.37+059.10+034.91+011.13+011.38+004.79-01206.4106.71.70+0673Waste Demineralizer Filter5.67+026.22+044.14+032.23+015.13+006.27-012.18-0187.317.82.17+04TABLE 12.2-4RADIOACTIVE SOURCES IN THE WASTE DISPOSAL BUILDING (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)Source No. Source 0.400.80 1.301.702.20 2.50 3.50Height (cm)Outside Diameter (cm)Volume (cc)

MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.67+11 = 1.67x10 11 Historical, not subject to future updating. This table ha s been retained to preser ve original design basis.TABLE 12.2-5RADIOACTIVE SOURCES IN THE FUEL BUILDING (HISTORICAL)Activity (MeV/cc-Sec) for Energy (MeV)Source 0.400.801.30 1.70 2.20 2.50 3.50 Height (cm)Diameter (cm)Volume (cc)69Most Radioactive Fuel Assembly 100Hours after Sutdown1.67+11 (1)1.02+127.21+105.10+111.49+103.68+101.01+0970Fuel Pool Purification Filter3.09+024.02+056.22+04--1.66+037.20+011.06+011.68+0245.72.75+0574Fuel Pool (Filled to Capacity 100Hours after Shutdown)5.58+39.02+31.05+31.14+25.16+12.42+14.48-2 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE: (1) 1.66+05=1.66x10 5 Historical, not subject to future updating. This table ha s been retained to preser ve original design basis.TABLE 12.2-5AOTHER RADIOACTIVE SOURCES (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)Source No.Source0.400.801.301.702.202.503.50 Height (cm)Outside Diameter (cm)Volume (cc)ESF Building 7RHR Pump1.66+05 (1)2.13+056.84+044.31+043.12+043.76+044.65+0343.299.13.32+058RHR Exchanger1.66+052.13+056.84+044.31+043.12+043.76+044.65+031,370100.01.09+07Yard Tanks71Boron Recovery Tank1.61+021.34+046.38+027.98+001.37+001.69+001.98+009.14+029.14+025.70+0875RWST1.68-021.17+008.11-021.09-051.89-061.19-062.27-101.80+031.80+034.41+09 Condensate Polishing Building76Condensate Polishing Demineralizer8.12+033.63+045.15+031.13+033.46+021.19+021.61+003.35+022.44+025.67+0677Cation Regeneration Tank3.38+037.16+048.96+034.26+021.22+026.80+013.58+004.27+022.08+022.21+0678Anion Regeneration Tank1.15+041.43+042.84+031.64+035.06+021.56+023.59-014.57+021.83+023.34+06 MPS-3 FSARPage 1 of 5Rev. 30TABLE 12.2-6INVENTORY OF AN AVERAGE FUEL ASSEMBLY AFTER 650 DAYS OF OPERATION AT 3,636 MW t AT SHUTDOWN AND 100 HOURS AFTER SHUTDOWN (µC i) (HISTORICAL)Isotope0 Hours100 HoursKr-83m8.19+10 *1.07-01Kr-85m2.05+112.91+04 Kr-854.58+094.58+09Kr-873.99+11 **Kr-885.60+119.17+00 Kr-897.25+11 **Xe-131m4.15+088.55+10Xe-133m2.53+101.09+10 Xe-1331.05+127.25+11Xe-135m2.85+119.53+06Xe-1352.79+111.45+09 Xe-1379.48+11 **Xe-1389.33+11 **Br-838.19+102.37-02 Br-841.47+11 **Br-852.05+11 **Br-873.95+11 **

I-1291.12+041.12+04I-1314.72+113.38+11I-1326.74+112.84+11 I-1331.06+123.77+10I-1341.23+12 **I-1359.74+113.06+07 I-1364.89+11 **Se-812.84+10 **Se-833.47+10 **Se-841.47+11 **Rb-885.65+111.03+01 MPS-3 FSARPage 2 of 5Rev. 30Rb-897.51+11 **Rb-909.17+11 **

Rb-918.55+11 **Rb-927.10+11 **Sr-897.51+117.10+11 Sr-903.73+103.73+10Sr-919.17+117.10+08Sr-928.34+115.34+00 Sr-939.33+11 **Sr-947.25+11 **Y-903.70+103.70+10 Y-91m5.39+114.66+08Y-919.33+118.91+11Y-929.33+111.08+04 Y-939.64+111.07+09Y-948.50+11 **Y-959.64+11 **

Zr-959.79+119.38+11Zr-979.33+111.48+10Nb-95m1.95+101.93+10 Nb-951.01+121.01+12Nb-97m8.91+111.42+10Nb-979.79+111.60+10 Mo-999.74+113.44+11Mo-1017.88+11**Mo-1026.58+11**Mo-1051.42+11 **Tc-99m8.55+113.28+11TABLE 12.2-6INVENTORY OF AN AVERAGE FUEL ASSEMBLY AFTER 650 DAYS OF OPERATION AT 3,636 MW t AT SHUTDOWN AND 100 HOURS AFTER SHUTDOWN (µC i) (HISTORICAL)Isotope0 Hours100 Hours MPS-3 FSARPage 3 of 5Rev. 30Tc-1011.58+12 **Tc-1026.58+11 **

Tc-1051.90+11 **Ru-1034.74+114.40+11Ru-1051.42+112.33+04 Ru-1064.46+104.42+10Ru-1073.00+10 **Rh-103m4.74+114.41+11 Rh-105m1.42+112.34+04Rh-1051.42+112.06+10Rh-1064.47+104.42+10 Rh-1073.00+10 **Sn-1271.74+101.10-04Sn-1285.85+10 **

Sn-1301.74+11 **Sb-1272.16+101.03+10Sb-1287.88+09 **

Sb-1291.58+111.80+04Sb-1303.16+11 **Sb-1314.26+11 **

Sb-1325.28+11 **Sb-1335.39+11 **Te-127m4.67+094.95+09 Te-1274.24+091.34+10Te-129m8.65+107.98+10Te-1291.69+118.03+10Te-131m6.94+106.89+09Te-1314.15+111.39+09TABLE 12.2-6INVENTORY OF AN AVERAGE FUEL ASSEMBLY AFTER 650 DAYS OF OPERATION AT 3,636 MW t AT SHUTDOWN AND 100 HOURS AFTER SHUTDOWN (µC i) (HISTORICAL)Isotope0 Hours100 Hours MPS-3 FSARPage 4 of 5Rev. 30Te-1326.74+112.76+11Te-133m7.41+11 **

Te-1334.74+11 **Te-1341.09+12 **Cs-1373.91+103.91+10 Cs-1381.06+12 **Cs-1391.02+12 **Cs-1409.33+11 **

Cs-1425.18+11 **Ba-137m3.65+103.59+10Ba-1391.02+12 **

Ba-1409.95+117.93+11Ba-1419.64+11 **Ba-1429.12+11 **

La-1401.00+128.86+11La-1419.64+111.77+04La-1429.17+09 **

La-1439.27+11 **Ce-1419.59+118.86+11Ce-1439.33+111.15+11 Ce-1446.79+116.68+11Ce-1456.11+11 **Ce-1464.54+11 **

Pr-1439.27+118.24+11Pr-1446.79+116.68+11Pr-1456.17+115.75+06Pr-1464.66+11 **Nd-1473.45+112.66+11TABLE 12.2-6INVENTORY OF AN AVERAGE FUEL ASSEMBLY AFTER 650 DAYS OF OPERATION AT 3,636 MW t AT SHUTDOWN AND 100 HOURS AFTER SHUTDOWN (µC i) (HISTORICAL)Isotope0 Hours100 Hours MPS-3 FSARPage 5 of 5Rev. 30NOTES:* 8.19+10 = 8.19x10 10** Less than 1.00x10

-6 µCi Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.Nd-1491.64+11 **Nd-1516.63+10 **

Pm-1471.13+111.14+11Pm-1491.64+114.59+10Pm-1516.63+105.80+09 Sm-1515.18+075.44+07Sm-1532.42+105.54+09TABLE 12.2-6INVENTORY OF AN AVERAGE FUEL ASSEMBLY AFTER 650 DAYS OF OPERATION AT 3,636 MW t AT SHUTDOWN AND 100 HOURS AFTER SHUTDOWN (µC i) (HISTORICAL)Isotope0 Hours100 Hours MPS-3 FSARPage 1 of 1Rev. 30NOTES:* Includes a radial peaking factor of 1.65

• • 1.77+17 = 1.77 x 10 17 Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.TABLE 12.2-7SOURCE INTENSITY IN THE MOST RADIOACTIVE FUEL ASSEMBLY

• AFTER 650 DAYS OF OPERATION AT 3636 MW t (HISTORICAL)Activity (MeV/Sec) for Energy Group (MeV)

Decay Time (Hrs)0.40.81.31.72.22.53.501.77+17 **7.83+177.91+176.47+171.35+172.15+175.57+1727.45+164.86+171.38+172.23+172.68+163.66+165.75+1546.69+164.11+179.75+161.74+171.88+162.14+162.24+1586.11+163.56+176.76+161.38+171.11+161.21+168.81+14165.47+163.03+174.70+161.13+176.24+158.33+152.84+14205.21+162.84+173.51+161.08+175.38+157.91+152.25+14244.97+162.69+173.03+161.04+174.86+157.70+152.02+14483.93+162.15+171.86+169.58+163.62+157.17+151.85+141002.82+161.72+171.22+168.63+162.52+156.23+151.71+141682.06+161.49+178.55+157.42+161.69+155.19+151.50+147205.81+159.49+161.97+152.13+166.81+141.40+154.68+13 MPS-3 FSARPage 1 of 2Rev. 30TABLE 12.2-8RADIONUCLIDE CO NCENTRATIONS IN TH E SPENT FUEL POOL FROM REFUELING 100 HOURS AFTER SHUTDOWN* (HISTORICAL)Nuclide Expected Concentration

(µCi/cc)Design Concentration (µCi/cc)I-1312.4E-022.2E-01I-1321.4E-031.3E-02 I-1331.8E-031.7E-02I-135 **8.2E-06Sr-894.1E-054.8E-04 Sr-901.2E-062.0E-05Y-90 **2.2E-05Y-918.1E-067.9E-05 Zr-957.1E-068.0E-05Nb-95m **1.7E-06Nb-956.3E-068.7E-05 Mo-993.7E-031.4E-01Tc-99m3.6E-031.3E-01Ru-1035.1E-063.8E-05 Ru-1061.2E-063.8E-06Rh-103m5.1E-063.8E-05Rh-105 **1.5E-06 Rh-1061.2E-063.8E-06Te-127m3.3E-052.4E-04Te-1273.3E-052.4E-04 Te-129m1.6E-044.2E-03Te-1291.6E-044.2E-03Te-131m3.2E-052.7E-04 Te-1316.4E-065.5E-05Te-1321.4E-031.3E-02Cs-1343.2E-033.9E-02Cs-1361.3E-031.6E-02Cs-1372.2E-031.9E-01 MPS-3 FSARPage 2 of 2Rev. 30NOTES:* The expected and design concentrations assume complete mixing of reactor coolant with refueling cavity water and spent fuel pool water 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown.

• • Concentrations less than 1.0E-06

µCi/cc are considered negligible and are not tabulated.

1.0E-06 = 1.0 x 10

-6 Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.Ba-137m2.1E-031.8E-01Ba-1402.2E-054.1E-04 La-1402.3E-054.0E-04Ce-1417.9E-067.5E-05Ce-143 **7.5E-06 Ce-1444.1E-065.8E-05Pr-1435.4E-066.9E-05Pr-1444.1E-065.8E-05 Nd-147 **2.2E-05Pm-147 **9.9E-06Pm-149 **3.3E-06 Cr-512.1E-042.1E-04Mn-543.8E-053.8E-05Fe-552.0E-042.0E-04 Fe-591.2E-041.2E-04Co-581.9E-041.9E-03Co-602.5E-042.5E-04 H-31.2E-014.1E-01TABLE 12.2-8RADIONUCLIDE CO NCENTRATIONS IN TH E SPENT FUEL POOL FROM REFUELING 100 HOURS AFTER SHUTDOWN* (HISTORICAL)Nuclide Expected Concentration

(µCi/cc)Design Concentration (µCi/cc)

MPS-3 FSARPage 1 of 1Rev. 30TABLE 12.2-9RADIATION SOURCES

• REACTOR COOLANT NITROGEN-16 ACTIVITY (HISTORICAL)

• Source: Westinghouse letter NEU-3492, dated July 31, 1980.

Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.Position in LoopLoop Transit Time (sec)Nitrogen-16 Activity (µCi/gm)Leaving Core0.0189Leaving Reactor Vessel1.1170Entering Steam Generator1.4164 Leaving Steam Generator5.4112Entering Reactor Coolant Pump6.0106Entering Reactor Vessel6.898 Entering Core9.086Leaving Core9.7189Nitrogen-16 Energy EmissionEnergy (MeV/gamma) Intensity (percent)1.750.132.749.766.1360.07.125.0 MPS-3 FSARPage 1 of 1Rev. 30NOTE:* Only the area above the fuel pool.

Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.TABLE 12.2-10ASSUMPTIONS USED IN THE CALCULATION OF AIRBORNE CONCENTRATIONS (HISTORICAL)

Containment BuildingTurbine Building Fuel Building1.Reactor coolant equilibrium concentrationsTable 11.1-22.Secondary side equilibrium concentrations-Table 11.1-6-3.Iodine and noble gas core inventory--Table 11.1-14.Leak rate into buildingsA.Equivalent hot reactor coolant (lb/day)4.7x10 3 --B.Equivalent main steam leakage (lb/hr)-1.7x10 3 -5.Normal moisture in atmosphere (%)60--6.Fraction of primary coolant activities released (%/day)A.Noble gases1.0--B.Iodines0.001--7.Mixing in building atmosphere (%)70100100 8.Building ventilation rate (cfm) 3.0x10 4 1.55x10 5 3.0x10 4 9.Building free volume (ft 3)2.32x10 6 4.06x10 6 2.30x10 5 *10.Recirculation - filtersYesNoNo11.Filter efficiency99%--

12.Fuel pool evaporation rate (lb/hr-ft 2)--1.7413.Recirculation rate (cfm) 2.4x10 4 --14.Fuel pool average volume (ft 3)--4.88x10 4 MPS-3 FSARPage 1 of 2Rev. 30TABLE 12.2-11AIRBORNE CONCENTRATION S INSIDE MAJOR BUILDINGS (µCI/CC) (HISTORICAL)

Containment Building Prior to Recirculation Containment Building After 16-Hour Recirculation (1) Turbine Building Fuel BuildingIsotopeDesignExpectedDesignExpectedDesignExpectedDesignExpectedH-3 1.6E-4 (2)4.6E-51.6E-44.6E-51.2E-82.9E-93.8E-61.1E-6I-1319.8E-71.1E-79.1E-91.0E-93.8E-112.9E-132.1E-102.3E-11I-1324.1E-92.3E-101.7E-92.2E-109.8E-127.7E-145.0E-125.4E-13I-1331.7E-71.7E-81.2E-81.3E-95.5E-114.0E-131.8E-121.9E-18I-1349.8E-109.8E-116.5E-106.5E-111.8E-121.3E-14--

I-1352.9E-82.9E-95.6E-95.6E-102.32E-111.7E-132.7E-16-Kr-83m1.6E-68.0E-81.6E-68.0E-84.1E-121.5E-14--Kr-85m1.5E-57.9E-71.5E-57.9E-71.7E-116.6E-14--

Kr-851.0E-46.1E-61.0E-46.1E-63.5E-131.7E-15--Kr-873.0E-61.6E-73.0E-61.6E-71.0E-114.0E-14--Kr-881.9E-51.0E-61.9E-51.0E-63.2E-111.3E-13--

Kr-891.1E-86.4E-101.1E-86.4E-101.7E-137.0E-16--Xe-131m6.2E-23.1E-66.2E-23.1E-61.2E-134.4E-152.0E-82.2E-9Xe-133m6.5E-54.3E-66.5E-54.3E-66.4E-123.2E-144.2E-124.3E-13 Xe-1336.5E-34.3E-46.5E-34.3E-42.8E-101.3E-121.8E-101.8E-11Xe-135m6.2E-79.0E-96.2E-79.0E-99.3E-123.3E-146.2E-15-Xe-1359.2E-54.0E-69.2E-54.0E-65.2E-111.7E-132.6E MPS-3 FSARPage 2 of 2Rev. 30NOTES:(1) Using 99% filter efficiency (2) 1.6E-4 = 1.6 x 10

-4 Historical, not subject to future updating. This table ha s been retained to preser ve original design basis.Xe-1372.1E-81.4E-92.1E-81.4E-93.1E-131.5E-15--Xe-1382.8E-72.5E-82.8E-72.5E-82.8E-121.8E-14--TABLE 12.2-11AIRBORNE CONCENTRATION S INSIDE MAJOR BUILDINGS (µCI/CC) (HISTORICAL)

Containment Building Prior to Recirculation Containment Building After 16-Hour Recirculation (1) Turbine Building Fuel BuildingIsotopeDesignExpectedDesignExpectedDesignExpectedDesignExpected MPS3 UFSAR12.3-1Rev. 3012.3RADIATION PROTECTION DESIGN FEATURES12.3.1SHIELDINGThe assessments performed to determine the original major shield designs were based on assumed source terms, occupancy times and acceptance criteria based on zone criteria. Although these criteria were used to establish the original sh ield design, they were ne ver intended to establish requirements for the radiation protection program implementation during plant operation. As time evolves, source terms change. Ac ceptable doses have typically decr eased with time as ambitious ALARA person-REM goals are established.

Current shielding requirements are non-specific and are established th rough the implementation of the Radiation Protection Program and ALARA Program. These programs evaluate the need for a combination of exposure saving principals such as reduced source term, decreasing occupancy time, or increased shielding. These programs use shielding as one me thod to help ensure compliance with 10 CFR 20.This section provides the basis for the original plant shielding design. Although current dose rates may not be consistent with the zone maps in this chapter, these maps are not being changed to be current, as that would make them inconsistent with the original design basis criteria for the shielding. Recent Heath Physics surveys should be consulted for information on current station radiological conditions.

Radiation shielding is designed to ensure that radiation exposure to the general public and to personnel in-plant is kept to leve ls as low as is reasonably achie vable (ALARA), c onsistent with the requirements set forth in 10 CFR 20 for normal operation and 10 CFR 50.67 for accident conditions and with the overall objectives set forth in USNRC Regulatory Guide 8.8. The original design of this radiation shieldi ng was based upon radiation zone criteria which were established in support of the expected access requirements and durations of occupancy during normal operations, during refueling outages, and during acc ident situations. Descriptions of the zone criteria are presented in Table 12.3-1, and the detailed radiati on zone criteria for normal and shutdown operations are illustrated on Fi gures 12.3-1 through 12.3-4 and 12.3-6 through 12.3-9.

These figures do not represent operational requirements.

Radiation shielding is provided on the basis of maximum concentrat ions of radioactive materials within each shield region (e.g., 1 percent failed fu el at the original desi gn basis core power level of 3636 MWt) rather than the annual average valu es. For batch processes, as an example, the point of highest radionuclide c oncentration in the batch process is assumed (e.g., just prior to draining of a tank). The shielding designs are, therefore, intenti onally conservative in that the dose rates reflect maximum, rather than average, sources to be shield ed. These maximum dose rates are based on anticipated occupancy requirements and are set such that the maximum exposure of plant personnel is within the li mits set by 10 CFR 20. The average exposures are expected to be a small fraction of the limiting va lues because it is not expected that the plant would run at 1 percent failed fuel with all tanks full to capacity, all demineralizer beds at saturation, etc.

MPS3 UFSAR12.3-2Rev. 30In computing the dose rates on which the confirmation of shielding thicknesses is based, a number of explicit and implicit conser vative measures are included. Do se points are generally calculated along vertical shield surfaces oppos ite the most intense source in the vicinity. These calculations are based on the inherent assumption that plant personnel spend th e required time in each zone in contact with the shield at this point. This is a demonstrably conservative approach, since the dose rate actually decreases dramatically as the dose points are moved along the surface of the shield due to the slant penetration invo lved. The additional reduction of intensity with distance is also ignored by this approach.

The shield wall thicknesses are derived from desi gn basis fuel defect of 1 percent and dose rate limitation of adjacent zones and are expected to provide adequate protection for abnormal conditions which may occur dur ing normal plant operations.Zone designations are based on th e annual occupational exposure limits, access requirements and occupancy time for the specific location in the plant as described in Table 12.3-1.

For the yard areas, the shield wa lls are designed to meet the Zone I criterion of 0.25 mRem per hour. The most significant structur es which contribute to the yard dose rate are the containment, fuel building, waste disposal bui lding, auxiliary building, refue ling water storage tank, and the boron recovery tanks.

The calculated dose rate levels in the unrestric ted areas are based upon full power normal plant operations assuming fuel defects producing e xpected quantities and concentrations of radionuclides consistent with NUREG-0017. At the site boundary, the calculated dose rate is approximately 0.43 mRem per year.

Dose rates are generally calcul ated at three and six foot le vels above walking surfaces, particularly if significa nt sources are located on the next level above the zone within the building.

Dose rates for post-shutdown conditions are com puted at the earliest reasonable time after shutdown. Subsequent decay is ignored for conservatism; i.e., the dose rate at that point in time is quoted despite the fact that the radiation le vels continue to decay to lower values.Transit times in coolant loops, etc, are computed as precisely as practicab le with no intentional conservatism. Simplified models which are used to describe large components are intentionally devised not to overestimate com ponent self-shielding. This provi des an amount of conservatism which varies from component to component. In some instances no com ponent self-shielding is included.The modeling reflects the knowledge of specific components and limitations which exist when details on components are not available. The models allow for this uncertainty in a conservative manner, thus ensuring that the actual radiation l eakage from the supplied component is less than or equal to predicted values.

Shielding in the Millstone 3 plant was designed using Stone & Webster Engineering Corporation's topical report, Ra diation Shielding Design and Analysis Approach for Light Water MPS3 UFSAR12.3-3Rev. 30 Reactor Power Plants (RP 8A) (SWEC 1975). Th is approach, recommended for guidance by the NRC in the Standard Review Plan, Section 12.3, incorporates the design features offered in Regulatory Guides 1.69 and 8.8. RP 8A defines the assumptions, codes, techniques, and parameters used in calculating shield thickness, material, and placement.12.3.1.1Primary Shielding

Primary shielding is provided to limit radiation emanating from the reactor vessel.

The primary shield is designed to: 1.attenuate neutron flux to minimize activa tion of plant compone nts and structures;2.reduce residual radiation from the core to a level that allows access to the region between the primary and secondary shield s at a reasonable time after shutdown; and3.optimize the combination of primary a nd secondary shielding by reducing the radiation level from the reactor so that it is commensu rate with radiation levels from other sources.

The primary shield consists of a water filled ne utron shield tank and a 4.5 foot thick reinforced concrete shield wall. The neutron shield tank has an annular thickness of 3 feet and is located between the reactor vessel and the concrete shield wall. To maintain the integrity of the primary shield, a streaming shield fa bricated from borat ed silicon rubber (Dow Corning Sylgard 170 silicon elastomer or equivalent) is installed in the upper annular ga p between the vessel flange and the neutron shield tank and around th e nozzles. (Refer to Figure 12.3-12.)This shield is designed to minimize the leakage of neutrons to the annular region and streaming to the upper levels of the containment, thus reducing the neutron dose rate on the operating floor, during normal operations, to acceptable levels.

It was estimated that the neutron dose rate in th e annulus area between th e containment wall and the crane wall at the operating floor level would not exceed 5 mR em per hour with the shield in place. Radiation protection surveys should be cons ulted to determine actual neutron dose rates.12.3.1.2Secondary Shielding Secondary shielding consists of reactor coolant loop shielding, the crane wall, containment structure shielding, fuel handling shielding, auxiliary equipment shielding, waste storage shielding, control room shielding, and yard shielding.

Secondary shielding thicknesses w ithin the containmen t structure are based on nitrogen 16 being the major source of radioactivity in the react or coolant during normal operation. This source establishes a required shielding thickness of th e reactor coolant loop shielding, crane wall, and containment structure wall. The shutdown radiation levels in the reactor coolant loop cubicles are MPS3 UFSAR12.3-4Rev. 30 established by the activities of th e activated corrosion and fission products in the reactor coolant system.The crane wall provides shielding for limited access to the annulus between the crane wall and the containment structure wall a nd provides additional exterior shielding during power operation.

The containment structure shielding consists of a steel lined reinforced concrete cylinder and hemispherical dome. This shielding, together with the crane wall, attenuate s radiation during full power operation and during an accide nt. This shielding keeps radiat ion levels within acceptable levels at the outside surface of the containm ent structure and at the exclusion area boundary (EAB).The fuel handling shielding, including both water and concrete, attenuates radiation from spent fuel assemblies, control rods, and reactor vessel internals to acceptable levels and permits the removal and transfer of spent fuel and contro l rods to the fuel pool in the fuel building.The refueling cavity above the reactor is formed by a stainless steel-lined, reinforced concrete structure. This refueling cavity becomes a pool when filled with bora ted water to provide shielding during the refueling operation.

The depth of the shielding water in the cavity is su ch that the radiation dose rate at the surface of the water from a spent fuel assembly should not exceed approximately 2.5 mRem per hour during the short time intervals when the fuel handling operation brings the spent fuel assembly to its closest approach to the pool surface.The cavity is large enough to provide storage space for the upper and lower internals and miscellaneous refueling tools.

The fuel pool in the fuel building is filled w ith water for shielding as discussed in Section 9.1.4.3.4. The fuel pool walls are a minimum of 6 foot thick concrete to ensure a dose rate less than 0.75 mRem per hour outside the fuel buildi ng and less than 2.5 mRem per hour inside both the fuel building and the adjacent auxiliary building from the fuel stored in the pool.In order to preclude unacceptable radiation dose rates during fuel transfer , a special radiation shield, fabricated from carbon steel has been provided inside containment, where the fuel transfer tube traverses the gap between th e containment wall and the refu eling cavity wall. (Refer to Figure 12.3-11). The design basis for the shielding concept is a dose rate at the surface of the shield of approximately 50 mRem per hour and a dose rate at the personnel access hatch of approximately 5 mRem per hour.

Outside containment, the fuel transfer tube is inaccessible to personne l by means of backfill covering the transfer tube and a security fenc e between the containment and the fuel building, assuring limited access to this area.

Three radiation monitors with local audible and visible alarms as well as remote alarms in the control room, are used to monitor fuel transfer operations. Two ra diation monitors are located in MPS3 UFSAR12.3-5Rev. 30the passageway and access area adjacent to the fuel transfer tube in the contai nment building. The other monitor is in the fuel building.

Auxiliary building components ma y exhibit varying degrees of radioactive contamination due to the handling of various fluids. The function of shielding in this building is to protect operating and maintenance personnel working near the various a uxiliary system components, such as those in the makeup and purification system, the boron re covery system, the radioactive liquid and gaseous waste systems, and the sampling system.Typically, major components of systems are individu ally shielded so that compartments may be entered without having to shut down and possibly decontaminate the entire system. Potentially highly contaminated ion exchangers and filters are located in i ndividual shielded cells in the auxiliary building. The concrete thicknesses provided around the shielded compartments was based on reducing the surrounding area dose rate to less than approximately 2.5 mRem per hour and the dose rate to any adjacent cubicle to less than approximately 100 mRem per hour.

In some areas, tornado missile protection in the form of concrete affords more shielding than that required for radiation protection.

The waste storage and processing facilities in the auxiliary bui lding and the waste disposal building are shielded to provide protection for operating personnel in accordance with radiation protection design criteria.

Boron recovery tanks, which may be used to store "letdown" prior to its recycling to the plant or processing as waste, are shielded to reduce dose ra tes to accessible levels within the yard area.12.3.1.3Accident Shielding 12.3.1.3.1Containment and Control Room DesignAccident shielding is provided by the containment structure, which is a reinforced concrete structure lined with steel. For structural reasons , the thickness of the cylindrical wall and the dome are 54 inches and 30 inches, respectively. These thicknesses are more than adequate to meet the shielding requirement during accident conditions.

The radiation design objective for the control room shielding helps limit the dose from external sources to personnel inside the c ontrol room to less than 5 Rem during any design basis accident.

This dose includes: (1) the external radiation contribution from the postulated radioactive plume leaking from the containment for a period of 30 days; (2) the 30 day radiation dose from radioactivity inside the containment; (3) the 30 day radiation dose due to post-LOCA leakage from the ECCS located outside of the Millstone 3 containment; a nd (4) the radiation dose due to radioactive components within the control room boundary (e.g., buildup of halogens in filters).

The Millstone 3 Control Room has also been ev aluated for the 30 day dose due to a postulated LOCA at Millstone 2 and those results are within the limits of GDC-19. Shielding calculations show that the 2 foot thick concrete walls which enclose the control room are sufficient to ensure that the radiation dose inside the control room remains below the radiation design basis during MPS3 UFSAR12.3-6Rev. 30 any postulated accident. The control room design includes special treatment of shield wall penetrations and structural details which ensures that this fa cility remains acceptably leak-resistant.12.3.1.3.2Post-Accident Access to Vital Areas A radiation and shielding design review was performed in accordance with NUREG-0737, Action Item II.B.2 (USNRC, 1980), in order to ensu re personnel accessibility after a design-basis accident (DBA). The DBA considered for this evaluation was the loss-of-coolant accident (LOCA). The projected dose to complete each activity necessary to mitigate a DBA LOCA, en route to and in vital areas, is less than the 5 rem design limit of NUREG-0737. At Millstone 3, this requirement is met by providing sufficient shie lding of components containing post-accident radioactive inventories, consistent with anticipated access routes and stay times.Areas requiring accessibility (vital areas) are those areas where post-LOCA actions can be taken over the short-term to en sure the capability of ope rators to control and mi tigate the consequences of an accident. A description of the post-accident activities is summarized below and in Table 12.3-3.1.Locally trip the reactor trip breakers and bypass breakers This action is performed at the 43 foot 6 inches elevation in the auxiliary building MCC rod control area. This is done in the ev ent that the reactor fa iled to trip. This action must take pla ce as soon as possible. Thus, th e 0 to 30 minute time frame is assumed. While this step is done only in the event of an ATWS (beyond the design basis scenario), it is conservatively included as a required operator action.2.Local actions needed to realign Spent Fu el Pool Cooling, RBCCW and Service Water for spent fuel pool cooling FSAR 9.1.3.3 states that spent fuel pool cool ing will be initiat ed approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the LOCA. This requires operator action in the spent fuel pool building.

The 2 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time frame is assumed.3.Powering the Plant Process ComputerThe Plant Process computer is normally not powered from an Emergency Bus. It is powered from an uninterrupt ible power supply that ma y last for only 30 minutes.

Thus, the 0 to 30 minute category is assume

d. The plant process computer is used for SPDS and OFIS. In order to restore power to the plant process computer, MCC 32-3T is energized on the 38 foot level in the turbine building.4.Powering the SI accumulator valvesFor post-LOCA cooldown and depressurization, the SI accumulator isolation valves are closed to prevent injection of nitrogen that might interrupt natural MPS3 UFSAR12.3-7Rev. 30 circulation. It is necessary to repower th e valves from the 24 foot 6 inch level in the auxiliary building. Since this would be done only after the plant is stabilized in preparation for a cooldown, the 30 minut e to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame is assumed.5.Initiate hydrogen monitor FSAR Section 6.2.5.2 states that this syst em will be available to provide continuous monitoring within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a nd 30 minutes of an accident. For dose consequence evaluation, availability within 30 minutes was assumed for conservatism. Thus, the 0 to 30 minute category is assumed. Access to the hydrogen recombiner building is needed in order to initiate hydrogen monitoring.6.Deleted7.Deleted8.Repower Monitor and Maintain the por ous concrete groundwater removal system A non-safety related pump (3SRW-P5) is credited with groundwater removal that circumvents the waterproof membrane th at surrounds the containment structure and the containment struct ure contiguous buildings.*3SRW-P5 is normally powered from 32-4T. If "A" Train Emergency Bus is not able to supply power to 32-4T, then 3SRW-P5 can be repowered from 32-3U, "B" Train Emergency Bus. It is estimated that repower may take 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. It is expected that for a de sign basis LOCA, this step is reached before 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The performance of this action is based on the radiological conditions near the RWST, which requi res work outside the ESF building to be completed between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.*The status of the groundwater rem oval system will be monitored and operated several times a day. These activ ities take place in the yard on the north side of the Refuel Water Storage Tank (RWST) at local panel 3SRW-CSP5. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time frame and beyond is assu med for monitoring. Due to dose considerations near the RWST as the accident progresses, the activities at panel 3SRW-CSP5 may need to be completed in as little as 2 minutes.*Should the single non-safety relate d groundwater sump pump become nonfunctional, it must be replaced or repaired. Due to dose considerations, the 1 day time frame and beyond is assumed for maintenance and repair activities for the sump pump. The sump pump is accessible from the ESFB roof. Access to the ESFB roof is ac hieved via the Hydrogen Recombiner Building stairway.9.Open the breakers for the non-safety grade sump pumps

MPS3 UFSAR12.3-8Rev. 30 The operation of the non-safety grade su mp pumps may mask the presence of a leak. Thus, the need to secure sump pumps in ECCS pump cubicles and common areas in the Auxiliary and ESF buildings. The 1 day to 4 day time frame is assumed. This action requires access to the 21 foot elevation of the ESF building and the 24 foot 6 inch elevation of the auxiliary building. However, if radiological conditions preclude entry into the ESF or Auxiliary Building, then the associated MCC's may be de-energized at its Load Centers in the 4 foot elevation of the Service Building. Therefore, local opera tor actions in the ESF and Auxiliary Building are not required.10.Align Alternate AFW Pump Suction Source or Replenish Demineralized Water Storage Tank (DWST) Inventory For a small break LOCA, steam generator inventory makeup beyond that provided by the DWST may be required for long term heat removal. Technical Specifications 3.7.1.3, "Demineralized Water Storage Tank" ensures that at least a 13 hour1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> inventory is availabl

e. In the longer term, the AFW pumps can be aligned to the condensate storage tank (CST).Travel Route 8 reflects th e travel route to manual valve 3FWA*HCV37 which is used to realign pump 3FWA*P2 to the CST.

Thus, the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> category is assumed. This action is performed on the 21 foot elevation of the ESF building.11.Reset MCC breakers for Diesel Generator keep warm systems This action is taken when off site pow er is available and the running diesel generator is stopped. The keep warm syst em assures that the diesel generator would be maintained in the optimum condi tion for a subsequent start if a loss of off site power occurs later in the transient. This action is performed in the emergency diesel generator building.

In addition to the areas and activities defined above, the Control Room and Technical Support Center (TSC) require post-accident access and continued occupancy as discussed in NUREG-0737.Post-accident control room habitability is discussed in Section 6.4. The post-accident dose consequences for the Control Room are presented in Table 15.0-8.

The potential radiation doses to a person occupyi ng the TSC have been evaluated for the Unit 3 LOCA. The TSC is designed for continuous operat ion for the duration of the accident (i.e., 30 days). The building roof and walls provide adequa te shielding to protect the occupants against direct radiation from the external radioact ive cloud and from the containment during the postulated LOCA. Double vestibul e doors are provided at the building entrance to minimize inleakage due to personnel ingress/egress. The TS C ventilation system is described in Section 9.4.13. The evaluated 30 day integrated dose for an individual occupying the TSC following the DBA is within the NUREG-0737 criteria of 5 rem whole body dose or equivalent.

MPS3 UFSAR12.3-9Rev. 30It has been determined that post-accident access to areas addressed in NUREG-0737, which have not been identified above, is not required for Millstone 3.Table 12.3-4 provides an estimate of the anticipated times after a LO CA that vital area access is required, with consideration given to the typical 30 minute minimum time frame assumed for operator action outside control room and the X/Q intervals assumed in the FSAR Chapter 15 accident analysis. Outside travel routes are shown on Figure 12.3-10 and are listed onTable 12.3-3. A general description of the ingres s travel routes, primar y and alternates, are described below (the egress path is the same as the ingress path except for alternate routes to the backup Chemistry Laborator y in travel route 4).Travel route 1:The primary route is from the c ontrol building through the service building to the auxiliary building (no outdoor travel). The alternate r oute is from the control building to the service building to the exit between the service and auxiliary building to the north entrance to the auxiliary building.Travel route 2:The primary route is from the c ontrol building through the service building to the exit between the service building and the auxiliary building, along the north side of the waste disposal building then south to outside of the ESF Building, (inside the Radioactive Materials Area fe nce). The alternate route is from the control building to the service building to the turbine building to the RR loading area, east along the roadway past the RWST to the outside of the ESF building.Travel route 3:The primary route is from th e control building through the service building corridor leading to the roadway beside the MSV building to the RCA gate south of the hydrogen recombiner building (HRB) and into the HRB. The alternate route is from the control building to the service bui lding to the turbine building to the RR loading area to the RCA gate adjacent to the HRB.Travel route 4:The primary route is from th e control building through the service building corridor leading to the roadway beside the MSV building to the RCA gate south of the hydrogen recombiner building (HRB) and into the HRB. The alternate route is from the control building to the service bui lding to the turbine building to the RR loading area to the RCA gate adjacent to the HRB. The sample analysis is performed in the MP3 chemistry lab wh ich is on the egress path. Figure 12.3-10, sheet 4, provides two additional routes for sample analysis in the MP1/MP2 service building.Travel route 5:The primary route is from the c ontrol building through the service building to the exit between the service building and the auxiliary building, following the roadway north and east of the waste di sposal building, then entering the fuel building. The alternate rout e is from the c ontrol building through the service building corridor to the turbine building to the RR loading area, then east along the roadway past the RWST, north to the fuel building.

MPS3 UFSAR12.3-10Rev. 30Travel route 6:The primary route is from the control building to turbine building 38 foot level. No alternate is given since doses would only increase.Travel route 7:The primary route is from the control building to the emergency diesel generator building. No alternate is given since dos es would only increase with any other route.Travel route 8:The primary route is from the c ontrol building through the service building to the exit between the service and auxiliary bui ldings, along the north side of the waste disposal building, south to the ESF build ing or the turbine driven Auxiliary Feedwater Pump Room. The alternate rout e is from the control building through the service building corridor to the turbine building to the RR loading area, then east along the roadway past the RWST and into the ESF building or the turbine driven Auxiliary Feedwater Pump Room.Travel route 9:The primary path is from the c ontrol building through the se rvice building corridor to the turbine building auxiliary bay, lower level then across the road to the auxiliary building. The altern ate path is from the contro l building to the service building, past the Chemistry Laboratory, ex it the service building to the auxiliary building.The following general assumptions and criteria are used as a basis for review of all vital areas and access routes as applicable:1.The starting point for all activitie s is the Unit 3 Control Building.2.In order for an access/egress pathway to be considered acceptable, the total dose for activities required for mitigation of th e design basis accident (which includes the dose to perform the activity and the associated transit dose) must be no greater than the 10 CFR Part 50 Appendix A GDC-19 or 10 CFR 50.67 dose criteria. The determination of total dose is based on the earliest time post-LOCA when access to the designated vital area is required as identified in Table 12.3-4.3.All calculated outside pathway doses are assumed to be comprised of contributions from (a) containment radiation (both dir ect shine and skyshine contributions) and (b) direct radiation fr om the overhead plume.

Gaseous and liquid LOCA source terms used in the review are not less than that stated in NUREG-0737,Section II.B.2, which provides the minimum source terms to be used for evaluation of the adequacy of radi ation protection to the operators.To determine post-accident doses to personnel fo r performance of and transit to identified activities, the following sources of radiation are considered.1.Auxiliary Building MPS3 UFSAR12.3-11Rev. 30*Radiation from containment atmo sphere shining through electrical penetrations.*Radio iodine buildup in the SLCRS filter.

• Sump water in the safety injecti on system piping located below the elevation 24 foot 6 inches floor.*Containment atmosphere shine through the personnel hatch and surrounding walls and floors.*Sump water in safety injection and charging system piping and associated shine through walls and floors.2.Fuel Building*Direct shine from containment*Plume shine*Shine from the RHR heat exchanger in the EFS building.

• Shine from the fuel pool cooling pumps.3.ESF Building*Shine from RSS and SIH piping

• Shine through the wall from the Recirculation Coolers*Shine from RWST piping*Shine from Auxiliary Steam piping4.Along routes from control building to the vital areas.*Skyshine from containment*Direct shine from containment.

• Plume shine.

• Direct shine from the RWST.Systems containing sources of radiation which are identified in NUREG-0737 but which have not been identified in buildings discussed above are c onsidered to be either irrelevant following an MPS3 UFSAR12.3-12Rev. 30accident or negligible contributors to personnel exposure following an accident. For example, the GWS system is a negligible contributor of radiation following an accident because when an accident occurs, the only use of this system is for post-LOCA hydrogen purge as the result of a beyond design basis event.

The results of the dose calculations indicate that the plant shielding and design provide adequate protection to operators followi ng a design basis LOCA to ensure compliance with the NUREG-0737 design dose requirements.12.3.2FACILITY DESIGN FEATURES The Millstone 3 design is consistent with th e guidance presented in Regulatory Guide 8.8, Revision 4, C2, which discusses specific features in the facility and equipment design that limit radiation exposure to levels that are ALARA.

The following features ha ve been incorporated.12.3.2.1Location and Design of Equipment to Minimize Service Time In the auxiliary building, nonradio active equipment, such as the reactor plant component cooling system and components used to process the waste evaporator distillate, are located outside high radiation cubicles in areas designated as Radiation Zones II or III (defined in Table 12.3-1). In the containment structure, nonradioactive equipment requiring servicing is typically located in Radiation Zone IV areas. Exceptions include thos e components attached to the reactor coolant system, such as the reactor coolant pump moto r cooling equipment and the equipment support snubbers.Major radioactive components which may require se rvicing are typically located in individually shielded cubicles. These cubicles are designed su ch that radiation contri butions from adjacent cubicles is small compared to sources within the cubicle. The resultant dose rate in any cubicle in which equipment is being serviced is due to sources within the cubicle to radiation penetrating through shield walls from adjace nt cubicles, and to radiati on streaming through shield wall penetrations. The design basis for shield walls enclosing cubicles containing process equipment is discussed in Section 12.3.1. Shie ld wall arrangement and di mensions are shown for the Containment Building, Figure 3.8-60; Auxiliary Building, Figur e 3.8-62; Fuel Building, Figure 3.8-63; and Waste Dispos al Building, Figure 3.8-74.Cubicle access openings generally incorporate a labyrinth design which precludes direct radiation shine. The openings are sized to allow for removal and replacement of minor fluid system components such as pumps and valves, as well as to provide access for maintenance equipment.

For example, pump cubicle openings for all horizontal pumps except the charging pumps are of sufficient size to skid such pumps through the entranceway. The openi ngs of other cubicles containing equipment requiring servicing are sized to allow the passage of components while still maintaining radiation safety conditions. Cubicles are sized to allow sufficient clearance around equipment for laydown of equipm ent and installation of tempor ary shielding as needed. The equipment service requirements for pull space and laydown space are provided within each cubicle, thus eliminating the need for dismantling of piping other than that directly connected to the equipment.

MPS3 UFSAR12.3-13Rev. 30 The corridor system is sized to allow hand car t and dolly access. Motor terminal boxes and other terminal boxes are located so as not to block acce ss, and are separated from radioactive piping if possible. Platforms for servicing specific components are provided where necessary.

Certain components have design features which mi nimize service time. For example, the reactor coolant pump design includes an assembled cartridge seal which results in reduced time required for replacement. The cartridge seal is also expected to have a us eful life which is double that of the older designs. The reactor coolant pump desi gn also includes a spool piece to facilitate separation and replacement of the motor from the pump.The reactor vessel nozzle welds insula tion is fabricated in sections wi th a thin reflective metallic sheet covering and quick disconnect clasps to facilitate removal for in spection of the welds.Typically, filters are designed to be removable from the top with li fting bails in the middle of the head. The filter assemblies usually have bolt lead-ins for tool entry, and the filters are contained in disposable cartridge assemblies. These features facilitate remote removal, disposal, and assembly.

The head closure system provided for Mill stone 3 includes quick disconnect/connect stud tensioners which have quick-act ing, hydraulically-operated stud gr ipper devices, as opposed to conventional tensioners which must be threaded onto the tops of the studs. Also provided are air-motor driven stud removal tools which can rapidly re move (or insert) the studs, in contrast to the much slower manual stud removal and insertion t ools used in older designs. The stud tensioners are designed to operate simultaneously, as are the stud removal tools.The primary system heat exchange rs are designed such that the shell-to-tube sheet joint need not be broken for inspection. The shell and tube assembly can be lifted intact above the channel head to expose the tube ends for inspection and leak testing.

Pumps are typically designed with flanged connections to facili tate removal for maintenance.

Depending on expected conditions, either canned pumps or pumps with high quality mechanical seals are used to reduce leakag e and maintenance requirements.12.3.2.2Location of Instruments Requiring In Situ Calibration Instruments which require in situ calibration are located, wherever possible, on exterior walls of shielded cubicles to minimize exposure of in strumentation and personnel. Instruments which cannot be located in this manner are located in the lowest practicable radiation area in the cubicles and are provided with convenient access. Where practical, instru ments are designed for removal to low radiation areas for calibration and maintenance.12.3.2.3Location of Equipment Requiring Servicing in Lowest Practicable Radiation Field (or Movable to Lowest Practicable Radiation Field)

As indicated above, radioactive equipment requiring servicing is typically located in shielded cubicles with access openings sized for ease of equipment removal. As an example, pump cubicles are designed to allow removal of the pump to the lowest practicable radiation field.

MPS3 UFSAR12.3-14Rev. 30Westinghouse has designed the Model F steam generators to reduce the radiation exposure during both normal operation and maintenance. The tube e nds are designed to be flush with the tube sheet in the steam generator channel head to el iminate a potential crud trap. The steam generator manways (entrance to channel head) are sized to facilitate entrance and exit with protective clothing. Handholes to the secondary side are positioned to faci litate maintenance operations.

Changes to increase steam generator reliabilit y also reduces occupati onal radiation exposures. Such changes include improved steam generator tube support plates (stainless steel and quatrefoil flow holes) and the use of all-volatile tr eatment chemistry on the secondary side.12.3.2.4Valve Location and SelectionValves are located in separate shielded valve cubi cles or areas outside equipment cubicles to the greatest extent practicable to minimize maintenance exposure. Valv e selections are usually based on "best product" available and maintenance time required. Westinghouse has supplied valves of the bolted body-to-bonnet forging type. This permit s the use of ultrasonic testing in place of radiography for inspection and facilitates assembly and disassembly, resulting in reduced inspection and maintenance time. Additionally, manual valves under 2 inches in diameter are designed for zero stem leakage.12.3.2.5Penetrations of Shielding and Containment Walls by Ducts and Other Openings There are numerous piping penetrat ions through shield walls in th e auxiliary building which are directed into adjacent cubicles, into the pipe ch ases for radioactive piping, and into the corridors for nonradioactive piping. To the greatest exte nt practicable, penetrations through walls separating higher radiation zone areas from lower radiation zone areas are located above head level, in corners, and in positions which are offs et from radiation source s in the higher radiation zone cubicles. This prevents line-of-sight radiation streaming from significant radiation sources to personnel working in adjacent cubicles. Notewort hy examples of this practice are provided as follows.1.Electrical penetrations through shield walls are made to prevent direct line-of-sight to any significant radiation sources.2.Instrument tubing penetrations through sh ield walls are made so as to prevent direct line-of-sight to any significant radiation sources.3.Ventilation duct penetrations through shield walls are made at the highest possible elevation and at locations which minimize direct line-of-sight to significant radiation sources. Where direct line-of-sight penetrations through cubicle walls are unavoidable, penetration shields are often employed either inside or outside such cubicles.12.3.2.6Radiation Sources and Occupied Areas Radiation sources (Section 12.2) are separated, as far as is practicable, from normally occupied areas by shield walls and cubicl es. Piping runs are also located as far as practicable from MPS3 UFSAR12.3-15Rev. 30 equipment cubicles. Radioactive piping (e.g., process piping carryi ng radioactive materials) is typically located behind shielding and also routed around, rather than through, normally occupied areas wherever practicable. Valves are located in shielded valve areas where practical and are separated from equipment cubicles, pi peways, and areas of general access.

Physically locked barriers (i.e., locked doors) are provided for areas havi ng radiation levels in excess of criteria as specified in the Technical Specifications.12.3.2.7Minimizing Spread of Contamination and Facilitation of Decont amination Following SpillsTypically sources of cont amination from leaks or spills from components located in cubicles are prevented from spreading by cubicle entrance dikes and/or low point drains to enclosed collection sumps. Floor surfaces and walls are sealed or painted as require d with a protective chemically-resistant coating to provide a surface which is easily decontaminated. Demineralized water hose stations are provided throughout the auxiliary building to allow flush water to be available to each cubicle in the auxiliary building.

Systems containing radioactive flui ds are usually fabricated of corrosion-resistant materials.

Airborne contamination is kept from spreadin g by ventilation systems which are described in Section 12.3.3. A personnel decontamin ation area is located in the radiation protection area. An equipment decontamination area is located in the waste disposal building.12.3.2.8Piping to Minimize Bu ildup of Contamination Interior surfaces of systems in radioactive liquid service typically are made of stainless steel or other corrosion-resistant material.The piping associated with thes e systems is normally routed to avoid sharp bends by carefully selecting the elevation between points and by attempting to run th is piping at no more than two elevations between these points. Pockets and low points are also avoided. Pipe runs for spent resin sluicing are provided with large radius bends rath er than welded elbows to prevent accumulation of resin fines and crud particles.

Resin piping is also butt-welded where possible to minimize the potential for crud particles. Valve stations are designed to minimize the buildup of crud by minimizing the number of pockets and stagnant ve rtical legs.Ventilation design features to minimize radioactiv e contamination buil dup are discussed in Section 12.3.3.12.3.2.9Flushing or Remote Chemical Cl eaning of Contaminated Systems Means for flushing and draining of potentially highly radioactive tanks, lines, and other components are considered in fluid system design. Waste collection tank design includes provisions for internal flushing with spray nozzles to remove potential coll ections of particulate material. All heat exchangers are provided w ith chemical cleaning connections which are connected prior to servicing.

MPS3 UFSAR12.3-16Rev. 30 Flushing and vent connections ar e provided to allow flushing of piping systems for maintenance.12.3.2.10Ventilation Design The ventilation systems are designed with sufficient capacity to control airborne radioactivity releases and concentrations during normal and maintenance c onditions. The ventilation flow through equipment cubicles is based upon unrestricte d air flow from general access areas into these cubicles. The design of ventilation syst ems typically ensures a positive flow from non-contaminated areas to potentially contaminated areas to prevent the spread of airborne radioactivity and to exhaust from the potentially contaminated areas. A more explicit description of ventilation systems is given in Section 12.3.3.12.3.2.11Radiation and Airborne Contamination Monitoring Area and airborne radiation monitoring ensure s that any substantial abnormal radioactivity release is promptly detected. The area and airborne radiation monitoring system is described in more detail in Section 12.3.4 and Section 11.5, respectively.12.3.2.12Temporary Shielding The use of temporary shielding to facilitate maintenance tasks is considered on a case-by-case basis. Convenient means for transport and plac ement of such shielding are provided by access corridors and elevators in the auxiliary building and an elevator in containment.12.3.2.13Solid Waste ShieldingAs shown on Figures 12.3-1 through 12.3-4, radioactive wastes in tanks, evaporators, process gas charcoal bed adsorbers and associated equipment are located in shielded cubicles. Solid waste is shielded both by the storage area walls in th e waste disposal build ing and by individual transportation shields.12.3.2.14Remote Handling Equipment As noted in Section 12.3.2.1, filter s are designed for remote removal, disposal, and assembly.

Equipment is provided for filter handling as well as for remote removal and replacement of ion exchange resins.Stations for potentially radioactiv e system valves are, in general, arranged either in segregated shielded cubicles away from the equipment serv ed and/or are provided with reach rods. The demineralizer and filter valves are in cubicles beneath the vess els and are also provided with reach rods.12.3.2.15Maximum Expected Failures of Fuel Element Cladding and Steam Generator Design features such as shielding and radiati on zones accommodate 1 percent fuel defects and primary to secondary steam generator tube leaks of 1,370 lb per day.

MPS3 UFSAR12.3-17Rev. 3012.3.2.16Sampling Stations Sample points are provided with sample sinks and ventilation hoods , splash screens, and valves located outside each splash screen. Samples are provided with recirculat ion paths behind shield walls at sample sinks, with reach rods for operators.12.3.2.17Cobalt Impuri ty SpecificationsCobalt weight percentages for materials in contact with reactor coolant are considered in purchase specifications.12.3.2.18Reactor Cavity Filtration System

During refueling, the reactor cavity water may become turbid, making it difficult to observe the removal and replacement of fuel assemblies. The portable reactor cavity filtration system, consisting basically of a pump and four filters, provides capability for cleanup of this water, thus minimizing the time required for, and dose due to, fuel and equipment handling operations.12.3.3VENTILATIONThis section provides the basis for the original plant ventilation design. A lthough current airborne levels may not be consistent with the tables in this chapter, these tables are not being changed to be current as that would make them inconsistent with the design basis criteria for the ventilation systems. Recent radiation protection surveys should be consulted for information on current radiological conditions. 12.3.3.1Design Objectives The function and design bases of the ventilation systems are given in Section 9.4. Consistent with these, the following specific objectives pertain to radiation protection and the commitment that occupational radiation exposures are ALARA, in accordance with Regulatory Guide 8.8.1.The airborne radioactivity concentrations from radioactive sources released into the fuel building and turbine building, as shown in Table 12.2-11, are small fractions of values in Column 1, Tabl e 1 of 10 CFR 20, Appendix B. Radwaste piping system and process components in the auxiliar y building and the waste disposal building are separated from normally accessed areas by walls, and are provided with ventilation systems which supply air from clean, occupied areas and exhaust from duct openings located within the process system cubicles. Ultimately, routine plant surveys by plant radiati on protection personnel provide appropriate controls and protective measures desc ribed in Section 12.5.3 when access is needed to areas which ar e not normally occupied.2.Concentrations in areas accessible to administrative personnel are less than 25 percent of the concentrations given in Column 1, Table II of Appendix B to 10 CFR 20.

MPS3 UFSAR12.3-18Rev. 303.The airborne concentrations in all plant areas are ALARA.4.The containment atmosphere filtration system, with only one of its two 12,000 cfm fan units in operation, is capable of reduci ng the airborne iodine concentration in the containment atmosphere to below 1 EC of I 131 in less than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of filter operation under the conditions of expect ed reactor coolant radioactivity concentration and leakag e described in NUREG-0017.5.The containment purge air system is capa ble of reducing airbor ne radiation levels in the containment to acceptable levels prior to and during extended personnel occupancy of the containment.6.The fuel building ventilat ion system operates in th e once-through mode without recirculation with the provision to exhaust through charcoal filters.7.Typically, air flow within the auxiliary, waste disposal, and fuel buildings during normal operation is from areas of lower to higher potential airborne contamination and then to monitored vents with provisions for terminati ng or filtering the ventilation flow upon a hi gh radioactivity alarm.8.Systems are designed so that filters containing radioactivity can easily be maintained to minimize the radiation dose to personnel.12.3.3.2Design Description Detailed descriptions of the ventilation systems for the plant buildings which contain radioactivity or potentially radi oactive systems are given in the following sections:

SectionTitle9.4.1Control room area ventilation system9.4.2Spent fuel pool area ventilation system9.4.3Auxiliary building ar ea ventilation system9.4.5Engineered safety fe ature ventilation system9.4.4Turbine building area ventilation system9.4.7Containment ventilation 9.4.9Waste disposal building ventilation system9.4.10Main steam valve bui lding ventilation system9.4.11Hydrogen recombiner building ventilation9.4.13Technical Support Center building ventilation system MPS3 UFSAR12.3-19Rev. 3012.3.3.3Personnel Protection Features The recommendations of Regulator y Guide 1.52, as described in S ection 1.8, are implemented in the design of the safety-related ve ntilation filter trains to help assure that occupation radiation exposures from service of these trains are ALARA.This is accomplished by utiliz ing the following criteria.1.Each filter train is housed in a shielded compartment, room, or cubicle except for the control building filters which occupy a common cubicle with the air conditioning unit.2.Adequate aisle space is provided for both personnel and equipment adjacent to the service side of the filter trains, and above those sections which require top access (i.e., charcoal adsorber).3.Convenient and accessible passageways and co rridors from the filter trains to the elevators and equipment hatc hes are provided for transpor t of replaceable filter train components and the equipment used in accomplishing their replacement.4.Replaceable elements, except for most downstream HEPA filters, are designed for ready removal from the clean filter si de, and minimal radiation exposure of personnel. A portable cart-mounted vac uum conveying system is provided for draining and recharging gasketless-type char coal adsorbers. Co ntaminated filters can be transported in shielded containers if necessary.5.Rigid, hinged access doors are provided in accordance with ANSI N 509 for man-entry filter trains.6.HEPA and prefilter arrangements are no mo re than 3 elements high to facilitate easy replacement without the use of ladders, temporary scaffolds, or platforms.7.A minimum of 2.5 linear feet is main tained from mounting frame to mounting frame between banks of components for removal of filter elements.8.Adequate vapor-tight lighting is provided on each side of the filter banks for man-entry filter trains.9.(Deleted) 10.Drains are provided to convey water from moisture separators, maintenance or fire protection discharge out of the filter train.11.Permanent test fittings are provide d for initial and periodic field testing.

Filter train arrangement is discussed in Section 6.4.

MPS3 UFSAR12.3-20Rev. 3012.3.3.4Radiological Evaluation Concentrations of airbor ne activity for the expected and design conditions in the containment structure, turbine building, and fuel building are tabulated in Table 12.2-11. The concentrations based upon design conditions are expected to enve lope anticipated operati onal occurrences. The airborne concentrations are averages based on assumed total leak rate s described in NUREG-0017 and the ventilation rates for the respective buildings. Corridors and areas normally occupied by operating personnel are expected to have negligible airborne activity concen trations since clean air ventilation flow is typically directed from areas with less potential for contamination (manned areas) to areas with gr eater potential for contamination.

Equipment cubicles are the most likely areas fo r airborne concentrations but are not normally occupied or accessible without prior survey and control. For purpose of quantification, the worst airborne concentration could conceivably exist in cubicles for which the combination of relatively high system volatile radionuclide concentrations and low cubi cle ventilation rate would simultaneously exist for a given leak rate. A cubicle such as the letdown heat exchanger cubicle in the auxiliary building could develop airbor ne concentrations of approximately 7x10

-5 Ci/cc assuming all the design basis leak rate takes place in that cubicle for expected coolant radioactivity concentrations.

Based on the above assumption, it is expected that other c ubicles would have airborne concentrations of less than 7x10

-5 Ci/cc.Section 12.2 includes the models and parameters used as a basis for calculated radioactivity values.12.3.4AREA RADIATION AND AIRBOR NE RADIOACTIVITY MONITORING12.3.4.1PurposeThe area radiation monitoring system (RMS) works in conjunction with the process, effluent, and airborne radiation monitoring group (Section 11.5). Its purpose is to protect plant personnel by measuring levels of radiation in various areas of the plant. It also provides a warning to operations of abnormal radiological conditions.

If high radiation levels are m onitored, the system sounds an alarm. It also produces a record of radiation levels.12.3.4.2System Design

The basis of the RMS area radiation monitoring group is the single channel, GM tube or ion chamber detector equipped with a dedicated mi croprocessor except the containment high range monitors which are analog and in designated cases, a rate meter. The microprocessor provides local display and control functions for the detector, computes and stores time-averaged detector outputs, stores all necessary opera ting parameters (e.g., alarm trip values), and also handles all communication between the detector and the RMS computer system. The rate meter, where provided, is located adjacent to the detector and provides local analog display. A high activity MPS3 UFSAR12.3-21Rev. 30 level is indicated by both audibl e and visible alarms which ma y be acknowledged at the rate meter. Area radiation monitors are located in normally accessible areas where changes in plant conditions could cause significant increases in personnel exposure rates in accordance with design criteria establishe d in ANSI/ANS HPSSC-6.8.1-1981.The RMS computer system provides centralized display and control, at the control room RMS workstations. A dual server based computer system, located in the control building, polls each monitors microprocessor every several seconds to obtain the latest readings, and to register any alarms present. All alarms are displayed on the control room RMS workstation and can be printed. Radiation alarms are annunciated both in the control room and if equipped with local indication locally at the microprocessor, and can be acknowle dged at either location. The system operator also uses the RMS workst ations to either output data fr om individual monitors or input commands to these non-Class 1E monitors. All commands sent are recorded in the message summary log. The last (30) 1 minute, 10 minute a nd hourly averages are stored and available for review at the RMS workstation for all radiation monitors excep t the containment high range. An RMS workstation is located in the radiation protection Office.

In addition, those monitors designated Class 1E except that containment high range monitors are connected directly to one of two control room 1E cabinets. The output of each monitor is digitally displayed and also recorded. A remote indicati on and control module (RIC) is furnished in the cabinets for each 1E monitor. The RIC handles al l remote control functions for 1E monitors. The containment high range monitors are displayed on the 1E control room cabinets and recorded on the plant process computer. The 1E cabinets are connected by electronic isolators to the RMS computers to allow data from the 1E monito rs to be displayed on the control room RMS workstations and to be written into the RMS computer.

The area RMS is calibrated both by a standard factory calibration a nd by onsite calibration.

Factory calibration included checks for linearity and energy response. Sources traceable to national standards are used. Onsite calibration includes detector response using sources of known energy and strength. The frequency of onsite calibrat ion of safety-related monitors is provided in the Technical Specifications.Table 12.3-2 gives the mark numbers, names, locati ons, and ranges of the area monitors in the RMS. The following paragraphs provide a brief description of the diff erent types of area monitors.12.3.4.3Class 1E Area Monitors Four of the area monitors are designated Cla ss 1E. (Section 8.3.1.1.2 discusses Class 1E power and its backup supply.) These are the two redundant fuel drop monitors and the two containment internal high range monitors. They differ from most other area monitors in that they use ion chamber detectors instead of GM detectors. The high range monitors are capable of withstanding a design basis accident inside containment. The fu el drop monitors are designed to operate in the normal containment environment and are discussed in Section 11.5.

MPS3 UFSAR12.3-22Rev. 30 The containment internal high-range monitors are located on the insi de face of the annulus wall approximately 180

° apart. Range is 1 to 10 8 R/hr.These monitors are qualified based on th e requirement of IEEE 323-1974, 279-1971, and 344-1975.Due to the normally high operational dose ra tes inside containment, the dedicated microprocessors for all detectors located inside the containment structure are located within the auxiliary building.

As with the Class 1E process monitors (Section 11.5), the outputs of these devices are displayed and recorded on the control room Class 1E panels. The display and recording of the containment high range monitor is as required by Regulatory Guide 1.97.12.3.4.4Non-Class 1E Area MonitorsNon-Class 1E area monitors measure and transmit local radiation levels, and annunciate an alarm upon a high radiation level. All monitors except the hydrogen recombiner control room have analog display rate meters located adjacent to the detectors and are powered from normal AC power (Section 8.3.1.1.1). The hydrogen recombiner control room monitor is powered by inverted normal DC power. Their purpose is to protect pl ant personnel from exces sive exposure and to provide a display radiation levels within the plant.12.3.4.5Airborne Radioactivity MonitoringThe process and effluent radia tion monitoring system described in Section 11.5 includes normal range particulate and gas monitors. Their purpose is to monitor airborne radioactivity in areas that may be occupied by plant personnel and to facil itate finding radioactive leaks. These monitors sample air from the reactor containment, the ESF building, the c ontrol room, and from locations in the reactor plant heating and ventilation system upstream of the ventilation vent monitor. They are capable of detecting airborne activated corrosion products and fission products at levels below the derived air concentration of 10 CFR 20. Both the particulate and gas detector channels of these monitors are provided with an "alert level" alarm, in addition to the "high" alarm, with setpoints established by ope rating personnel to allow observation of increases in airborne radioactivity. These moni tors are polled every several seconds by the radiation monitoring computer system. Once elevated readings are noticed, personnel with portable air samplers can determin e which area, associated with the ventilation stream with high radioactivity, c ontains the source of the problem.12.

3.5REFERENCES

FOR SECTION 12.3 12.3-1NUREG-75/087, USNRC. Standard Review Plan, Revision 1.

MPS3 UFSAR12.3-23Rev. 3012.3-2Regulatory Guide 1.52, USNRC. Design, Test ing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, Revision 2.12.3-3Regulatory Guide 1.69, USNRC.

Concrete Radiation Shields for Nuclear Power Plants.12.3-4Regulatory Guide 1.70, USNRC. Standard Fo rmat and Contents of Safety Analysis Reports for Nuclear Power Plants, Revision 3.12.3-5Regulatory Guide 1.97, USNRC. Instrumentation for Light-Water-Cooled Nuclear Power Plant to Assess Plant and Environs Conditi ons During and Followi ng Accident, Revision

2. (Compliance provided in a separate report, Section 1.7.4.)12.3-6Regulatory Guide 8.8, USNRC. Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low as is Reasonably Achievable, Revision 4.12.3-7Stone & Webster Engineering Corporation (SWEC) 1975. Ra diation Shielding Design and Analysis Approach for Light Water Reactor Power Plants, RP-8A, May 1975.

MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:*Based upon the 5 rem per year criterion given in 10CFR20 and the maximum personnel occupancy time corresponding to each radiation zone.TABLE 12.3-1RADIATION ZONESZone DesignationZone DescriptionMaximum Allowable Dose Rate* (mRem/hr) IUnrestricted area - Continuous access0.25IIUnrestricted area - Periodic access - 40 hrs/wk2.5IIIRestricted area - Controll ed periodic access - 6 hrs/wk15IVRestricted area - Controlled infrequent access - 1 hr/wk100VHigh radiation area - Not normally accessible> 100 MPS-3 FSARPage 1 of 2Rev. 30TABLE 12.3-2RADIATION MONITORING SYSTEM - AREA RADIATION DETECTOR LOCATIONMark NumberNameLocation Range mr/hr3RMS-RE01Refueling MachineContainment - 51 feet 4 inches 1-10 53RMS-RE02Fuel Transfer TubeContainment - 3 feet 8 inches1-10 53RMS-RE03Incore Inst. Transfer AreaContainment - 24 feet 6 inches1-10 53RMS*RE04AContainment High Range Internal Containment - 52 feet 4 inches 10 3-10 113RMS*RE05AContainment High Range Internal Containment - 51 feet 4 inches 10 3-10 113RMS-RE06Decontamination AreaFuel - 24 feet 6 inches0.1-10 43RMS-RE07HVAC AreaAuxiliary - 66 feet 6 inches0.1-10 43RMS-RE08Spent Fuel Pit Bridge/HoistFuel - 52 feet 4 inches0.1-10 43RMS-RE09Auxiliary Bldg General (A)Auxiliary - 18 feet 6 inches0.1-10 43RMS-RE10Auxiliary Bldg General (B)Auxiliary - 4 feet 6 inches0.1-10 43RMS-RE11Auxiliary Bldg General (C)Auxiliary - 4 feet 6 inches0.1-10 43RMS-RE12Auxiliary Bldg General (D)Auxiliary - 24 feet 6 inches0.1-10 43RMS-RE13Auxiliary Bldg General (E)Auxiliary - 24 feet 6 inches0.1-10 43RMS-RE14Auxiliary Bldg General (F)Auxiliary - 24 feet 6 inches0.1-10 43RMS-RE15Auxiliary Bldg General (G)Auxiliary - 43 feet 6 inches0.1-10 43RMS-RE16Volume Control Tank Cubicle Auxiliary - 43 feet 6 inches0.1-10 43RMS-RE17Waste Disposal Bldg (A)Waste Dispos al - 4 feet 6 inches0.1-10 43RMS-RE18Waste Disposal Bldg (B)Waste Dispos al - 4 feet 6 inches0.1-10 43RMS-RE19Solid Waste Storage AreaWaste Disposal -24 feet 6 inches0.1-10 43RMS-RE20Sample RoomAuxiliary - 43 feet 6 inches0.1-10 43RMS-RE21LaboratoryService - 24 feet 6 inches0.01-10 33RMS-RE22Control RoomControl - 47 feet 6 inches0.01-10 3 MPS-3 FSARPage 2 of 2Rev. 303RMS-RE24Waste Disposal Bldg (C)Waste Dispos al - 4 feet 6 inches0.1-10 43RMS-RE25Waste Disposal Bldg (D)Waste Dispos al - 4 feet 6 inches0.1-10 43RMS-RE28Fuel Building Pipe RackFuel - 11 feet 0 inches0.1-10 43RMS-RE29Spent Fuel Cask AreaFuel - 52 feet 4 inches0.1-10 43RMS-RE31Fuel Transfer TubeContainment - 24 feet 6 inches0.1-10 43RMS-RE32Containment Sump AreaC ontainment - (-24 feet 6 inches)0.1-10 43RMS-RE33RHR Cubi cle "A" Normal Range ESF - 4 feet 6 inches0.1-10 43RMS-RE34RHR Cubi cle "B" Normal Range ESF - 4 feet 6 inches0.1-10 43RMS-RE35Incore Inst. Thimble AreaContainment - 3 feet 8 inches0.1-10 43RMS-RE36Fuel Pool MonitorFuel - 52 feet 4 inches0.1-10 43RMS-RE37Condensate DemineralizerCond. Polishing

- 14 feet 6 inches0.01-10 33RMS-RE38Regeneration AreaCond. Polishing - 38 feet 6 inches0.01-10 33RMS*RE41Fuel Drop MonitorContainment - 51 feet 4 inches 10 1-10 83RMS*RE42Fuel Drop MonitorContainment- 51 feet 4 inches 10 1-10 83RMS-RE52 Recombiner Control RoomHRB- 24 feet 6 inches 10 1-10 8TABLE 12.3-2RADIATION MONITORING SYSTEM - AREA RADIATION DETECTOR LOCATIONMark NumberNameLocation Range mr/hr MPS-3 FSAR Page 1 of 2 Rev. 30TABLE 12.3-3OPERATOR ACTIVITY LOCATIONS AND TIME DURATIONSActivityLocationApprox. Duration (minutes)Travel Route*1Locally trip the reactor trip breakers and bypass breakers 43 feet 6 inches Auxilary Building (MCC Rod Control)< 592Deleted3Local actions needed to realign Spent Fuel Pool Cooling, RBCCW and Service Water for spent fuel pool coolingSpent Fuel Building< 1554Powering the Plant Process Computer38 feet Turbine Building< 1065Powering the SI accumulator valves24 feet 6 inches Auxilary Building< 596Initiate hydrogen monitorHRB< 3037Deleted8Deleted 9Deleted10Monitor and maintain the porous concrete groundwater removal systema. Repower sump pumpOutside of ESF Building< 902ESF Building 38 feet 6 inches< 102b. Monitor sump pump and operate systemOutside of ESF Building.< 22

c. Maintain/repair sump pumpESF Building roof< 2403 MPS-3 FSAR Page 2 of 2 Rev. 30*Figure 12.3-10 graphically depicts each route by route number.**There are no appreciable dose rates in the Emergency Diesel Generator Building.11Open the breakers for the non-safety grade sump pumps in the ESF and Auxiliary buildings21 foot ESF Building< 152 24 feet 6 inches Auxilary Building< 1514 foot Service Building< 1512Deleted13Align Alternate AFW Pump Suction Source or Replenish Demineralized Water Storage Tank (DWST) Inventory.21 foot ESF Building< 60814Reset MCC breakers for Diesel Generator keep warm systemsEmergency Diesel Generator BuildingN/A **7TABLE 12.3-3OPERATOR ACTIVITY LOCATIONS AND TIME DURATIONSActivityLocationApprox. Duration (minutes)Travel Route*

MPS-3 FSARPage 1 of 1Rev. 30* The starting time of the time frame listed is used for source term decay correction.** Work must be completed between 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.TABLE 12.3-4ACTIVITY INITIATION TIMETime Frame *Activity0 to 30 minutes1Locally trip the react or trip breakers and bypass breakers2Deleted4Powering the process computer6Initiate hydrogen monitor 14 Reset MCC breakers for Diesel Generator keep warm systems30 minutes to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s5Powering the SI accumulator valves7Deleted 9Deleted12Deleted2 hours to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s3Local actions needed to realign Spent Fuel Pool Cooling, RBCCW and Service Water fo r spent fuel pool cooling10 Repower porous conc rete groundwater pump **8 hours to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s10Monitor porous concrete groundwater system13Align Alternate AFW Pump Suction Source or Replenish Demineralized Water Storage Tank (DWST) Inventory1 day to 4 days10Monitor and maintain the porous concrete groundwater system11Open the breakers for the non-sa fety grade sump pumps in the ESF and Auxiliary Building4 days to 30 days8Deleted 10Monitor and maintain the por ous concrete groundwater system MPS3 UFSAR12.4-1Rev. 3012.4DOSE ASSESSMENT This section was applicable during the presta rt-up period as it provi ded estimates of the occupational radiological consequences of futu re operation and the dose to workers during unit construction. Now that Millstone Unit 3 is operational, Annual Reports submitted to the NRC per Regulatory Guide 1.16 should be consulted for data on the station occupational person-rem requirements. Information on desi gn dose rates in unrestricted areas has been moved to Section 12.3-1.

MPS3 UFSAR12.5-1Rev. 3012.5HEALTH PHYSICS PROGRAM The regulatory guides and other references cited in this section were used as basis documentation for the development of the ra diation protection program. Docu mented methods and solutions different from those set out in the guidance have also been incorporated in the radiation protection program.12.5.1ORGANIZATION The radiation protection program is established to provide an effective means of radiation protection for permanent and tempor ary employees and for visitors at the station. To provide an effective means of radiation protection, the radiation protection program incorporates a philosophy from management (Sect ion 12.1.1); employs qualified personnel to supervise and implement the program; provides appropriate equipment and facilities; and utilizes written procedures designed to provide protection of station personnel ag ainst exposure to radiation and radioactive materials in a manne r consistent with Federal and State regulations (Section 13.5).

The radiation protection program is developed and implemented through th e applicable guidance of INPO 05-008, Regulatory Guid es 8.2, Revision 1; 8.8, Revi sion 3; and 8.10, Revision 1-R.

The radiation protection department and line function management implement and enforce the radiation protection pr ogram. Dominion Corporate commitment to the radiation protection program is provided in DNAP-0100, "Dominion Nuclear Operations Standard."

The Radiation Protection Manager shall meet or exceed the qualifications specified in Regulatory Guide 1.8, Revision 1. The Site Vice President will designate the individual or position that will serve in the position of Radi ation Protection Mana ger (RPM) that is required in the Administrative Section of the Technical Specifications for each Unit. Radiation protection technicians meet or exceed the qualifications specified in ANSI N18.1-1971. The radiation protection organization includes radiation prot ection operations, support and waste services.

The radiation protection departme nt coordinates with all stati on, corporate, and contractor organizations to provide radiati on protection coverage fo r all activities that involve radiation or radioactive material. The radiation protection department is orga nized to provide the following services:1.preparation and implementa tion of radiation protection procedures for routine and nonroutine activities associated with th e operation, maintenance, inspection, and testing at the station;2.compliance with regulatory requirements for maximum permi ssible dose limits and contamination control;3.maintenance of a personnel radiation dos imetry program and dosimetry records;4.the surveying of station areas, maintenan ce of survey records, and the posting of survey results for daily activities within the station; MPS3 UFSAR12.5-2Rev. 305.assistance in the station training pr ogram by providing specialized radiation protection training;6.procurement, maintenance, and calibrati on of radiation detection instruments and equipment for assessment of the radiation areas;7.procurement, maintenance, and issuance of protective clothing and equipment;8.shipping, storage, and rece iving of all radioactive material to assure compliance with regulatory requirements;9.assistance in the decontamination of personnel, equipment, and facilities;10.preparation, maintenance, and issuance of the required regulatory, station, and personnel reports that are associated with radiation or radiation exposure; and11.preparation, maintenance, and implemen tation of the radiological respiratory protection program.12.Ensure stop work authority when requi red by actual or potential radiological conditions.

The chemistry department is res ponsible for measuring the radioactive content of all gaseous and liquid effluents from the site in accordance with the requirements of the Technical Specifications, Radiological Effluent Monitoring and Offs ite Dose Calculation Manual and 10 CFR 20.It is a policy of the Millstone Power Station to keep personnel radiation exposure within the applicable regulations , and beyond that, to keep it as low as reasonably achievable.12.5.2EQUIPMENT, INSTRUMENTATION, FACILITIES

The criteria for purchasing the various types of portable and laboratory equipment used in the radiation protection and chemistry department is based on several factors. Portable survey and laboratory radiation detection eq uipment is selected to provi de the appropriate detection capabilities, ranges, accuracy and durability required for the expected types and levels of radiation anticipated during normal operating or emergency conditi ons. Selection of respiratory protection equipment such as full-face masks, se lf-contained breathing apparatus, and respirator filters is made followi ng the guidance of applicab le approval regulations.

Radiation protection equipment, such as portab le survey meters, is maintained by radiation protection. Survey equipment for use in emergency situations is stored in emergency kits which are located in such areas as the control room and the emergency operation facility. Special portable equipment, such as pers onnel air samplers, is available fr om radiation protection, and is utilized at the discretion of ra diation protection supervision. Resp iratory protection equipment is primarily stored at the respirat ory storage and issue facilities.

MPS3 UFSAR12.5-3Rev. 30 Portable instruments for measuring radiation or radioactivity are used as required by 10 CFR 20, and by the provisions set forth in Regulatory Gu ide 1.97, Revision 2. Millstone 2 and 3 maintain a common inventory of hand-held radiation meters, electronic dosimeters, and National Voluntary Laboratory Accreditation Program (NVLAP) accredited individua l monitoring devices. The Millstone Station radiat ion protection group maintains adequate supplies of hand-held radiation meters, secondary dosimeters, and NVLAP accredi ted dosimeters for normal station activities, multiple unit shutdowns, and/or potential accident conditions.

The station will maintain an adequate supply of portable radi ation protection instrumentation strategically located at the facility to ensure the radiation protection staff is properly e quipped to perform their required functions. These instruments will be calibrated as specified by the manufacturers instructions and procedural requirements or as deemed necessary by radiation protection supervision. Calibration, operation, and maintenance procedures are followed for each specific type of instrument.

Detailed records of calibration a nd maintenance of each instrument are maintained at the station. Calibrations are performed using radiation sources of known activity. Thes e sources are calibrated or certified accurate by the National Institute of Standards and Technology (NIST). Calibration sources are stored by radiation pr otection. Actual calibr ation of equipment is performed in the calibration laboratories or ot her appropriate facilities.

The radiation protection group and chemistry group maintain appr opriate laboratory instruments to perform the required radiological evaluations to support the station needs. Radiation protection or chemistry personnel check each counting system at re gular intervals with standard radioactive sources to determine counting efficiencies, prope r voltage settings, and background count rates. Records are maintained for each instrument or counting system. Repair and maintenance of laboratory equipment is performed by station personnel or through vendor repair contracts.

The Millstone site contai ns the following areas:Unrestricted Area - access to which is ne ither limited nor controlled by the licensee.

Controlled Area - an area, outside a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason.Restricted Area - an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risk s from exposure to radiati on and radioactive materials.Radiologically Controlled Area/Radiological Control Area (RCA) - an area, posted with a sign by the licensee for the purpose of protecting indi viduals from exposure to radiation and/or radioactive materials. Dosimetry is always required within the RCA.

The Millstone Restricted Area generally corresponds to the area inside the protected area fence.

Millstone, with three units, has a number of RCAs within the Restricted Area. Within the RCA, Radiation Areas, High Radiation Areas, Technical Specification Locked High Radiation Areas, Very High Radiation Areas, Contaminated Areas, Airborne Radioactivity Areas, Radioactive Materials Areas, and Hot Particle Areas can be found.

MPS3 UFSAR12.5-4Rev. 30 A portal monitor and/or frisker and/or personnel contamination m onitor is located at selected control points to detect the spread of radioactive contamination to the areas outside the RCA. At the discretion of radiation protec tion supervision, a personnel moni tor or frisker is placed in specific areas at the station where contamination or the pote ntial for contamination may be present.Any areas where radioactive materi als and radiation may result in doses in excess of the dose limits in 10 CFR 20, Section 20.1301 is surveyed, cl assified, and conspicuously posted with the appropriate radiation caution si gns, labels, and signals in accordance with 10 CFR 20, Sections 20.1902 and 20.1903, except as described below.

The station employs admin istrative and physical security meas ures to prevent unauthorized entry of personnel into any high or very high radiation areas. The NRC gr anted the station approval in accordance with 10 CFR 20.1601(c), to use alternat ive methods for controlling access to high radiation areas in place of th e controls required by 10 CFR 20.1601 (a) and (b). These alternative methods are described in the administrative section of each unit's T echnical Specifications.Very High Radiation Areas are t hose areas where an individual could receive in excess of 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter (3 feet) from a radiation source or from any surface that the radiation penetrates. These areas, in addition to the controls specified in the Technical Specifications, have a unique key and a specific proc edure for entry into the area.Traffic patterns normally discourage or prevent access to radiation or poten tial radiation areas. Warning signs, audible and visible indicators, barricades, or locked doors are employed to protect personnel from access to high radiation areas that may exist temporarily or semi-permanently as a result of unit operations and maintenance.

Radiation protection services a nd facilities around the site provi de all workers the necessary protection and controls for work in radioactive environments.

Administrative radiation protec tion activities are cente red around the radiation protection office. Standard office equipment, equipment storag e areas, records storage, and some personnel dosimetry equipment are among the items to be included in the radiation protection office.

Personnel decontamination supplies and equipmen t are stored in the radiation protection decontamination facility. This room contains stainless steel showers and sinks, with drains directed to the wastewater treatment system (Section 9.2.3). A low-b ackground count laboratory is used for counting and/or identifying radi oactivity in airborne and liquid samples in conformance with 10 CFR 20, and to 10 CFR 50 App. A General Desi gn Criterion 64. The chemistry laboratory is used to perform chemical and elemental analyses of environmental effluents. All sink and floor drains in this room are directed to the wastewater treatment system; fume hood exhaust is directed to the ventilation system. Equi pment used to perform routine counting and analyses on all plan t radioactivity samples, as required by 10 CFR 20, are acquired, maintained and calibrated as appropriate.

All personnel entering contaminated areas are required to wear pr otective clothing. The nature of the work to be done, the contamination level in the area, and the total in dustrial risks, are the MPS3 UFSAR12.5-5Rev. 30 governing factors in the selecti on of protective clothing to be worn by individuals. Additional protective clothing stations are established at temporary dressing rooms or strategic locations, as required, to ensure efficient operations and to pr eclude the spread of c ontamination. Protective clothing available at the st ation includes the following:*modesty garments

• shoe covers*overshoe rubbers*head covers

• gloves*coveralls and lab coats Additional items of specialized apparel are available for operations involving high-level contamination, such as:*plastic or rubber suits

• surgeon's masks*face shields*bubble hoods All protective clothing is cleaned and decontaminated at a vendor laundry, on-site laundry facility, or disposed of as radioactive waste.Appropriate training and written guidance govern the proper use of protective clothing, where and how it is to be worn and removed, the decontam ination facilities for pe rsonnel and equipment, and the areas to be used.

Respiratory protective equipment is available to qualified station personnel and issued to individuals, as required by actual or potential occupational risk of the work assignment. The respiratory protection program follows the gui dance of Regulatory Guide 8.15, Revision 1, and complies with 10 CFR 20, Subpart H. Respirat ory protection equipment is stored at the respiratory storage and issue faciliti es. Respiratory equipment may include:*pressure demand full-face-pi ece air line respirators;*continuous air fl ow hoods or suits;*pressure demand full-face-piece self-contained breathing apparatus; and MPS3 UFSAR12.5-6Rev. 30*full-face mechanical filter respirators.

Respiratory protective equipment is cleaned, sanitized, repaired and decontaminated at respiratory storage and issue facilities , or at vendor facilities.All radiation workers are issued NVLAP accredit ed dosimeters and are required to wear such dosimeters at all times while within any RCA.

All other individuals who enter an RCA are required to wear an individua l radiation monitoring device.

Electronic dosimeters or direct-reading pocket dosimeters are issued as an additional method for determining gamma exposure. All individuals are required to examine their dosimeters at frequent intervals while in radiation areas. The use, care, and testing of these direct readi ng dosimeters will follow applicable guidan ce of Regulatory Guides 8.4, Revision 1, and 8.28, Revision 0.Special or additional dosimetry, such as finger ring dosimeters and teledosimetry, are issued under special conditions at the discretion of radiation protection supervision. The NVLAP accredited individual monitoring devices are processed periodically at the discretion of radiation protection personnel.

In addition, they can be pro cessed promptly whenever it appears that an overexposure may have occurred.

Dosimeter records furnish the exposure data for th e administrative control of radiation exposure.

Exposure records for each individual are mainta ined in accordance with the guidance of Regulatory Guide 8.7, Revision 2.12.5.3PROCEDURES All radiation protection proce dures and methods of operation for ensuring that occupational radiation exposure is as low as reasonably achievable (ALARA) follow the provisions and suggestions of Regulatory Guides 8.8, Revision 3; 8.10, Revision 1-R; and 1.33, Revision 2, as applicable. Such procedures are implemented by qualified pers onnel whose qualifications meet the requirements of Regulatory Guide 1.8, Revi sion 3. In addition, all administrative and procedural practices associated with the monitoring of occupational radiation exposure follow the guidance of Regulatory Guides 8.2, Revision 1; 8.4, Revision 1; 8.7, Revision 2; 8.9, Revision 1; and 8.34, Revision 0.

Many radiation protection procedures at Millstone Nuclear Power Station are common to Units 2 and 3. Radiation protection procedures are an inte gral part of the ALARA program at the station.

Access to restricted/radiologicall y controlled areas is controlle d by administrative and physical security measures as require d by 10 CFR 20, Subparts G and J.Station management assures entry control to hi gh radiation areas throug h the administration of radiation work permits (RWPs) that stipulate purpose of entry, work location, radiological conditions, surveillance and dosimetry requirements, stay time, protective clothing, respiratory MPS3 UFSAR12.5-7Rev. 30 protective equipment, special t ools, engineering controls, special personnel monitoring devices, and other procedur al requirements.The following are some objectives for issuing RWPs.1.Provide a detailed assessment of the actual and potential radiation hazards that are associated with the job function and area.2.Ensure that proper protective measures are taken to safely perform the required tasks in the area and to maintain the Total Effective Dose Equivalent as low as reasonably achievable (ALARA).3.Provide a mechanism for individuals to acknowledge their understanding of the radiological conditions, the protective and safety equipment and measures required, and the willingness to follow the requirements designated on the RWP.4.Provide a system for recording the sour ces (station systems and components), job types and functions, and personnel categories where exposures occur.RWPs are issued for general and specific activities performed in radiation areas, contaminated areas, airborne radioactivity areas , and for all activities that requi re entrance into high radiation areas, and very high radiation areas as defined in 10 CFR 20, Section 20.1003. RWPs are also issued prior to maintenance or inspection of contaminated or radioactive equipment with removable contamination in excess of 1,000 dpm/100 cm 2 beta-gamma and/or 100 dpm/100 cm 2 alpha. RWPs are also required prior to entrance into the reactor containment of any unit.

Under limited situations and at the discretion of radiation protection supervision, continuous radiation protection personnel coverage may be substituted for an RWP, such as an emergency which threatens personnel or plant safety.

Radiation protection personnel routinely survey selected areas of the station to assess and control exposure to radiation and radioactive materi als in accordance with 10 CFR 20, Section 20.1501.

Depending on the type of survey required and an ticipated types and levels of radioactivity, various portable instruments and techniques are us ed to perform these surveys. Results of all surveys are recorded and kept on file at the radiation protection office on a short term basis.If necessary, survey sheets may be posted. Permanen t storage is provided by forwarding records to the nuclear records facility. Re porting practices for all normal a nd accident conditions comply with the regulations set fo rth in 10 CFR 20, Subpart M.

Area surveys are performed at scheduled frequencie s, based on location, radiation levels, station status, and occupancy. All area su rvey readings are recorded and filed as required by 10 CFR 20, Section 20.2103 and Regulatory Guide 8.2, Revi sion 0. Caution placards, describing the radiological conditions, are posted to comply with the requirements of 10 CFR 20, Section 20.1902.

MPS3 UFSAR12.5-8Rev. 30Surveys for contamination are used to assess c ontainment of radioactive materials and the need for decontamination of an area.Contamination is measured at selected locations throughout the station, where the potential for the spread of contamination exists. C ontamination surveys are made using the "smear" or "swi pe" technique, or by using an a ppropriate portable instrument.

Scheduled frequencies are based on location, radiation levels, station status and occupancy, or as required by actual operating conditions, and as directed by radiation protection supervision.

Contamination surveys are performed on personnel, equipment, and in uncontrolled areas to ensure that radiological control methods are adequate. Personnel, equipment, and material leaving contaminated areas are monitored to prevent the sp read of contamination in to clean areas. Areas, equipment, and personnel that may be contaminated with radioactive material are decontaminated using applicable methods and techniques, such as those suggested in NCRP65 and IE Circular 81-07.Levels of contamination are also used to judge the potential for ai rborne radioactive material and the need for monitoring air, a nd the use of engineering contro ls or respiratory protection.It is management's intent to control airborne radioactivity levels as effectively as practicable by proper preventive measures, engineering cont rols, and good housekeeping techniques. In the event of a radioactive airborne problem, every effort is made to promptly assess the situation.

Section 12.3.4 provides information on the inst alled airborne radi oactivity monitoring instrumentation.

Control of airborne radioactivity levels is assured through the use of the station's heating, ventilation, and air-conditioning (HVAC) systems and portable air movers and filters. The HVAC systems provide controlled air movement and filtration capability for those areas with a high potential for airborne radioactiv ity problems. As required, special control techniques are used to minimize airborne exposure arising from special work projects. Respiratory protection equipment is available for use in those situations where ai rborne radioactivity haza rds exist and where other control measures are inadequate at the location and time. Respirat ory protection equipment use is assessed based upon the principle of keeping the Total Effective Dose Equivalent ALARA, consistent with minimizing total occupational risk.

The special control techniques used to minimiz e airborne exposure incl ude decontamination of the component or area prior to performing work, keeping work surfaces damp while work is in progress, and using tents or glove bags in c onjunction with appropriate , filtered ventilation systems.Techniques for obtaining breathing zone air samples are grab samples taken in areas representative of the worker's breat hing zone and/or lapel air samplers.

Some conditions which require special air samp ling include lifting the reactor vessel head, venting a contaminated system, and wo rking on an open contaminated system.

In regard to reporting practices for airborne contamination surveys, radiation protection supervision is notified when airborne concen trations read 30 percen t of DAC and the area requires posting if this condition persists for a sustained period.

MPS3 UFSAR12.5-9Rev. 30 All airborne contamination survey sheets are reviewed by radiation pr otection supervision and filed.The air sampling program provides information on the potential inhalation of radioactive material by workers. The information is used to determin e what remedial action or protective measures such as respirators, glove boxes, or engineering controls are necessary to protect the worker. Air samples are taken for all work on systems whic h have the potential for release of airborne radioactivity. Surveys are performed on a routine basis, dependi ng on location, station status, and occupancy. In addition, surveys are performe d whenever work is required on a known or potentially contaminated system that must be opened to the working environment or whenever welding, burning, or grinding is performed on a known or potentially co ntaminated system.

Surveys are also performed whenever the conti nuous air monitor indicate s an airborne problem and prior to containment entry. Additional surveys are performed as deemed necessary by radiation protection supervision.Prior to issuance and use of required respiratory protection equipment, each individual must have satisfactorily completed the following:*a satisfactory medical evaluation to ensure that the individual is medically fit to use respiratory protection devices;*training for the device to be used;*a fit test (face sealing devices only); and The air sampling and respirator y protection programs meet the recommendations and provisions of 10 CFR 20, Subpart H, Regulator y Guide 8.15, Revision 1, and NUREG-0041.

Special procedures control the ha ndling or movement of material within and from restricted and radiologically controlled areas, such as the shipme nt and receipt of radioactive materials. These procedures comply with the regulations stipulated in 49 CFR 170-178, 10 CFR 70, 10 CFR 71, and 10 CFR 20.1906.As previously discussed in th is section, all radiation workers receive a NVLAP accredited individual monitoring device and direct-reading pocket ion chambe r and/or electronic dosimetry to monitor personnel exposure. Exposure records are filed and retained for each individual in accordance with the recommendations of Regulat ory Guides 8.2, Revisi on 1 and 8.7, Revision 2, and as required by 10 CFR 20, Subpart L. Any reports of overexposur es and excessive levels and concentrations comply with the regulations of 10 CFR 20, Subpart M. Reports of personnel monitoring, and reports of theft or loss of licensed material are issued in accordance with the regulations required by 10 CFR 20, Subpart M.

The bioassay program at the Millstone Point Nuclear Power Station follows the guidance of Regulatory Guides 8.9, Revision 1 and meets the requiremen ts of 10 CFR 20, Section 20.1204.

The bioassay program includes:

MPS3 UFSAR12.5-10Rev. 30*determination of the conditions under which bioassays should be required;*selection of measurement techniques, measurement frequency, and program participants;*action points and actions to be take n based on measurement results; and*interpretation of measurement results in term s of location of radioactive material in the body, the quantity present, the rate of elimin ation, and the resulting dose commitment and*use of personnel contaminati on monitors, located at RCA ex its and the Protected Areas exits, which may serve as "passive monitors" for detection of intern al contamination in lieu of periodic whole body counts for all workers.

A whole-body counter is located at the station as needed for in vivo measurement of station personnel, visitors, or support personnel.

The whole-body counter provides preliminary background information, periodic evaluation, and emergency capabil ity for detecting internal exposure conditions. Assessment of internal radiation exposur e for station personnel may be performed, for example when:*individuals have a known or suspected inta ke of four or more DAC hours within a calendar week,*incidents involve contamination around the nose or mouth; and*accidents involve a potential intake. Excret a samples from suspected individuals may be sent to a qualified laboratory for analysis.Training in radiation protection principles and procedures is performed by the Nuclear Training Department or by qualifi ed station personnel. New employees , contractors, and other supporting personnel receive validation of prior training and orientation training, as appropriate, before the beginning of their work assignments.

All permanent station personnel who are required to work in the RCA are required to successfully complete basic training courses and practical exercises to dem onstrate their proficiency and competence.All radiological workers participate in the radiological worker training program. The radiological worker training program maintains the profic iency of employees th rough training and annual retraining on selected material.

Additional training is given for selected tasks which require increased radiation protection.

The content of the radiation protection related training program meets th e intent of Regulatory Guide 8.27, Revision 0; Regulatory Guide 8.13, Re vision 1; Regulatory Guide 8.29, Revision 0; and NUREG-0731. The program content is detailed in Section 13.2, Training Program. Details of the Emergency Plan which meet the inte nt of NUREG-0731, dated 1980, are given in Section 13.3, Emergency Planning.

MPS3 UFSAR12.5-11Rev. 30 Assessments are performed on al l radiation protection procedur es including those such as emergency procedures and instrument storage, calibration, and maintenance procedures, in addition to the procedures specifically required by Regulatory Guide 1.33, Revision 2.12.5.4REFERENCE FOR SECTION 12.512.5-1Nuclear Regulatory Co mmission, Code of Federal Regulations 10 CFR Part 20, "Standards for Protecti on Against Radiation"12.5-2Nuclear Regulatory Commission, Code of Federal Regulations 10 CFR Part 50, App. A, "General Design Criteria for Nuclear Power Plants," "Domestic Licensing of Production and Utilization Facilities"12.5-3Nuclear Regulatory Co mmission, Code of Federal Regulations 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material"12.5-4Nuclear Regulatory Co mmission, Code of Federal Regulations 10 CFR Part 71, "Packaging and Transportation of Radioactive Material"12.5-5Department of Transportation, Code of Federal Regulat ions 49 CFR 170-178, "Subchapter C-Hazardous Materials Regulations"12.5-6NUREG-0731, "Guidelines for Utility Management Structure and Technical Resources," 198012.5-7NUREG-0800, USNRC. "Standard Review Plan", Revision 1.12.5-8Regulatory Guide 1.8, Rev. 3 "Qualification and Traini ng of Personnel for Nuclear Power Plants" 12.5-9Regulatory Guide 1.33, Rev. 2, "Qua lity Assurance Program Requirements"12.5-10Regulatory Guide 8.2, Rev. 1, "Guide for Administrative Prac tices in Radiation Monitoring"12.5-11Regulatory Guide 8.4, Rev.

1, "Personnel Monitoring Device - Direct-Reading Pocket Dosimeters"12.5-12Regulatory Guide 8.7, Rev.

2, "Instructions for Recordi ng and Reporting Occupational Radiation Exposure Data"12.5-13Regulatory Guide 8.8, Rev.

3, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Sta tions will be as Low as is Reasonable Achievable" MPS3 UFSAR12.5-12Rev. 3012.5-14Regulatory Guide 8.9, Rev. 1, "Acceptable Concepts, Models, Equations and Assumptions for a Bioassay Program"12.5-15Regulatory Guide 8.10, Rev. 1-R, "Opera ting Philosophy for Main taining Occupational Radiation Exposures As Low As is Reasonably Achievable"12.5-16Regulatory Guide 8.13, Rev. 3. "Instructi on Concerning Prenatal Radiation Exposure"12.5-17Regulatory Guide 8.15, Rev. 1, "Accepta ble Programs for Respiratory Protection"12.5-18Regulatory Guide 8.27. "Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants", Revision 0.12.5-19Regulatory Guide 8.28, Rev.

0, "Audible-Alarm Dosimeters"12.5-20Regulatory Guide 8.29. "Ins tructions Concerning Risks fr om Occupational Radiation Exposure", Revision 1.12.5-21Regulatory Guide 8.34, Rev. 0, "Monitoring Criteria and Methods To Calculate Occupational Radiation Doses"12.5-22Regulatory Guide 1.33. "Quality Assu rance Program Require ments", Revision 2.12.5-23INPO 05-008, "Guidelines for Radiological Protection at Nuclear Power Stations" 12.5-24IE Circular 81-007, "Control of Ra dioactively Contam inated Material"12.5-25NCRP Number 65, "Management of Pe rsons Accidently Contaminated with Radionuclides"12.5-26NUREG-0041, Rev. 1, "Manual of Respiratory Protection Against Airborne Radioactive Material" MPS-3 FSARPage 1 of 1Rev. 30TABLE 12.5-1DELETED BY FSARCR 04-MP3-040 MPS-3 FSARPage 1 of 1Rev. 30TABLE 12.5-2DELETED BY FSARCR 04-MP3-040 MPS-3 FSAR FIGURE 12.2 - 1 ARRA NGEMENT - OPERATING PERSONNEL ACCESS AND EGRESS (PLA N ELEVATION 3 FEET 8 INCHES AND ABOVE)Amendment 10 October 1984 Rev. 20.3 MPS-3 FSAR FIGURE 12.2-2 ARRANGEMEN T - OPERATING PERSONNEL ACCESS AND EGRESS (PLAN ELEVATION 24 FEET 6 INCH ES AND ABOVE)

August 1999 Rev. 20.3 MPS-3 FSAR December 1997Rev. 20.3FIGURE 12.2-3 ARRANGEMENT - OPERATING PERSONNEL ACCESS AND EGRE SS (PLAN ELEVATION 38 FE ET 6 INCHES AND ABOVE)

MPS-3 FSARFIGURE 12.2-4 I-131 CONCENTRATION CONTAINMENT (HISTORICAL) . Rev. 21.3 MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-1 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-2 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-3 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-4 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

MPS-3 FSARFIGURE 12.3-5 CONTAINMENT MONITORING SYSTEM THIS FIGURE NOW IN SECTION 11.5 (FIGURE 11.5-2)

Amendment 12 Februrary 1985 Rev. 20.3 MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-6 DESIGN BASIS RADIATION ZONES FOR SH IELDING (SHUTDOWN/REFUELING)

MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-7 DESIGN BASIS RADIATION ZONES FOR SH IELDING (SHUTDOWN/REFUELING)

MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-8 DESIGN BASIS RADIATION ZONES FOR SH IELDING (SHUTDOWN/REFUELING)

MPS-3 FSAR December 1997Rev. 20.3 FIGURE 12.3-9 DESIGN BASIS RADIATION ZONES FOR SH IELDING (SHUTDOWN/REFUELING)

MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 1ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 1 TO AUXILIARY BUILDING MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 2ROUTES TO POST-ACCIDENT VITAL AREA S - ROUTE 2 TO THE ESF BUILDING MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 3ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 3 TO THE HYDROGEN RECOMBINER BUILDING MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 4ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 4 TO THE HYDROGEN RE COMBINER BUILDING AND TO UNIT 2 CHEMISTRY L ABS MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 5ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 5 TO FUEL BUILDING MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 6ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 6 TO TURBINE BUILDING MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 7ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 7 TO DIESEL GENERATOR MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 8ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 8 TO ESF BUILDING MPS-3 FSARRev. 21.3FIGURE12.3-10SHEET 9ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 9 TO AUXILIARY BUILDING MPS-3 FSARFIGURE 12.3 - 11 FUEL TRANSFER TUBE SHIELDINGAmendment 6 January 1984 Rev. 20.3 MPS-3 FSARFIGURE 12.3 - 12 UPPER REACTOR CAVITY NEUTRON SHIELDAmendment 8 May 1984 Rev. 20.3