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 FSAR Millstone Power Station Unit 3 Safety Analysis Report Chapter 12

Table of Contents tion Title Page INTRODUCTION .................................................................................... 12.0-1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) ....................................................................... 12.1-1 RADIATION SOURCES ......................................................................... 12.2-1

.1 CONTAINED SOURCES ........................................................................ 12.2-1

.1.1 Sources for Design Basis Loss-of-Coolant Accident ............................... 12.2-2

.2 AIRBORNE RADIOACTIVE SOURCES............................................... 12.2-2

.3 REFERENCES FOR SECTION 12.2....................................................... 12.2-4 RADIATION PROTECTION DESIGN FEATURES ............................. 12.3-1

.1 SHIELDING ............................................................................................. 12.3-1

.1.1 Primary Shielding ..................................................................................... 12.3-3

.1.2 Secondary Shielding ................................................................................. 12.3-3

.1.3 Accident Shielding.................................................................................... 12.3-5

.1.3.1 Containment and Control Room Design................................................... 12.3-5

.1.3.2 Post-Accident Access to Vital Areas ........................................................ 12.3-6

.2 FACILITY DESIGN FEATURES ......................................................... 12.3-12

.2.1 Location and Design of Equipment to Minimize Service Time ........................................................................................................ 12.3-12

.2.2 Location of Instruments Requiring In Situ Calibration .......................... 12.3-13

.2.3 Location of Equipment Requiring Servicing in Lowest Practicable Radiation Field (or Movable to Lowest Practicable Radiation Field).................................................................... 12.3-13

.2.4 Valve Location and Selection ................................................................. 12.3-14

.2.5 Penetrations of Shielding and Containment Walls by Ducts and Other Openings ................................................................................ 12.3-14

.2.6 Radiation Sources and Occupied Areas.................................................. 12.3-14

.2.7 Minimizing Spread of Contamination and Facilitation of Decontamination Following Spills ......................................................... 12.3-15

.2.8 Piping to Minimize Buildup of Contamination ...................................... 12.3-15

.2.9 Flushing or Remote Chemical Cleaning of Contaminated Systems ................................................................................................... 12.3-15

.2.10 Ventilation Design .................................................................................. 12.3-16

.2.11 Radiation and Airborne Contamination Monitoring............................... 12.3-16

.2.12 Temporary Shielding .............................................................................. 12.3-16 12-i Rev. 30

tion Title Page

.2.13 Solid Waste Shielding............................................................................. 12.3-16

.2.14 Remote Handling Equipment.................................................................. 12.3-16

.2.15 Maximum Expected Failures of Fuel Element Cladding and Steam Generator ..................................................................................... 12.3-16

.2.16 Sampling Stations ................................................................................... 12.3-17

.2.17 Cobalt Impurity Specifications ............................................................... 12.3-17 3.2.18 Reactor Cavity Filtration System............................................................ 12.3-17

.3 VENTILATION...................................................................................... 12.3-17

.3.1 Design Objectives ................................................................................... 12.3-17

.3.2 Design Description ................................................................................. 12.3-18

.3.3 Personnel Protection Features................................................................. 12.3-19

.3.4 Radiological Evaluation.......................................................................... 12.3-20

.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING....................................................................................... 12.3-20

.4.1 Purpose.................................................................................................... 12.3-20

.4.2 System Design ........................................................................................ 12.3-20

.4.3 Class 1E Area Monitors .......................................................................... 12.3-21

.4.4 Non-Class 1E Area Monitors.................................................................. 12.3-22 3.4.5 Airborne Radioactivity Monitoring ........................................................ 12.3-22

.5 REFERENCES FOR SECTION 12.3..................................................... 12.3-22 DOSE ASSESSMENT ............................................................................. 12.4-1 HEALTH PHYSICS PROGRAM ............................................................ 12.5-1

.1 ORGANIZATION .................................................................................... 12.5-1

.2 EQUIPMENT, INSTRUMENTATION, FACILITIES............................ 12.5-2

.3 PROCEDURES ........................................................................................ 12.5-6

.4 REFERENCE FOR SECTION 12.5 ....................................................... 12.5-11 12-ii Rev. 30

List of Tables mber Title

-1 Parameters Used in Calculation of Design Radiation Source Inventories (HISTORICAL)

-2 Radioactive Sources in Containment Building (HISTORICAL)

-3 Radioactive Sources in the Auxiliary Building (HISTORICAL)

-4 Radioactive Sources in the Waste Disposal Building (HISTORICAL)

-5 Radioactive Sources in the Fuel Building (HISTORICAL)

-5A Other Radioactive Sources (HISTORICAL)

-6 Inventory of an Average Fuel Assembly after 650 Days of Operation at 3,636 MWt at Shutdown and 100 Hours after Shutdown (Ci) (HISTORICAL)

-7 Source Intensity in the Most Radioactive Fuel Assembly

• after 650 Days of Operation at 3636 Mwt (HISTORICAL) 2-8 Radionuclide Concentrations in the Spent Fuel Pool from Refueling 100 Hours after Shutdown* (HISTORICAL)

-9 Radiation Sources

• Reactor Coolant Nitrogen-16 Activity (HISTORICAL)

-10 Assumptions Used in the Calculation of Airborne Concentrations (HISTORICAL)

-11 Airborne Concentrations Inside Major Buildings (mci/cc) (HISTORICAL)

-1 Radiation Zones

-2 Radiation Monitoring System - Area Radiation Detector Location

-3 Operator Activity Locations and Time Durations

-4 Activity Initiation Time

-1 Deleted by FSARCR 04-MP3-040

-2 Deleted by FSARCR 04-MP3-040 12-iii Rev. 30

List of Figures mber Title

-1 Arrangement - Operating Personnel Access and Egress

-2 Arrangement - Operating Personnel Access and Egress

-3 Arrangement - Operating Personnel Access and Egress

-4 I131 Concentration Containment

-1 Design Basis Radiation Zones for Shielding (Normal Operations)

-2 Design Basis Radiation Zones for Shielding (Normal Operations)

-3 Design Basis Radiation Zones for Shielding (Normal Operations)

-4 Design Basis Radiation Zones for Shielding (Normal Operations)

-5 This figure moved to Section 11.5 (Figure 11.5-2)

-6 Design Basis Radiation Zones for Shielding (Shutdown/Refueling)

-7 Design Basis Radiation Zones for Shielding (Shutdown/Refueling)

-8 Design Basis Radiation Zones for Shielding (Shutdown/Refueling)

-9 Design Basis Radiation Zones for Shielding (Shutdown/Refueling)

-10 Routes to Post-Accident Vital Areas

-11 Fuel Transfer Tube Shielding

-12 Upper Reactor Cavity Neutron Shield

-1 Figure has been deleted 12-iv Rev. 30

INTRODUCTION r to the licensing and operation of a nuclear power reactor, the applicant must include, in pter 12 of the FSAR, an estimate of the radiation dose expected to be received by station onnel. This includes an estimate for both whole body dose from direct radiation and internal e from airborne activity. This is provided to ensure the proposed station design related to upational radiation exposure control (e.g., shielding and airborne activity control) will be icient to ensure compliance with 10 CFR 20. The assessments presented in Chapter 12 are ed on nominal assumptions and generic models and criteria that were appropriate at the time original FSAR was written. They represent estimates chosen for the purpose of projecting upational dose consequences. They do not represent design or operational requirements. It was y expected that actual operational data would not match the chosen assumptions and criteria ented in Chapter 12, but in general, the estimates were expected to be conservative.

e the plant is operational, compliance with occupational exposure limits and controls is ured and controlled by compliance with the Technical Specifications and 10 CFR 20. These uments require a Radiation Protection Program and an ALARA Program. These are dynamic grams that change to meet changing regulations, industry initiatives and state-of-the-art tices. These programs provide detailed controls on occupational exposure. The Radiation tection Program ensures that the requirements of 10 CFR 20 are met. The ALARA Program ures that controls are imposed and assessments are performed to reduce occupational exposure evels that meet current standards and are not the original estimates of Chapter 12. These grams are routinely audited by licensee and NRC staff for compliance and effectiveness.

ual occupational exposure reports are provided to the NRC to provide a real time measure of effectiveness of occupational exposure controls.

refore, compliance with occupational exposure regulations is controlled by the Radiation tection and ALARA programs, which are described in Chapter 12 of the FSAR. Current gram measures are contained in Chapter 12, however, the design parameters or quantities vided here are not updated from original values. The original bases for station design and ation control are relegated to historical perspective, as these bases were never intended to cribe conditions of operation. More accurate information on radiological quantities or ditions should be determined by referencing current radiation protection data available in the ation protection department.

12.0-1 Rev. 30

Millstone policy to implement a program that meets the intent of 10 CFR 20 and ensure that upational radiation exposures at its nuclear facilities are kept as low as reasonably ievable (ALARA).

ALARA Program criteria shall be in accordance with 10 CFR 20, Regulatory Guide 8.8, Rev.

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

program shall ensure that:

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

Annual collective doses, (person rem) are ALARA.

Individual doses within work groups are balanced to be consistent with:

a. experience

b. manpower availability

c. existing agreements Annual and three year goals are developed for collective doses at each unit.

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

An ALARA job review process exists for jobs with the potential for significant exposure.

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

Personnel are aware of ALARA program philosophy and trained in ALARA concepts.

Millstone management provides the necessary policy, resources and commitment for ALARA program.

An ALARA feedback system exists for workers to identify ALARA concerns or suggest ALARA improvements.

Corrective actions are considered when the attainment of specific ALARA goals are jeopardized.

12.1-1 Rev. 30

.1 CONTAINED SOURCES radioactivity values provided in this section are the design basis values used for the design of t shielding. As such they are considered historical and not subject to future updating. This rmation is retained to avoid loss of original licensing bases. As discussed in Chapter 12.0, pliance with occupational exposure limits and controls is ensured and controlled by pliance with the Technical Specifications and 10 CFR 20 which was implemented at MPS-3 the Radiation Protection Program and the ALARA Program.

source of radioactivity contained in the streams of the various radioactive waste management ems are the nuclides generated in the reactor core and activation of nuclides in the reactor lant system and the air surrounding the reactor vessel. These sources are described in Chapter Table 12.2-1 presents the principal parameters which are used to establish design radiation rce inventories. The design basis for the shielding source terms for fission products in this ion is cladding defects in fuel rods producing 1 percent of the core thermal power. The design s for activation and corrosion product activities are derived from measurements at operating ts and are independent of fuel defect level. The radionuclide activity levels in the reactor lant at the design basis level are given in Section 11.1. The models and assumptions used in rmining these sources are also given in Section 11.1.

reactor core source description is similar in that given in Topical Report RP-8A (Stone &

bster Engineering Corporation), Section 4.1.1, with appropriate adjustment to account for er level difference.

activity of a spent fuel assembly is calculated using appropriate fission yields, decay stants, and thermal neutron cross sections. Isotopic inventories are based on full power ration for 650 days. The inventory of an average fuel assembly at shutdown and 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> r shutdown is given in Table 12.2-6. The source strengths in MeV/sec for the most radioactive assembly at several decay times (assuming a radial peaking factor of 1.65) is given in le 12.2-7. The isotopic activities (expected and design) in the fuel pool water are given in le 12.2-8 based on expected and design primary coolant activities homogeneously mixed with eling 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 tdown.

location and geometry of significant sources of radiation in the containment building, iliary building, fuel building, waste disposal building, condensate polishing building, ESF ding, and the tank yard are presented in Tables 12.2-2 through 12.2-5A and on Figures 12.2-1 ugh 12.2-3. The method used to arrive at the source terms presented in Tables 12.2-2 through

-5A considers the operating parameters described in Table 11.1-3. The evaporator bottoms vity is based on liquid particulate concentrations at the start of plant operations, thereby imizing the decay time to produce a batch of concentrate. Normal operating flow rates are also d to develop demineralizer nuclide concentrations consistent with the decontamination factors d to determine radioactive concentrations of waste liquid streams. The inventory of radioactive lides on filters downstream of the demineralizer assumes a fraction of the resin fines are 12.2-1 Rev. 30

iation source terms are also presented in these tables in terms of seven discrete energy levels V) used in plant shielding design.

eighth energy level corresponds to the N-16 activity in the reactor coolant for various ponents, listed in Table 12.2-2. N-16 activity is the controlling source in the design of the ondary shield and is tabulated in Table 12.2-9, in Ci per gram of coolant, as a function of sport time through the reactor coolant loop.

method used to calculate the source strength of N-16 in each component is described in detail he Stone & Webster Topical Report RP-8A (Stone & Webster Engineering Corporation). The uired parameters for this calculation are transit time to the component, transit time through the ponent. N-16 is not a factor in the radiation sources for the systems and components located ide the containment due to its short (7.11 seconds) half life and the greater than one minute sport time before the letdown flow exits the containment.

2.1.1 Sources for Design Basis Loss-of-Coolant Accident radiation sources of importance for the design basis accident are the sources within the tainment and sources transported via the emergency safeguards features (ESF) cooling em.

fission product radiation sources considered to be released from the fuel to the containment owing a maximum credible accident are based on the assumptions given in Regulatory Guide and NUREG-0737. This source term was used to evaluate the original plant shielding design, is not used in Chapter 15 analyses.

sources in the ESF system are based on the nongaseous activity, i.e., 50 percent halogens and rcent remainder, being retained in the coolant water. Noble gases formed by the decay of gens in the sump water are assumed to be released to the containment and not retained in the er. Credit has been taken for dilution by the reactor coolant system volume plus the contents of refueling water storage tank and other ESF system component volumes.

opic fission product sources in the Fuel are given in Section 11.1.

2.2 AIRBORNE RADIOACTIVE SOURCES principal sources of airborne radionuclides are the reactor coolant system and the air ounding the reactor vessel. Reactor coolant leaking into plant buildings results in the release irborne contamination. The radioactivity sources which contribute to the radioactive airborne ases from the plant waste management system and the plant ventilation system are described hapter 11.

centrations of airborne activity for the expected and design conditions in the containment cture, turbine building, and fuel building are listed in Table 12.2-11. The bases used to derive 12.2-2 Rev. 30

sidered negligible.

borne levels in general access areas of the auxiliary, turbine, and fuel buildings are expected to ower than equipment cubicles of these buildings, since ventilation flow paths are normally cted from areas with less potential contamination to areas with greater potential for tamination.

tainment Structure containment structure is not normally occupied during power operation. Radiation protection cedures control access. Two recirculating charcoal filters can be operated to ensure that orne iodine in the containment is as low as is reasonably achievable for work in that area ction 9.4.7).

ure 12.2-4 presents expected iodine-131 concentrations in the containment structure after the charcoal filters have been placed in operation.

er shutdown, the containment purge air system can be used to reduce the airborne activity hin the containment structure. The filtered purge is rated at 30,000 cfm.

ioactivity associated with primary coolant leakage is mixed in the containment atmosphere by containment atmosphere recirculation system (Section 9.4.7.1). Removal occurs through oactive decay. At appropriate times, the radioactive inventory may be reduced by means of rculating the containment air through the charcoal filters or containment purging.

containment atmosphere tritium assumes the same relative concentration (Ci of tritium per m of water) as exists in the reactor coolant leakage.

bine Building ioactivity associated with steam leakage is assumed to be uniformly distributed by the turbine ding ventilation system (Section 9.4.4).

oval occurs through decay and ventilation exhaust. The tritium concentration in the turbine ding is calculated assuming that all the steam leakage into the turbine building remains eous.

iliary Building borne radioactivity associated with primary coolant leakage is assumed to be limited to cess equipment cubicles and is removed by auxiliary building ventilation system ction 9.4.3). Removal occurs through decay and ventilation exhaust. The ventilation system is figured to preclude mixing of the atmosphere in process equipment cubicles with the osphere in the general access areas as described above.

12.2-3 Rev. 30

age is collected in sumps and drains and is not generally available for evaporation.

l Building fuel building ventilation system is described in Section 9.4.2.

tritium concentrations in the fuel building atmosphere assumes that the atmosphere above nt fuel pool has the same relative tritium concentration, Ci of tritium per gram of water, as the water. Airborne concentrations are presented in Table 12.2-11.

.3 REFERENCES FOR SECTION 12.2

-1 NUREG-0017 United States Nuclear Regulatory Commission. Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE CODE). 1976 Office of Standards Development.

-2 Regulatory Guide 1.4.

-3 Stone & Webster Engineering Corporation (SWEC) 1975. Radiation Shielding Design and Analysis Approach for Light Water Reactor Power Plants RP 8A. Topical Report.

Cambridge, Mass.

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

12.2-4 Rev. 30

Parameter ower level (MWt) 3,636 ailed Fuel Fraction 0.01 rimary-to-Secondary Leak Rate (lb/day) 1,370 eactor Operating Time (days) 650 scape Rate Coefficients (sec-1):

1. Noble Gases 6.5 x 10-8

2. Br, Rb, I, and Cs nuclides 1.3 x 10-8

3. Te nuclides 1.0 x 10-9

4. Mo nuclides 2.0 x 10-9

5. Sr and Ba nuclides 1.0 x 10-11

6. Y, La, Ce, Pr nuclides 1.6 x 10-12 urification Letdown Flow Rate (gpm) 75 egasification Charcoal Delay Bed Holdup Time (days):

1. Kr 6.1

2. Xe 142.1 torical, not subject to future updating. This table has been retained to preserve original design s.

Page 1 of 1 Rev. 30

TABLE 12.2-2 RADIOACTIVE SOURCES IN CONTAINMENT BUILDING (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)

Outside Source Height Diameter Volum No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 6.10 (cm) (cm) (cc) 1 Reactor Core 1.05+12 (1) 4.63+12 4.68+12 3.83+12 7.98+11 1.27+12 3.29+12 2.23+07 1303 439 1.97+08 2 Regenerative Heat Exchanger Shell 1.73+05 2.91+05 1.26+05 1.22+05 9.74+04 1.36+05 3.03+04 2.01+07 427.0 30.48 3.12+05 Tube 7.23+03 6.11+04 2.67+04 3.99+04 1.92+03 1.20+04 9.42+03 -

3 Steam Generator Shell 1.44+05 2.43+05 1.05+05 1.02+05 8.14+04 1.14+05 2.53+04 1.94+07 335 2.11+08 Tube 2.97+01 1.38+02 3.19+01 1.62+01 1.79+00 9.84+01 1.21+00 - 1970 427 4 Pressurizer Liquid 2.83+04 1.62+05 8.61+04 6.30+04 1.13+04 1.01+04 1.85+04 4.93+05 1499 213 3.06+07 Gas 8.79+04 6.22+04 4.92+03 3.52+04 9.16+04 1.30+05 4.20+03 - 2.04+07 5 Containment Drains 1.98+05 3.33+05 1.43+05 1.40+05 1.11+05 1.55+05 3.46+04 - 259 145.0 3.75+06 Transfer Tank 6 Reactor Coolant 1.34+05 2.26+05 9.74+04 9.49+04 7.57+04 1.06+05 2.35+04 (See - - -

Piping Table 12.2-9) 9 Excess Letdown Heat 1.81+05 3.05+05 1.31+05 1.28+05 1.02+05 1.43+05 3.17+04 1.32+07 366 30.48 2.67+05 Exchanger 10 Pressurizer Relief 3.87+04 2.74+04 2.16+03 1.55+04 4.03+04 5.72+04 1.85+03 - 826 290 5.10+07 Tank Page 1 of 2 Rev

Activity (MeV/cc-Sec) for Energy (MeV)

Outside Source Height Diameter Volum No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 6.10 (cm) (cm) (cc) 11 Pressurizer Relief 3.87+04 2.74+04 2.16+03 1.55+04 4.03+04 5.72+04 1.85+03 - 29.2 38.1 3.02+04 Tank Transfer Pump 12 Containment Drains 1.98+05 3.33+05 1.43+05 1.40+05 1.11+05 1.55+05 3.46+04 - 29.2 38.1 3.02+04 Transfer Pump 13 Reactor Coolant Pump 1.51+05 2.54+05 1.09+05 1.07+05 8.52+04 1.19+05 2.64+04 1.25+07 427.0 122.0 5.0+06 (1) 1.05+12 = 1.05 x 1012 Historical, not subject to future updating. This table has been retained to preserve original design basis.

Page 2 of 2 Rev

TABLE 12.2-3 RADIOACTIVE SOURCES IN THE AUXILIARY BUILDING (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)

Source Height Outside Volume No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 (cm) Diameter (cm) (cc) 14 Cesium Removal Ion 1.87+07 (1) 5.02+08 2.67+07 1.48+06 3.34+05 1.46+05 6.02+04 206.4 106.7 1.71+06 Exchanger 15 Boron Evaporator Reboiler 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 441.9 53.3 9.21+05 16 Boron Evaporator 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 983.0 229 2.14+07 17 Boron Evaporator Reboiler 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 104.1 43.2 5.29+04 Pump 18 Boron Evaporator Bottoms 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 35.6 18.4 3.82+03 Pump 19 Boron Recovery Filters 9.00+06 2.45+08 6.06+07 9.57+05 1.60+05 6.97+04 2.88+04 83.3 17.7 2.08+04 20 Boron Evaporator Bottoms 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 78.4 17.7 1.95+04 Filter 21 Reactor Coolant Filter 1.72+08 4.11+08 6.33+07 1.32+07 3.08+06 2.88+05 1.74+05 48.2 16.8 1.07+04 22 Sealwater Injection Filter 4.02+05 4.72+07 4.68+08 2.44+06 -- -- -- 55.8 7.0 2.14+03 23 Sealwater Heat Exchanger 1.09+04 5.96+04 2.62+04 4.45+04 1.48+03 1.28+04 1.14+04 419.1 35.5 4.17+05 24 Thermal Regeneration 5.05+06 4.28+06 2.73+05 1.37+05 3.94+04 3.87+01 1.30+02 253.6 122.0 2.10+06 Demineralizer 25 Cation Bed Demineralizer 2.38+05 4.34+08 1.39+07 0.00+00 0.00+00 0.00+00 0.00+00 274.9 66.0 9.42+05 26 Mixed Bed Demineralizer 2.25+08 5.37+08 8.27+07 1.73+07 4.02+06 3.76+05 2.27+05 270.8 81.3 1.40+06 27 Letdown Heat Exchanger 1.67+05 2.75+05 1.27+05 1.32+05 1.05+05 1.43+05 4.14+04 541.0 53.3 1.21+06 28 Letdown Reheat Heat 1.67+05 2.75+05 1.27+05 1.32+05 1.05+05 1.44+05 4.15+04 221 22.9 9.10+04 Exchanger Page 1 of 3 Rev

Activity (MeV/cc-Sec) for Energy (MeV)

Source Height Outside Volume No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 (cm) Diameter (cm) (cc) 29 Letdown Chiller Heat 1.37+05 1.24+05 3.07+04 8.16+04 9.96+04 1.51+05 1.59+04 518.2 53.3 1.16+06 Exchanger 30 Moderating Heat Exchanger 1.37+05 1.24+05 3.07+04 8.16+04 9.96+04 1.51+05 1.59+04 556.3 45.7 9.12+05 31 Volume Control Tank Liquid 8.39+03 8.87+04 3.24+04 4.30+04 3.09+03 1.30+04 1.25+04 125.5 228.6 4.53+06 Volume Control Tank Gases 6.67+04 4.24+04 3.00+03 2.16+04 6.55+04 9.67+04 1.29+03 188.3 228.6 6.80+06 32 Degasifier Recirculation 1.13+04 9.46+04 2.64+04 3.87+04 1.40+03 1.16+04 8.19+03 35.1 27.0 1.57+04 Pump 33 Sealwater Return Filter 7.51+04 8.82+06 8.74+07 4.56+05 -- -- -- 48.3 16.8 1.07+04 34 Letdown Filter 7.56+05 8.88+07 8.80+08 4.59+06 -- -- -- 48.3 16.8 1.07+04 35 Boric Acid Tanks 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 579 488 1.08+08 36 Boric Acid Filters 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 48.3 16.8 1.07+04 37 Boric Acid Transfer Pumps 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 34.3 28.6 2.19+04 38 Primary Drains Transfer 1.98+05 3.33+05 1.48+05 1.40+05 1.11+05 1.55+05 3.46+04 356.9 152.4 6.51+06 Tanks 39 Primary Drains Transfer 1.98+05 3.33+05 1.48+05 1.40+05 1.11+05 1.55+05 3.46+04 29.8 30.5 2.17+04 Pump 40 Process Gas Charcoal Bed 2.56+07 2.67+05 1.46+05 1.90+05 4.73+05 7.48+05 7.91+04 492.7 182.8 1.29+07 Adsorbers 41 Degasifier Recovery Heat 1.55+05 1.45+05 3.32+04 8.30+04 1.02+05 1.58+05 1.28+04 709.0 38.1 7.11+05 Exchanger 42 Degasifier Trim Cooler 1.55+05 1.45+05 3.32+04 8.30+04 1.02+05 1.58+05 1.28+04 652.8 27.3 3.86+05 43 Degasifier Liquid 1.13+04 9.46+04 2.64+04 3.87+04 1.40+03 1.16+04 8.19+03 538.8 167.6 9.78+06 Page 2 of 3 Rev

Activity (MeV/cc-Sec) for Energy (MeV)

Source Height Outside Volume No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 (cm) Diameter (cm) (cc)

Degasifier Vapor 1.51+03 7.78+02 4.89+01 3.91+02 1.05+03 1.50+03 3.68+01 152.4 167.6 1.12+06 44 Degasifier Feed Preheater 1.51+05 1.40+05 3.22+04 8.05+04 9.88+04 1.53+05 1.24+04 373.4 30.48 2.72+05 45 Fuel Pool Post Filter 2.01+05 3.77+05 4.31+04 5.77+03 2.50+03 1.18+03 3.03-02 114 45.7 1.87+05 46 Fuel Pool Demineralizer 2.01+07 3.75+07 4.27+06 5.76+05 2.49+05 1.18+05 3.03+00 259.1 60.9 7.34+05 47 Process Gas Receiver 0.0+00 8.55+02 0.0+00 0.0+00 0.0+00 0.0+00 0.0+00 183 61.0 5.34+05 48 Charging Pumps 4.76+05 1.44+06 1.48+05 7.03+04 1.18+04 1.32+04 1.25+04 145 91.4 9.50+05 49 Boron Evaporator Bottoms 4.68+05 1.35+06 1.16+05 2.73+04 8.74+03 1.70+02 2.09+01 336.5 21.9 1.26+05 Cooler 50 Boron Evaporator Distillate 4.89-01 1.35+00 3.43-01 1.78-01 3.40-02 3.75-02 3.40-03 340.3 21.9 1.28+05 Cooler 53 Degasifier Condenser 1.23+06 6.32+05 3.97+04 3.17+05 8.54+05 1.21+06 2.99+04 152.4 76.2 6.95+04 54 Effluent Filters 7.58-01 6.79+00 3.02+00 1.79+00 3.74-01 9.46-04 1.56-04 147.3 32.4 1.21+05 (1) 1.87+07 = 1.87x107 Historical, not subject to future updating. This table has been retained to preserve original design basis.

Page 3 of 3 Rev

TABLE 12.2-4 RADIOACTIVE SOURCES IN THE WASTE DISPOSAL BUILDING (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)

Source Outside No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 Height (cm) Diameter (cm) Volume (c 51 Boron Demineralizer 1.31+03 (1) 3.26+04 9.59+02 5.12+01 1.38+01 9.80-01 2.31-02 206.4 106.7 1.7+06 52 Boron Demineralizer Filter 5.97+02 1.48+04 4.36+02 2.33+01 6.27+00 4.46-01 1.05-02 87.3 17.8 2.17+04 55 Waste Evaporator 3.80+04 1.31+05 1.38+04 1.80+03 5.57+02 2.18+02 6.70-03 973.5 182.9 2.10+07 56 Waste Evaporator Reboiler 3.80+04 1.31+05 1.38+04 1.80+03 5.57+02 2.18+02 6.70-03 621.3 55.9 1.52+06 57 Waste Evaporator Reboiler 3.80+04 1.31+05 1.38+04 1.80+03 5.57+02 2.18+02 6.70-03 109.2 43.2 6.43+04 Pump 58 Boron Evaporator Feed 1.61+02 1.34+04 6.38+02 7.98+00 1.37+00 1.69+00 1.98+00 43.2 18.4 5.06+03 Pumps 59 Waste Evaporator Feed 1.61+03 5.51+03 6.70+02 1.46+02 3.59+01 1.21+01 2.48+00 35.6 18.4 3.82+03 Pumps 60 Waste Evaporator Bottoms 3.80+04 1.31+05 1.38+04 1.80+03 5.57+02 2.18+02 6.70-03 35.6 18.4 3.82+03 Pump 61 Spent Resin Hold Tank 2.34+07 8.36+07 9.07+06 1.75+06 3.88+05 6.64+04 4.39+04 360.0 182.9 8.90+06 62 Spent Resin Transfer Pump 2.25+06 5.37+06 8.27+05 1.73+05 4.02+04 3.76+03 2.27+03 83.8 17.8 2.08+04 Filter 63 Spent Resin Transfer Pump 2.25+06 5.37+06 8.27+05 1.73+05 4.02+04 3.76+03 2.27+03 30.5 30.5 2.22+04 64 High-Level Waste Drain Tank 1.61+03 5.51+03 6.70+02 1.46+02 3.59+01 1.21+01 2.48+00 1295.4 320.0 1.01+08 65 Low-Level Waste Drain Tank 7.64-01 7.11+00 3.13+00 1.85+00 3.74-01 9.46-04 1.56-04 340.4 274.3 1.80+07 66 Low Level Waste Drain Pump 7.64-01 7.11+00 3.13+00 1.85+00 3.74-01 9.46-04 1.56-04 35.6 18.4 3.82+03 67 Waste Evaporator Bottoms 3.80+04 1.31+04 1.38+04 1.80+03 5.57+02 2.18+02 6.70-03 335.3 21.9 1.27+05 68 Waste Distillate Cooler 3.80+00 1.31+01 1.38+00 1.80-01 5.57-02 2.18-02 6.70-07 365.8 27.3 2.17+05 Page 1 of 2 Rev

Activity (MeV/cc-Sec) for Energy (MeV)

Source Outside No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 Height (cm) Diameter (cm) Volume (c 72 Waste Demineralizer 1.25+03 1.37+05 9.10+03 4.91+01 1.13+01 1.38+00 4.79-01 206.4 106.7 1.70+06 73 Waste Demineralizer Filter 5.67+02 6.22+04 4.14+03 2.23+01 5.13+00 6.27-01 2.18-01 87.3 17.8 2.17+04 (1) 1.31+03 = 1.31 x 103 Historical, not subject to future updating. This table has been retained to preserve original design basis.

Page 2 of 2 Rev

TABLE 12.2-5 RADIOACTIVE SOURCES IN THE FUEL BUILDING (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)

Height Diameter Vo Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 (cm) (cm) (

69 Most 1.67+11(1) 1.02+12 7.21+10 5.10+11 1.49+10 3.68+10 1.01+09 Radioactive Fuel Assembly 100 Hours after Sutdown 70 Fuel Pool 3.09+02 4.02+05 6.22+04 -- 1.66+03 7.20+01 1.06+01 1.68+02 45.7 2.75 Purification Filter 74 Fuel Pool 5.58+3 9.02+3 1.05+3 1.14+2 5.16+1 2.42+1 4.48-2 (Filled to Capacity 100 Hours after Shutdown)

NOTE:

(1) 1.67+11 = 1.67x1011 Historical, not subject to future updating. This table has been retained to preserve original design basis.

Page 1 of 1 Rev

TABLE 12.2-5A OTHER RADIOACTIVE SOURCES (HISTORICAL)

Activity (MeV/cc-Sec) for Energy (MeV)

Outside Source Height Diameter No. Source 0.40 0.80 1.30 1.70 2.20 2.50 3.50 (cm) (cm)

ESF Building 7 RHR Pump 1.66+05 2.13+05 6.84+04 4.31+04 3.12+04 3.76+04 4.65+03 43.2 99.1 3 (1) 8 RHR Exchanger 1.66+05 2.13+05 6.84+04 4.31+04 3.12+04 3.76+04 4.65+03 1,370 100.0 1 Yard Tanks 71 Boron Recovery 1.61+02 1.34+04 6.38+02 7.98+00 1.37+00 1.69+00 1.98+00 9.14+02 9.14+02 5 Tank 75 RWST 1.68-02 1.17+00 8.11-02 1.09-05 1.89-06 1.19-06 2.27-10 1.80+03 1.80+03 4 Condensate Polishing Building 76 Condensate 8.12+03 3.63+04 5.15+03 1.13+03 3.46+02 1.19+02 1.61+00 3.35+02 2.44+02 5 Polishing Demineralizer 77 Cation Regeneration 3.38+03 7.16+04 8.96+03 4.26+02 1.22+02 6.80+01 3.58+00 4.27+02 2.08+02 2 Tank 78 Anion Regeneration 1.15+04 1.43+04 2.84+03 1.64+03 5.06+02 1.56+02 3.59-01 4.57+02 1.83+02 3 Tank NOTE:

(1) 1.66+05=1.66x105 Historical, not subject to future updating. This table has been retained to preserve original design basis.

Page 1 of 1 Rev

BLE 12.2-6 INVENTORY OF AN AVERAGE FUEL ASSEMBLY AFTER 650 DAYS OF OPERATION AT 3,636 MWt AT SHUTDOWN AND 100 HOURS AFTER SHUTDOWN (Ci) (HISTORICAL)

Isotope 0 Hours 100 Hours

-83m 8.19+10

• 1.07-01

-85m 2.05+11 2.91+04

-85 4.58+09 4.58+09

-87 3.99+11 **

-88 5.60+11 9.17+00

-89 7.25+11 **

-131m 4.15+08 8.55+10

-133m 2.53+10 1.09+10

-133 1.05+12 7.25+11

-135m 2.85+11 9.53+06

-135 2.79+11 1.45+09

-137 9.48+11 **

-138 9.33+11 **

83 8.19+10 2.37-02 84 1.47+11 **

85 2.05+11 **

87 3.95+11 **

29 1.12+04 1.12+04 31 4.72+11 3.38+11 32 6.74+11 2.84+11 33 1.06+12 3.77+10 34 1.23+12 **

35 9.74+11 3.06+07 36 4.89+11 **

-81 2.84+10 **

-83 3.47+10 **

-84 1.47+11 **

-88 5.65+11 1.03+01 Page 1 of 5 Rev. 30

SHUTDOWN (Ci) (HISTORICAL)

Isotope 0 Hours 100 Hours

-89 7.51+11 **

-90 9.17+11 **

-91 8.55+11 **

-92 7.10+11 **

89 7.51+11 7.10+11 90 3.73+10 3.73+10 91 9.17+11 7.10+08 92 8.34+11 5.34+00 93 9.33+11 **

94 7.25+11 **

90 3.70+10 3.70+10 91m 5.39+11 4.66+08 91 9.33+11 8.91+11 92 9.33+11 1.08+04 93 9.64+11 1.07+09 94 8.50+11 **

95 9.64+11 **

95 9.79+11 9.38+11 97 9.33+11 1.48+10

-95m 1.95+10 1.93+10

-95 1.01+12 1.01+12

-97m 8.91+11 1.42+10

-97 9.79+11 1.60+10

-99 9.74+11 3.44+11

-101 7.88+11 **

-102 6.58+11 **

-105 1.42+11 **

99m 8.55+11 3.28+11 Page 2 of 5 Rev. 30

SHUTDOWN (Ci) (HISTORICAL)

Isotope 0 Hours 100 Hours 101 1.58+12 **

102 6.58+11 **

105 1.90+11 **

-103 4.74+11 4.40+11

-105 1.42+11 2.33+04

-106 4.46+10 4.42+10

-107 3.00+10 **

-103m 4.74+11 4.41+11

-105m 1.42+11 2.34+04

-105 1.42+11 2.06+10

-106 4.47+10 4.42+10

-107 3.00+10 **

-127 1.74+10 1.10-04

-128 5.85+10 **

-130 1.74+11 **

-127 2.16+10 1.03+10

-128 7.88+09 **

-129 1.58+11 1.80+04

-130 3.16+11 **

-131 4.26+11 **

-132 5.28+11 **

-133 5.39+11 **

127m 4.67+09 4.95+09 127 4.24+09 1.34+10 129m 8.65+10 7.98+10 129 1.69+11 8.03+10 131m 6.94+10 6.89+09 131 4.15+11 1.39+09 Page 3 of 5 Rev. 30

SHUTDOWN (Ci) (HISTORICAL)

Isotope 0 Hours 100 Hours 132 6.74+11 2.76+11 133m 7.41+11 **

133 4.74+11 **

134 1.09+12 **

-137 3.91+10 3.91+10

-138 1.06+12 **

-139 1.02+12 **

-140 9.33+11 **

-142 5.18+11 **

-137m 3.65+10 3.59+10

-139 1.02+12 **

-140 9.95+11 7.93+11

-141 9.64+11 **

-142 9.12+11 **

-140 1.00+12 8.86+11

-141 9.64+11 1.77+04

-142 9.17+09 **

-143 9.27+11 **

-141 9.59+11 8.86+11

-143 9.33+11 1.15+11

-144 6.79+11 6.68+11

-145 6.11+11 **

-146 4.54+11 **

143 9.27+11 8.24+11 144 6.79+11 6.68+11 145 6.17+11 5.75+06 146 4.66+11 **

-147 3.45+11 2.66+11 Page 4 of 5 Rev. 30

SHUTDOWN (Ci) (HISTORICAL)

Isotope 0 Hours 100 Hours

-149 1.64+11 **

-151 6.63+10 **

-147 1.13+11 1.14+11

-149 1.64+11 4.59+10

-151 6.63+10 5.80+09

-151 5.18+07 5.44+07

-153 2.42+10 5.54+09 TES:

• 8.19+10 = 8.19x1010

• • Less than 1.00x10-6 Ci torical, not subject to future updating. This table has been retained to preserve original design s.

Page 5 of 5 Rev. 30

TABLE 12.2-7 SOURCE INTENSITY IN THE MOST RADIOACTIVE FUEL ASSEMBLY

• AFTER 650 DAYS OF OPERATION AT 3636 MWt (HISTORICAL)

Activity (MeV/Sec) for Energy Group (MeV) ecay ime rs) 0.4 0.8 1.3 1.7 2.2 2.5 3.5 1.77+17 ** 7.83+17 7.91+17 6.47+17 1.35+17 2.15+17 5.57+17 7.45+16 4.86+17 1.38+17 2.23+17 2.68+16 3.66+16 5.75+15 6.69+16 4.11+17 9.75+16 1.74+17 1.88+16 2.14+16 2.24+15 6.11+16 3.56+17 6.76+16 1.38+17 1.11+16 1.21+16 8.81+14 5.47+16 3.03+17 4.70+16 1.13+17 6.24+15 8.33+15 2.84+14 5.21+16 2.84+17 3.51+16 1.08+17 5.38+15 7.91+15 2.25+14 4.97+16 2.69+17 3.03+16 1.04+17 4.86+15 7.70+15 2.02+14 3.93+16 2.15+17 1.86+16 9.58+16 3.62+15 7.17+15 1.85+14 0 2.82+16 1.72+17 1.22+16 8.63+16 2.52+15 6.23+15 1.71+14 8 2.06+16 1.49+17 8.55+15 7.42+16 1.69+15 5.19+15 1.50+14 0 5.81+15 9.49+16 1.97+15 2.13+16 6.81+14 1.40+15 4.68+13 TES:

• Includes a radial peaking factor of 1.65

• • 1.77+17 = 1.77 x 1017 torical, not subject to future updating. This table has been retained to preserve original design s.

Page 1 of 1 Rev. 30

ABLE 12.2-8 RADIONUCLIDE CONCENTRATIONS IN THE SPENT FUEL POOL FROM REFUELING 100 HOURS AFTER SHUTDOWN* (HISTORICAL)

Expected Concentration Nuclide (Ci/cc) Design Concentration (Ci/cc) 31 2.4E-02 2.2E-01 32 1.4E-03 1.3E-02 33 1.8E-03 1.7E-02 35 ** 8.2E-06 89 4.1E-05 4.8E-04 90 1.2E-06 2.0E-05 90 ** 2.2E-05 91 8.1E-06 7.9E-05 95 7.1E-06 8.0E-05

-95m ** 1.7E-06

-95 6.3E-06 8.7E-05

-99 3.7E-03 1.4E-01 99m 3.6E-03 1.3E-01

-103 5.1E-06 3.8E-05

-106 1.2E-06 3.8E-06

-103m 5.1E-06 3.8E-05

-105 ** 1.5E-06

-106 1.2E-06 3.8E-06 127m 3.3E-05 2.4E-04 127 3.3E-05 2.4E-04 129m 1.6E-04 4.2E-03 129 1.6E-04 4.2E-03 131m 3.2E-05 2.7E-04 131 6.4E-06 5.5E-05 132 1.4E-03 1.3E-02

-134 3.2E-03 3.9E-02

-136 1.3E-03 1.6E-02

-137 2.2E-03 1.9E-01 Page 1 of 2 Rev. 30

Expected Concentration Nuclide (Ci/cc) Design Concentration (Ci/cc)

-137m 2.1E-03 1.8E-01

-140 2.2E-05 4.1E-04

-140 2.3E-05 4.0E-04

-141 7.9E-06 7.5E-05

-143 ** 7.5E-06

-144 4.1E-06 5.8E-05 143 5.4E-06 6.9E-05 144 4.1E-06 5.8E-05

-147 ** 2.2E-05

-147 ** 9.9E-06

-149 ** 3.3E-06 51 2.1E-04 2.1E-04

-54 3.8E-05 3.8E-05

-55 2.0E-04 2.0E-04

-59 1.2E-04 1.2E-04

-58 1.9E-04 1.9E-03

-60 2.5E-04 2.5E-04 3 1.2E-01 4.1E-01 TES:

• 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 Ci/cc are considered negligible and are not tabulated.

1.0E-06 = 1.0 x 10-6 torical, not subject to future updating. This table has been retained to preserve original design s.

Page 2 of 2 Rev. 30

Position in Loop Loop Transit Time (sec) Nitrogen-16 Activity (Ci/gm) aving Core 0.0 189 aving Reactor Vessel 1.1 170 tering Steam Generator 1.4 164 aving Steam Generator 5.4 112 tering Reactor Coolant Pump 6.0 106 tering Reactor Vessel 6.8 98 tering Core 9.0 86 aving Core 9.7 189 Nitrogen-16 Energy Emission Energy (MeV/gamma) Intensity (percent) 1.75 0.13 2.74 9.76 6.13 60.0 7.12 5.0

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

torical, not subject to future updating. This table has been retained to preserve original design s.

Page 1 of 1 Rev. 30

TABLE 12.2-10 ASSUMPTIONS USED IN THE CALCULATION OF AIRBORNE CONCENTRATIONS (HISTORICAL)

Containment Turbine Fuel Building Building Building Reactor coolant equilibrium concentrations Table 11.1-2 Secondary side equilibrium concentrations - Table 11.1-6 -

Iodine and noble gas core inventory - - Table 11.1-1 Leak rate into buildings A. Equivalent hot reactor coolant 4.7x103 - -

(lb/day)

B. Equivalent main steam leakage - 1.7x103 -

(lb/hr)

Normal moisture in atmosphere (%) 60 - -

Fraction of primary coolant activities released (%/day)

A. Noble gases 1.0 - -

B. Iodines 0.001 - -

Mixing in building atmosphere (%) 70 100 100 Building ventilation rate (cfm) 3.0x104 1.55x105 3.0x104 Building free volume (ft3) 2.32x106 4.06x106 2.30x105

• Recirculation - filters Yes No No Filter efficiency 99% - -

Fuel pool evaporation rate (lb/hr-ft2) - - 1.74 Recirculation rate (cfm) 2.4x104 - -

Fuel pool average volume (ft3) - - 4.88x104 TE:

Only the area above the fuel pool.

torical, not subject to future updating. This table has been retained to preserve original design s.

Page 1 of 1 Rev. 30

TABLE 12.2-11 AIRBORNE CONCENTRATIONS INSIDE MAJOR BUILDINGS (CI/CC) (HISTORICAL)

Containment Building Containment Building After 16-Hour Prior to Recirculation Recirculation (1) Turbine Building Fuel Building Isotope Design Expected Design Expected Design Expected Design Expecte H-3 1.6E-4(2) 4.6E-5 1.6E-4 4.6E-5 1.2E-8 2.9E-9 3.8E-6 1.1E-6 I-131 9.8E-7 1.1E-7 9.1E-9 1.0E-9 3.8E-11 2.9E-13 2.1E-10 2.3E-11 I-132 4.1E-9 2.3E-10 1.7E-9 2.2E-10 9.8E-12 7.7E-14 5.0E-12 5.4E-13 I-133 1.7E-7 1.7E-8 1.2E-8 1.3E-9 5.5E-11 4.0E-13 1.8E-12 1.9E-18 I-134 9.8E-10 9.8E-11 6.5E-10 6.5E-11 1.8E-12 1.3E-14 - -

I-135 2.9E-8 2.9E-9 5.6E-9 5.6E-10 2.32E-11 1.7E-13 2.7E-16 -

Kr-83m 1.6E-6 8.0E-8 1.6E-6 8.0E-8 4.1E-12 1.5E-14 - -

Kr-85m 1.5E-5 7.9E-7 1.5E-5 7.9E-7 1.7E-11 6.6E-14 - -

Kr-85 1.0E-4 6.1E-6 1.0E-4 6.1E-6 3.5E-13 1.7E-15 - -

Kr-87 3.0E-6 1.6E-7 3.0E-6 1.6E-7 1.0E-11 4.0E-14 - -

Kr-88 1.9E-5 1.0E-6 1.9E-5 1.0E-6 3.2E-11 1.3E-13 - -

Kr-89 1.1E-8 6.4E-10 1.1E-8 6.4E-10 1.7E-13 7.0E-16 - -

Xe-131m 6.2E-2 3.1E-6 6.2E-2 3.1E-6 1.2E-13 4.4E-15 2.0E-8 2.2E-9 Xe-133m 6.5E-5 4.3E-6 6.5E-5 4.3E-6 6.4E-12 3.2E-14 4.2E-12 4.3E-13 Xe-133 6.5E-3 4.3E-4 6.5E-3 4.3E-4 2.8E-10 1.3E-12 1.8E-10 1.8E-11 Xe-135m 6.2E-7 9.0E-9 6.2E-7 9.0E-9 9.3E-12 3.3E-14 6.2E-15 -

Xe-135 9.2E-5 4.0E-6 9.2E-5 4.0E-6 5.2E-11 1.7E-13 2.6E-14 -

Page 1 of 2 Rev

Containment Building Containment Building After 16-Hour Prior to Recirculation Recirculation (1) Turbine Building Fuel Building Isotope Design Expected Design Expected Design Expected Design Expecte Xe-137 2.1E-8 1.4E-9 2.1E-8 1.4E-9 3.1E-13 1.5E-15 - -

Xe-138 2.8E-7 2.5E-8 2.8E-7 2.5E-8 2.8E-12 1.8E-14 - -

NOTES:

(1) Using 99% filter efficiency (2) 1.6E-4 = 1.6 x 10-4 Historical, not subject to future updating. This table has been retained to preserve original design basis.

Page 2 of 2 Rev

.1 SHIELDING assessments performed to determine the original major shield designs were based on assumed rce terms, occupancy times and acceptance criteria based on zone criteria. Although these eria were used to establish the original shield design, they were never intended to establish uirements for the radiation protection program implementation during plant operation. As time lves, source terms change. Acceptable doses have typically decreased with time as ambitious ARA person-REM goals are established.

rent shielding requirements are non-specific and are established through the implementation he Radiation Protection Program and ALARA Program. These programs evaluate the need for mbination of exposure saving principals such as reduced source term, decreasing occupancy e, or increased shielding. These programs use shielding as one method to help ensure pliance with 10 CFR 20.

s section provides the basis for the original plant shielding design. Although current dose rates not be consistent with the zone maps in this chapter, these maps are not being changed to be ent, as that would make them inconsistent with the original design basis criteria for the lding. Recent Heath Physics surveys should be consulted for information on current station ological conditions.

iation shielding is designed to ensure that radiation exposure to the general public and to onnel in-plant is kept to levels as low as is reasonably achievable (ALARA), consistent with requirements set forth in 10 CFR 20 for normal operation and 10 CFR 50.67 for accident ditions and with the overall objectives set forth in USNRC Regulatory Guide 8.8. The original gn of this radiation shielding was based upon radiation zone criteria which were established in port of the expected access requirements and durations of occupancy during normal rations, during refueling outages, and during accident situations. Descriptions of the zone eria are presented in Table 12.3-1, and the detailed radiation zone criteria for normal and tdown operations are illustrated on Figures 12.3-1 through 12.3-4 and 12.3-6 through 12.3-9.

se figures do not represent operational requirements.

iation shielding is provided on the basis of maximum concentrations of radioactive materials hin each shield region (e.g., 1 percent failed fuel at the original design basis core power level 636 MWt) rather than the annual average values. For batch processes, as an example, the nt of highest radionuclide concentration in the batch process is assumed (e.g., just prior to ning of a tank). The shielding designs are, therefore, intentionally conservative in that the e rates reflect maximum, rather than average, sources to be shielded. These maximum dose s are based on anticipated occupancy requirements and are set such that the maximum osure of plant personnel is within the limits set by 10 CFR 20. The average exposures are ected to be a small fraction of the limiting values because it is not expected that the plant ld run at 1 percent failed fuel with all tanks full to capacity, all demineralizer beds at ration, etc.

12.3-1 Rev. 30

g vertical shield surfaces opposite the most intense source in the vicinity. These calculations based on the inherent assumption that plant personnel spend the required time in each zone in tact with the shield at this point. This is a demonstrably conservative approach, since the dose actually decreases dramatically as the dose points are moved along the surface of the shield to the slant penetration involved. The additional reduction of intensity with distance is also ored by this approach.

shield wall thicknesses are derived from design basis fuel defect of 1 percent and dose rate tation of adjacent zones and are expected to provide adequate protection for abnormal ditions which may occur during normal plant operations.

e designations are based on the annual occupational exposure limits, access requirements and upancy time for the specific location in the plant as described in Table 12.3-1.

the yard areas, the shield walls are designed to meet the Zone I criterion of 0.25 mRem per

r. The most significant structures which contribute to the yard dose rate are the containment, building, waste disposal building, auxiliary building, refueling water storage tank, and the on recovery tanks.

calculated dose rate levels in the unrestricted areas are based upon full power normal plant rations assuming fuel defects producing expected quantities and concentrations of onuclides consistent with NUREG-0017. At the site boundary, the calculated dose rate is roximately 0.43 mRem per year.

e rates are generally calculated at three and six foot levels above walking surfaces, icularly if significant sources are located on the next level above the zone within the building.

e rates for post-shutdown conditions are computed at the earliest reasonable time after tdown. Subsequent decay is ignored for conservatism; i.e., the dose rate at that point in time is ted despite the fact that the radiation levels continue to decay to lower values.

nsit times in coolant loops, etc, are computed as precisely as practicable with no intentional servatism. Simplified models which are used to describe large components are intentionally ised not to overestimate component self-shielding. This provides an amount of conservatism ch varies from component to component. In some instances no component self-shielding is uded.

modeling reflects the knowledge of specific components and limitations which exist when ils on components are not available. The models allow for this uncertainty in a conservative ner, thus ensuring that the actual radiation leakage from the supplied component is less than qual to predicted values.

elding in the Millstone 3 plant was designed using Stone & Webster Engineering porations topical report, Radiation Shielding Design and Analysis Approach for Light Water 12.3-2 Rev. 30

ulatory Guides 1.69 and 8.8. RP 8A defines the assumptions, codes, techniques, and meters used in calculating shield thickness, material, and placement.

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

primary shield is designed to:

1. attenuate neutron flux to minimize activation of plant components and structures;

2. reduce residual radiation from the core to a level that allows access to the region between the primary and secondary shields at a reasonable time after shutdown; and

3. optimize the combination of primary and secondary shielding by reducing the radiation level from the reactor so that it is commensurate with radiation levels from other sources.

primary shield consists of a water filled neutron shield tank and a 4.5 foot thick reinforced crete shield wall. The neutron shield tank has an annular thickness of 3 feet and is located ween the reactor vessel and the concrete shield wall. To maintain the integrity of the primary ld, a streaming shield fabricated from borated silicon rubber (Dow Corning Sylgard 170 on elastomer or equivalent) is installed in the upper annular gap between the vessel flange and neutron shield tank and around the nozzles. (Refer to Figure 12.3-12.)

s shield is designed to minimize the leakage of neutrons to the annular region and streaming to upper levels of the containment, thus reducing the neutron dose rate on the operating floor, ng normal operations, to acceptable levels.

as estimated that the neutron dose rate in the annulus area between the containment wall and crane wall at the operating floor level would not exceed 5 mRem per hour with the shield in

e. Radiation protection surveys should be consulted to determine actual neutron dose rates.

.1.2 Secondary Shielding ondary shielding consists of reactor coolant loop shielding, the crane wall, containment cture shielding, fuel handling shielding, auxiliary equipment shielding, waste storage lding, control room shielding, and yard shielding.

ondary shielding thicknesses within the containment structure are based on nitrogen 16 being major source of radioactivity in the reactor coolant during normal operation. This source blishes a required shielding thickness of the reactor coolant loop shielding, crane wall, and tainment structure wall. The shutdown radiation levels in the reactor coolant loop cubicles are 12.3-3 Rev. 30

crane wall provides shielding for limited access to the annulus between the crane wall and the tainment structure wall and provides additional exterior shielding during power operation.

containment structure shielding consists of a steel lined reinforced concrete cylinder and ispherical dome. This shielding, together with the crane wall, attenuates radiation during full er operation and during an accident. This shielding keeps radiation levels within acceptable ls at the outside surface of the containment structure and at the exclusion area boundary B).

fuel handling shielding, including both water and concrete, attenuates radiation from spent assemblies, control rods, and reactor vessel internals to acceptable levels and permits the oval and transfer of spent fuel and control rods to the fuel pool in the fuel building.

refueling cavity above the reactor is formed by a stainless steel-lined, reinforced concrete cture. This refueling cavity becomes a pool when filled with borated water to provide lding during the refueling operation.

depth of the shielding water in the cavity is such that the radiation dose rate at the surface of water from a spent fuel assembly should not exceed approximately 2.5 mRem per hour during short time intervals when the fuel handling operation brings the spent fuel assembly to its est approach to the pool surface.

cavity is large enough to provide storage space for the upper and lower internals and cellaneous refueling tools.

fuel pool in the fuel building is filled with water for shielding as discussed in Section 4.3.4. The fuel pool walls are a minimum of 6 foot thick concrete to ensure a dose rate less 0.75 mRem per hour outside the fuel building and less than 2.5 mRem per hour inside both fuel building and the adjacent auxiliary building from the fuel stored in the pool.

rder to preclude unacceptable radiation dose rates during fuel transfer, a special radiation ld, fabricated from carbon steel has been provided inside containment, where the fuel transfer traverses the gap between the containment wall and the refueling cavity wall. (Refer to ure 12.3-11). The design basis for the shielding concept is a dose rate at the surface of the ld of approximately 50 mRem per hour and a dose rate at the personnel access hatch of roximately 5 mRem per hour.

side containment, the fuel transfer tube is inaccessible to personnel by means of backfill ering the transfer tube and a security fence between the containment and the fuel building, ring limited access to this area.

ee radiation monitors with local audible and visible alarms as well as remote alarms in the trol room, are used to monitor fuel transfer operations. Two radiation monitors are located in 12.3-4 Rev. 30

iliary building components may exhibit varying degrees of radioactive contamination due to handling of various fluids. The function of shielding in this building is to protect operating and ntenance personnel working near the various auxiliary system components, such as those in makeup and purification system, the boron recovery system, the radioactive liquid and eous waste systems, and the sampling system.

ically, major components of systems are individually shielded so that compartments may be red without having to shut down and possibly decontaminate the entire system. Potentially hly contaminated ion exchangers and filters are located in individual shielded cells in the iliary building. The concrete thicknesses provided around the shielded compartments was ed on reducing the surrounding area dose rate to less than approximately 2.5 mRem per hour the dose rate to any adjacent cubicle to less than approximately 100 mRem per hour.

ome areas, tornado missile protection in the form of concrete affords more shielding than that uired for radiation protection.

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

on recovery tanks, which may be used to store letdown prior to its recycling to the plant or cessing as waste, are shielded to reduce dose rates to accessible levels within the yard area.

.1.3 Accident Shielding

.1.3.1 Containment and Control Room Design ident shielding is provided by the containment structure, which is a reinforced concrete cture lined with steel. For structural reasons, the thickness of the cylindrical wall and the e are 54 inches and 30 inches, respectively. These thicknesses are more than adequate to meet shielding requirement during accident conditions.

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

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

Millstone 3 Control Room has also been evaluated for the 30 day dose due to a postulated CA at Millstone 2 and those results are within the limits of GDC-19. Shielding calculations w that the 2 foot thick concrete walls which enclose the control room are sufficient to ensure the radiation dose inside the control room remains below the radiation design basis during 12.3-5 Rev. 30

stant.

.1.3.2 Post-Accident Access to Vital Areas diation and shielding design review was performed in accordance with NUREG-0737, Action II.B.2 (USNRC, 1980), in order to ensure personnel accessibility after a design-basis dent (DBA). The DBA considered for this evaluation was the loss-of-coolant accident CA). The projected dose to complete each activity necessary to mitigate a DBA LOCA, en e to and in vital areas, is less than the 5 rem design limit of NUREG-0737. At Millstone 3, this uirement is met by providing sufficient shielding of components containing post-accident oactive inventories, consistent with anticipated access routes and stay times.

as requiring accessibility (vital areas) are those areas where post-LOCA actions can be taken r the short-term to ensure the capability of operators to control and mitigate the consequences n accident. A description of the post-accident activities is summarized below and in Table

-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 event that the reactor failed to trip. This action must take place as soon as possible. Thus, the 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 Fuel Pool Cooling, RBCCW and Service Water for spent fuel pool cooling FSAR 9.1.3.3 states that spent fuel pool cooling will be initiated 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 Computer The Plant Process computer is normally not powered from an Emergency Bus. It is powered from an uninterruptible power supply that may last for only 30 minutes.

Thus, the 0 to 30 minute category is assumed. 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 valves For post-LOCA cooldown and depressurization, the SI accumulator isolation valves are closed to prevent injection of nitrogen that might interrupt natural 12.3-6 Rev. 30

preparation for a cooldown, the 30 minute 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 system 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 /> and 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. Deleted

7. Deleted

8. Repower Monitor and Maintain the porous concrete groundwater removal system A non-safety related pump (3SRW-P5) is credited with groundwater removal that circumvents the waterproof membrane that surrounds the containment structure and the containment structure 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 design 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 requires 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 removal system will be monitored and operated several times a day. These activities 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 assumed 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 related 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 achieved via the Hydrogen Recombiner Building stairway.

9. Open the breakers for the non-safety grade sump pumps 12.3-7 Rev. 30

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 MCCs may be de-energized at its Load Centers in the 4 foot elevation of the Service Building. Therefore, local operator 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 available. In the longer term, the AFW pumps can be aligned to the condensate storage tank (CST).Travel Route 8 reflects the 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 power is available and the running diesel generator is stopped. The keep warm system assures that the diesel generator would be maintained in the optimum condition 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.

ddition to the areas and activities defined above, the Control Room and Technical Support ter (TSC) require post-accident access and continued occupancy as discussed in NUREG-7.

t-accident control room habitability is discussed in Section 6.4. The post-accident dose sequences for the Control Room are presented in Table 15.0-8.

potential radiation doses to a person occupying the TSC have been evaluated for the Unit 3 CA. The TSC is designed for continuous operation for the duration of the accident (i.e., 30 s). The building roof and walls provide adequate shielding to protect the occupants against ct radiation from the external radioactive cloud and from the containment during the tulated LOCA. Double vestibule doors are provided at the building entrance to minimize akage due to personnel ingress/egress. The TSC ventilation system is described in Section

13. The evaluated 30 day integrated dose for an individual occupying the TSC following the A is within the NUREG-0737 criteria of 5 rem whole body dose or equivalent.

12.3-8 Rev. 30

le 12.3-4 provides an estimate of the anticipated times after a LOCA that vital area access is uired, with consideration given to the typical 30 minute minimum time frame assumed for rator action outside control room and the X/Q intervals assumed in the FSAR Chapter 15 dent analysis. Outside travel routes are shown on Figure 12.3-10 and are listed on le 12.3-3. A general description of the ingress travel routes, primary and alternates, are cribed below (the egress path is the same as the ingress path except for alternate routes to the kup Chemistry Laboratory in travel route 4).

vel route 1:The primary route is from the control building through the service building to the auxiliary building (no outdoor travel). The alternate route 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.

vel route 2:The primary route is from the control 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 fence). 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.

vel route 3:The primary route is from the 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 building to the turbine building to the RR loading area to the RCA gate adjacent to the HRB.

vel route 4:The primary route is from the 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 building 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 which is on the egress path. Figure 12.3-10, sheet 4, provides two additional routes for sample analysis in the MP1/MP2 service building.

vel route 5:The primary route is from the control 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 disposal building, then entering the fuel building. The alternate route 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, north to the fuel building.

12.3-9 Rev. 30

vel route 7:The primary route is from the control building to the emergency diesel generator building. No alternate is given since doses would only increase with any other route.

vel route 8:The primary route is from the control building through the service building to the exit between the service and auxiliary buildings, along the north side of the waste disposal building, south to the ESF building or the turbine driven Auxiliary Feedwater Pump Room. The alternate route 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.

vel route 9:The primary path is from the control building through the service building corridor to the turbine building auxiliary bay, lower level then across the road to the auxiliary building. The alternate path is from the control building to the service building, past the Chemistry Laboratory, exit the service building to the auxiliary building.

following general assumptions and criteria are used as a basis for review of all vital areas and ess routes as applicable:

1. The starting point for all activities 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 the 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 direct shine and skyshine contributions) and (b) direct radiation from the overhead plume.

eous and liquid LOCA source terms used in the review are not less than that stated in REG-0737,Section II.B.2, which provides the minimum source terms to be used for luation of the adequacy of radiation protection to the operators.

determine post-accident doses to personnel for performance of and transit to identified vities, the following sources of radiation are considered.

1. Auxiliary Building 12.3-10 Rev. 30

• Radio iodine buildup in the SLCRS filter.

• Sump water in the safety injection 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 piping

4. Along routes from control building to the vital areas.

• Skyshine from containment

• Direct shine from containment.

• Plume shine.

• Direct shine from the RWST.

tems containing sources of radiation which are identified in NUREG-0737 but which have not n identified in buildings discussed above are considered to be either irrelevant following an 12.3-11 Rev. 30

dent occurs, the only use of this system is for post-LOCA hydrogen purge as the result of a ond design basis event.

results of the dose calculations indicate that the plant shielding and design provide adequate ection to operators following a design basis LOCA to ensure compliance with the NUREG-7 design dose requirements.

.2 FACILITY DESIGN FEATURES Millstone 3 design is consistent with the guidance presented in Regulatory Guide 8.8, ision 4, C2, which discusses specific features in the facility and equipment design that limit ation exposure to levels that are ALARA. The following features have been incorporated.

.2.1 Location and Design of Equipment to Minimize Service Time he auxiliary building, nonradioactive equipment, such as the reactor plant component cooling em and components used to process the waste evaporator distillate, are located outside high ation cubicles in areas designated as Radiation Zones II or III (defined in Table 12.3-1). In the tainment structure, nonradioactive equipment requiring servicing is typically located in iation Zone IV areas. Exceptions include those components attached to the reactor coolant em, such as the reactor coolant pump motor cooling equipment and the equipment support bbers.

or radioactive components which may require servicing are typically located in individually lded cubicles. These cubicles are designed such that radiation contributions from adjacent icles is small compared to sources within the cubicle. The resultant dose rate in any cubicle in ch equipment is being serviced is due to sources within the cubicle to radiation penetrating ugh shield walls from adjacent cubicles, and to radiation streaming through shield wall etrations. The design basis for shield walls enclosing cubicles containing process equipment is ussed in Section 12.3.1. Shield wall arrangement and dimensions are shown for the tainment Building, Figure 3.8-60; Auxiliary Building, Figure 3.8-62; Fuel Building, Figure 63; and Waste Disposal Building, Figure 3.8-74.

icle access openings generally incorporate a labyrinth design which precludes direct radiation

e. The openings are sized to allow for removal and replacement of minor fluid system ponents such as pumps and valves, as well as to provide access for maintenance equipment.

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

12.3-12 Rev. 30

sible. Platforms for servicing specific components are provided where necessary.

tain components have design features which minimize service time. For example, the reactor lant pump design includes an assembled cartridge seal which results in reduced time required replacement. The cartridge seal is also expected to have a useful life which is double that of older designs. The reactor coolant pump design also includes a spool piece to facilitate aration and replacement of the motor from the pump.

reactor vessel nozzle welds insulation is fabricated in sections with a thin reflective metallic et covering and quick disconnect clasps to facilitate removal for inspection of the welds.

ically, filters are designed to be removable from the top with lifting bails in the middle of the

d. The filter assemblies usually have bolt lead-ins for tool entry, and the filters are contained in osable cartridge assemblies. These features facilitate remote removal, disposal, and assembly.

head closure system provided for Millstone 3 includes quick disconnect/connect stud ioners which have quick-acting, hydraulically-operated stud gripper devices, as opposed to ventional tensioners which must be threaded onto the tops of the studs. Also provided are air-or driven stud removal tools which can rapidly remove (or insert) the studs, in contrast to the h slower manual stud removal and insertion tools used in older designs. The stud tensioners designed to operate simultaneously, as are the stud removal tools.

primary system heat exchangers are designed such that the shell-to-tube sheet joint need not roken for inspection. The shell and tube assembly can be lifted intact above the channel head xpose the tube ends for inspection and leak testing.

ps are typically designed with flanged connections to facilitate removal for maintenance.

ending on expected conditions, either canned pumps or pumps with high quality mechanical s are used to reduce leakage and maintenance requirements.

.2.2 Location of Instruments Requiring In Situ Calibration ruments which require in situ calibration are located, wherever possible, on exterior walls of lded cubicles to minimize exposure of instrumentation and personnel. Instruments which not be located in this manner are located in the lowest practicable radiation area in the cubicles are provided with convenient access. Where practical, instruments are designed for removal ow radiation areas for calibration and maintenance.

.2.3 Location of Equipment Requiring Servicing in Lowest Practicable Radiation Field (or Movable to Lowest Practicable Radiation Field) ndicated above, radioactive equipment requiring servicing is typically located in shielded icles with access openings sized for ease of equipment removal. As an example, pump icles are designed to allow removal of the pump to the lowest practicable radiation field.

12.3-13 Rev. 30

et in the steam generator channel head to eliminate a potential crud trap. The steam generator ways (entrance to channel head) are sized to facilitate entrance and exit with protective hing. Handholes to the secondary side are positioned to facilitate maintenance operations.

nges to increase steam generator reliability also reduces occupational radiation exposures.

h changes include improved steam generator tube support plates (stainless steel and quatrefoil holes) and the use of all-volatile treatment chemistry on the secondary side.

.2.4 Valve Location and Selection ves are located in separate shielded valve cubicles or areas outside equipment cubicles to the test extent practicable to minimize maintenance exposure. Valve selections are usually based best product available and maintenance time required. Westinghouse has supplied valves of bolted body-to-bonnet forging type. This permits the use of ultrasonic testing in place of ography for inspection and facilitates assembly and disassembly, resulting in reduced ection and maintenance time. Additionally, manual valves under 2 inches in diameter are gned for zero stem leakage.

.2.5 Penetrations of Shielding and Containment Walls by Ducts and Other Openings re are numerous piping penetrations through shield walls in the auxiliary building which are cted into adjacent cubicles, into the pipe chases for radioactive piping, and into the corridors nonradioactive piping. To the greatest extent practicable, penetrations through walls arating higher radiation zone areas from lower radiation zone areas are located above head l, in corners, and in positions which are offset from radiation sources in the higher radiation e cubicles. This prevents line-of-sight radiation streaming from significant radiation sources to onnel working in adjacent cubicles. Noteworthy examples of this practice are provided as ows.

1. Electrical penetrations through shield walls are made to prevent direct line-of-sight to any significant radiation sources.

2. Instrument tubing penetrations through shield 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.

.2.6 Radiation Sources and Occupied Areas iation sources (Section 12.2) are separated, as far as is practicable, from normally occupied s by shield walls and cubicles. Piping runs are also located as far as practicable from 12.3-14 Rev. 30

s wherever practicable. Valves are located in shielded valve areas where practical and are arated from equipment cubicles, pipeways, and areas of general access.

sically locked barriers (i.e., locked doors) are provided for areas having radiation levels in ess of criteria as specified in the Technical Specifications.

.2.7 Minimizing Spread of Contamination and Facilitation of Decontamination Following Spills ically sources of contamination from leaks or spills from components located in cubicles are vented from spreading by cubicle entrance dikes and/or low point drains to enclosed collection ps. Floor surfaces and walls are sealed or painted as required with a protective chemically-stant coating to provide a surface which is easily decontaminated. Demineralized water hose ions are provided throughout the auxiliary building to allow flush water to be available to each icle in the auxiliary building. Systems containing radioactive fluids are usually fabricated of osion-resistant materials.

borne contamination is kept from spreading by ventilation systems which are described in tion 12.3.3. A personnel decontamination area is located in the radiation protection area. An ipment decontamination area is located in the waste disposal building.

.2.8 Piping to Minimize Buildup of Contamination rior surfaces of systems in radioactive liquid service typically are made of stainless steel or r corrosion-resistant material.

piping associated with these systems is normally routed to avoid sharp bends by carefully cting the elevation between points and by attempting to run this piping at no more than two ations between these points. Pockets and low points are also avoided. Pipe runs for spent resin cing are provided with large radius bends rather than welded elbows to prevent accumulation esin fines and crud particles. Resin piping is also butt-welded where possible to minimize the ntial for crud particles. Valve stations are designed to minimize the buildup of crud by imizing the number of pockets and stagnant vertical legs.

tilation design features to minimize radioactive contamination buildup are discussed in tion 12.3.3.

3.2.9 Flushing or Remote Chemical Cleaning of Contaminated Systems ans for flushing and draining of potentially highly radioactive tanks, lines, and other ponents are considered in fluid system design. Waste collection tank design includes visions for internal flushing with spray nozzles to remove potential collections of particulate erial. All heat exchangers are provided with chemical cleaning connections which are nected prior to servicing.

12.3-15 Rev. 30

.2.10 Ventilation Design ventilation systems are designed with sufficient capacity to control airborne radioactivity ases and concentrations during normal and maintenance conditions. The ventilation flow ugh equipment cubicles is based upon unrestricted air flow from general access areas into e cubicles. The design of ventilation systems typically ensures a positive flow from non-taminated areas to potentially contaminated areas to prevent the spread of airborne oactivity and to exhaust from the potentially contaminated areas. A more explicit description entilation systems is given in Section 12.3.3.

.2.11 Radiation and Airborne Contamination Monitoring a and airborne radiation monitoring ensures that any substantial abnormal radioactivity ase is promptly detected. The area and airborne radiation monitoring system is described in e detail in Section 12.3.4 and Section 11.5, respectively.

.2.12 Temporary Shielding use of temporary shielding to facilitate maintenance tasks is considered on a case-by-case

s. Convenient means for transport and placement of such shielding are provided by access idors and elevators in the auxiliary building and an elevator in containment.

.2.13 Solid Waste Shielding shown on Figures 12.3-1 through 12.3-4, radioactive wastes in tanks, evaporators, process gas rcoal bed adsorbers and associated equipment are located in shielded cubicles. Solid waste is lded both by the storage area walls in the waste disposal building and by individual sportation shields.

.2.14 Remote Handling Equipment noted in Section 12.3.2.1, filters are designed for remote removal, disposal, and assembly.

ipment is provided for filter handling as well as for remote removal and replacement of ion hange resins.

ions for potentially radioactive system valves are, in general, arranged either in segregated lded cubicles away from the equipment served and/or are provided with reach rods. The ineralizer and filter valves are in cubicles beneath the vessels and are also provided with h rods.

.2.15 Maximum Expected Failures of Fuel Element Cladding and Steam Generator ign features such as shielding and radiation zones accommodate 1 percent fuel defects and mary to secondary steam generator tube leaks of 1,370 lb per day.

12.3-16 Rev. 30

ple points are provided with sample sinks and ventilation hoods, splash screens, and valves ted outside each splash screen. Samples are provided with recirculation paths behind shield ls at sample sinks, with reach rods for operators.

.2.17 Cobalt Impurity Specifications alt weight percentages for materials in contact with reactor coolant are considered in purchase cifications.

.2.18 Reactor Cavity Filtration System ing refueling, the reactor cavity water may become turbid, making it difficult to observe the oval and replacement of fuel assemblies. The portable reactor cavity filtration system, sisting basically of a pump and four filters, provides capability for cleanup of this water, thus imizing the time required for, and dose due to, fuel and equipment handling operations.

.3 VENTILATION s section provides the basis for the original plant ventilation design. Although current airborne ls may not be consistent with the tables in this chapter, these tables are not being changed to urrent as that would make them inconsistent with the design basis criteria for the ventilation ems. Recent radiation protection surveys should be consulted for information on current ological conditions.

.3.1 Design Objectives function and design bases of the ventilation systems are given in Section 9.4. Consistent with e, the following specific objectives pertain to radiation protection and the commitment that upational 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, Table 1 of 10 CFR 20, Appendix B. Radwaste piping system and process components in the auxiliary 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 radiation protection personnel provide appropriate controls and protective measures described in Section 12.5.3 when access is needed to areas which are 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.

12.3-17 Rev. 30

4. The containment atmosphere filtration system, with only one of its two 12,000 cfm fan units in operation, is capable of reducing 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 expected reactor coolant radioactivity concentration and leakage described in NUREG-0017.

5. The containment purge air system is capable of reducing airborne radiation levels in the containment to acceptable levels prior to and during extended personnel occupancy of the containment.

6. The fuel building ventilation system operates in the 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 terminating or filtering the ventilation flow upon a high radioactivity alarm.

8. Systems are designed so that filters containing radioactivity can easily be maintained to minimize the radiation dose to personnel.

.3.2 Design Description ailed descriptions of the ventilation systems for the plant buildings which contain radioactivity otentially radioactive systems are given in the following sections:

Section Title 9.4.1 Control room area ventilation system 9.4.2 Spent fuel pool area ventilation system 9.4.3 Auxiliary building area ventilation system 9.4.5 Engineered safety feature ventilation system 9.4.4 Turbine building area ventilation system 9.4.7 Containment ventilation 9.4.9 Waste disposal building ventilation system 9.4.10 Main steam valve building ventilation system 9.4.11 Hydrogen recombiner building ventilation 9.4.13 Technical Support Center building ventilation system 12.3-18 Rev. 30

recommendations of Regulatory Guide 1.52, as described in Section 1.8, are implemented in design of the safety-related ventilation filter trains to help assure that occupation radiation osures from service of these trains are ALARA.

s is accomplished by utilizing 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 corridors from the filter trains to the elevators and equipment hatches are provided for transport 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 side, and minimal radiation exposure of personnel. A portable cart-mounted vacuum conveying system is provided for draining and recharging gasketless-type charcoal adsorbers. Contaminated 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 more 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 maintained 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 provided for initial and periodic field testing.

er train arrangement is discussed in Section 6.4.

12.3-19 Rev. 30

centrations of airborne activity for the expected and design conditions in the containment cture, turbine building, and fuel building are tabulated in Table 12.2-11. The concentrations ed upon design conditions are expected to envelope anticipated operational occurrences. The orne concentrations are averages based on assumed total leak rates described in NUREG-7 and the ventilation rates for the respective buildings. Corridors and areas normally occupied perating personnel are expected to have negligible airborne activity concentrations since n air ventilation flow is typically directed from areas with less potential for contamination nned areas) to areas with greater potential for contamination.

ipment cubicles are the most likely areas for airborne concentrations but are not normally upied or accessible without prior survey and control. For purpose of quantification, the worst orne concentration could conceivably exist in cubicles for which the combination of relatively h system volatile radionuclide concentrations and low cubicle ventilation rate would ultaneously exist for a given leak rate. A cubicle such as the letdown heat exchanger cubicle in auxiliary building could develop airborne concentrations of approximately 7x10-5 Ci/cc ming all the design basis leak rate takes place in that cubicle for expected coolant oactivity concentrations.

ed on the above assumption, it is expected that other cubicles would have airborne centrations of less than 7x10-5 Ci/cc.

tion 12.2 includes the models and parameters used as a basis for calculated radioactivity es.

.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING

.4.1 Purpose area radiation monitoring system (RMS) works in conjunction with the process, effluent, and orne radiation monitoring group (Section 11.5). Its purpose is to protect plant personnel by suring levels of radiation in various areas of the plant. It also provides a warning to operations bnormal radiological conditions. If high radiation levels are monitored, the system sounds an

m. It also produces a record of radiation levels.

.4.2 System Design basis of the RMS area radiation monitoring group is the single channel, GM tube or ion mber detector equipped with a dedicated microprocessor except the containment high range nitors which are analog and in designated cases, a rate meter. The microprocessor provides l display and control functions for the detector, computes and stores time-averaged detector puts, stores all necessary operating parameters (e.g., alarm trip values), and also handles all munication between the detector and the RMS computer system. The rate meter, where vided, is located adjacent to the detector and provides local analog display. A high activity 12.3-20 Rev. 30

ditions could cause significant increases in personnel exposure rates in accordance with gn criteria established in ANSI/ANS HPSSC-6.8.1-1981.

RMS computer system provides centralized display and control, at the control room RMS kstations. A dual server based computer system, located in the control building, polls each nitors microprocessor every several seconds to obtain the latest readings, and to register any ms present. All alarms are displayed on the control room RMS workstation and can be ted. Radiation alarms are annunciated both in the control room and if equipped with local cation locally at the microprocessor, and can be acknowledged at either location. The system rator also uses the RMS workstations to either output data from individual monitors or input mands to these non-Class 1E monitors. All commands sent are recorded in the message mary log. The last (30) 1 minute, 10 minute and hourly averages are stored and available for ew at the RMS workstation for all radiation monitors except the containment high range. An S workstation is located in the radiation protection Office.

ddition, those monitors designated Class 1E except that containment high range monitors are nected directly to one of two control room 1E cabinets. The output of each monitor is digitally layed and also recorded. A remote indication and control module (RIC) is furnished in the inets for each 1E monitor. The RIC handles all remote control functions for 1E monitors. The tainment high range monitors are displayed on the 1E control room cabinets and recorded on plant process computer. The 1E cabinets are connected by electronic isolators to the RMS puters to allow data from the 1E monitors to be displayed on the control room RMS kstations and to be written into the RMS computer.

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

tory calibration included checks for linearity and energy response. Sources traceable to onal standards are used. Onsite calibration includes detector response using sources of known rgy and strength. The frequency of onsite calibration of safety-related monitors is provided in Technical Specifications.

le 12.3-2 gives the mark numbers, names, locations, and ranges of the area monitors in the S. The following paragraphs provide a brief description of the different types of area nitors.

3.4.3 Class 1E Area Monitors r of the area monitors are designated Class 1E. (Section 8.3.1.1.2 discusses Class 1E power its backup supply.) These are the two redundant fuel drop monitors and the two containment rnal high range monitors. They differ from most other area monitors in that they use ion mber detectors instead of GM detectors. The high range monitors are capable of withstanding sign basis accident inside containment. The fuel drop monitors are designed to operate in the mal containment environment and are discussed in Section 11.5.

12.3-21 Rev. 30

roximately 180° apart. Range is 1 to 108 R/hr.

se monitors are qualified based on the requirement of IEEE 323-1974, 279-1971, and 344-5.

to the normally high operational dose rates inside containment, the dedicated roprocessors for all detectors located inside the containment structure are located within the iliary building.

with the Class 1E process monitors (Section 11.5), the outputs of these devices are displayed recorded on the control room Class 1E panels. The display and recording of the containment h range monitor is as required by Regulatory Guide 1.97.

.4.4 Non-Class 1E Area Monitors

-Class 1E area monitors measure and transmit local radiation levels, and annunciate an alarm n a high radiation level. All monitors except the hydrogen recombiner control room have log display rate meters located adjacent to the detectors and are powered from normal AC er (Section 8.3.1.1.1). The hydrogen recombiner control room monitor is powered by inverted mal DC power. Their purpose is to protect plant personnel from excessive exposure and to vide a display radiation levels within the plant.

.4.5 Airborne Radioactivity Monitoring process and effluent radiation monitoring system described in Section 11.5 includes normal ge particulate and gas monitors. Their purpose is to monitor airborne radioactivity in areas that be occupied by plant personnel and to facilitate finding radioactive leaks. These monitors ple air from the reactor containment, the ESF building, the control room, and from locations he reactor plant heating and ventilation system upstream of the ventilation vent monitor. They capable of detecting airborne activated corrosion products and fission products at levels below derived air concentration of 10 CFR 20.

h the particulate and gas detector channels of these monitors are provided with an alert level m, in addition to the high alarm, with setpoints established by operating personnel to allow ervation of increases in airborne radioactivity. These monitors are polled every several onds by the radiation monitoring computer system. Once elevated readings are noticed, onnel with portable air samplers can determine which area, associated with the ventilation am with high radioactivity, contains the source of the problem.

3.5 REFERENCES

FOR SECTION 12.3

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

12.3-22 Rev. 30

Adsorption Units of Light-Water-Cooled Nuclear Power Plants, Revision 2.

-3 Regulatory Guide 1.69, USNRC. Concrete Radiation Shields for Nuclear Power Plants.

-4 Regulatory Guide 1.70, USNRC. Standard Format and Contents of Safety Analysis Reports for Nuclear Power Plants, Revision 3.

-5 Regulatory Guide 1.97, USNRC. Instrumentation for Light-Water-Cooled Nuclear Power Plant to Assess Plant and Environs Conditions During and Following Accident, Revision

2. (Compliance provided in a separate report, Section 1.7.4.)

-6 Regulatory 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.

-7 Stone & Webster Engineering Corporation (SWEC) 1975. Radiation Shielding Design and Analysis Approach for Light Water Reactor Power Plants, RP-8A, May 1975.

12.3-23 Rev. 30

TABLE 12.3-1 RADIATION ZONES Zone Designation Zone Description Maximum Allowable Dose Rate* (mRem/h I Unrestricted area - Continuous access 0.25 II Unrestricted area - Periodic access - 40 hrs/wk 2.5 III Restricted area - Controlled periodic access - 6 hrs/wk 15 IV Restricted area - Controlled infrequent access - 1 hr/wk 100 V High radiation area - Not normally accessible > 100 NOTE:

• Based upon the 5 rem per year criterion given in 10 CFR 20 and the maximum personnel occupancy time corresponding to e radiation zone.

Page 1 of 1 Rev

TABLE 12.3-2 RADIATION MONITORING SYSTEM - AREA RADIATION DETECTOR LOCATION Range ark Number Name Location mr/hr MS-RE01 Refueling Machine Containment - 51 feet 4 inches 1-105 MS-RE02 Fuel Transfer Tube Containment - 3 feet 8 inches 1-105 MS-RE03 Incore Inst. Transfer Area Containment - 24 feet 6 inches 1-105 MS*RE04A Containment High Range Containment - 52 feet 4 inches 103-1011 Internal MS*RE05A Containment High Range Containment - 51 feet 4 inches 103-1011 Internal MS-RE06 Decontamination Area Fuel - 24 feet 6 inches 0.1-104 MS-RE07 HVAC Area Auxiliary - 66 feet 6 inches 0.1-104 MS-RE08 Spent Fuel Pit Bridge/Hoist Fuel - 52 feet 4 inches 0.1-104 MS-RE09 Auxiliary Bldg General (A) Auxiliary - 18 feet 6 inches 0.1-104 MS-RE10 Auxiliary Bldg General (B) Auxiliary - 4 feet 6 inches 0.1-104 MS-RE11 Auxiliary Bldg General (C) Auxiliary - 4 feet 6 inches 0.1-104 MS-RE12 Auxiliary Bldg General (D) Auxiliary - 24 feet 6 inches 0.1-104 MS-RE13 Auxiliary Bldg General (E) Auxiliary - 24 feet 6 inches 0.1-104 MS-RE14 Auxiliary Bldg General (F) Auxiliary - 24 feet 6 inches 0.1-104 MS-RE15 Auxiliary Bldg General (G) Auxiliary - 43 feet 6 inches 0.1-104 MS-RE16 Volume Control Tank Auxiliary - 43 feet 6 inches 0.1-104 Cubicle MS-RE17 Waste Disposal Bldg (A) Waste Disposal - 4 feet 6 inches 0.1-104 MS-RE18 Waste Disposal Bldg (B) Waste Disposal - 4 feet 6 inches 0.1-104 MS-RE19 Solid Waste Storage Area Waste Disposal -24 feet 6 inches 0.1-104 MS-RE20 Sample Room Auxiliary - 43 feet 6 inches 0.1-104 MS-RE21 Laboratory Service - 24 feet 6 inches 0.01-103 MS-RE22 Control Room Control - 47 feet 6 inches 0.01-103 Page 1 of 2 Rev. 30

Range ark Number Name Location mr/hr MS-RE24 Waste Disposal Bldg (C) Waste Disposal - 4 feet 6 inches 0.1-104 MS-RE25 Waste Disposal Bldg (D) Waste Disposal - 4 feet 6 inches 0.1-104 MS-RE28 Fuel Building Pipe Rack Fuel - 11 feet 0 inches 0.1-104 MS-RE29 Spent Fuel Cask Area Fuel - 52 feet 4 inches 0.1-104 MS-RE31 Fuel Transfer Tube Containment - 24 feet 6 inches 0.1-104 MS-RE32 Containment Sump Area Containment - (-24 feet 6 inches) 0.1-104 MS-RE33 RHR Cubicle A Normal ESF - 4 feet 6 inches 0.1-104 Range MS-RE34 RHR Cubicle B Normal ESF - 4 feet 6 inches 0.1-104 Range MS-RE35 Incore Inst. Thimble Area Containment - 3 feet 8 inches 0.1-104 MS-RE36 Fuel Pool Monitor Fuel - 52 feet 4 inches 0.1-104 MS-RE37 Condensate Demineralizer Cond. Polishing - 14 feet 6 inches 0.01-103 MS-RE38 Regeneration Area Cond. Polishing - 38 feet 6 inches 0.01-103 MS*RE41 Fuel Drop Monitor Containment - 51 feet 4 inches 101-108 MS*RE42 Fuel Drop Monitor Containment- 51 feet 4 inches 101-108 MS-RE52 Recombiner Control Room HRB- 24 feet 6 inches 101-108 Page 2 of 2 Rev. 30

TABLE 12.3-3 OPERATOR ACTIVITY LOCATIONS AND TIME DURATIONS Approx. Duration Activity Location (minutes) Travel Rout 1 Locally trip the reactor trip breakers and bypass 43 feet 6 inches Auxilary <5 9 breakers Building (MCC Rod Control) 2 Deleted 3 Local actions needed to realign Spent Fuel Pool Spent Fuel Building < 15 5 Cooling, RBCCW and Service Water for spent fuel pool cooling 4 Powering the Plant Process Computer 38 feet Turbine Building < 10 6 5 Powering the SI accumulator valves 24 feet 6 inches Auxilary <5 9 Building 6 Initiate hydrogen monitor HRB < 30 3 7 Deleted 8 Deleted 9 Deleted 10 Monitor and maintain the porous concrete groundwater removal system

a. Repower sump pump Outside of ESF Building < 90 2 ESF Building 38 feet 6 inches < 10 2

b. Monitor sump pump and operate system Outside of ESF Building. <2 2

c. Maintain/repair sump pump ESF Building roof < 240 3 Page 1 of 2 Rev

Approx. Duration Activity Location (minutes) Travel Rout 11 Open the breakers for the non-safety grade sump 21 foot ESF Building < 15 2 pumps in the ESF and Auxiliary buildings 24 feet 6 inches Auxilary < 15 1 Building 4 foot Service Building < 15 12 Deleted 13 Align Alternate AFW Pump Suction Source or 21 foot ESF Building < 60 8 Replenish Demineralized Water Storage Tank (DWST) Inventory.

14 Reset MCC breakers for Diesel Generator keep Emergency Diesel Generator N/A ** 7 warm systems Building

• Figure 12.3-10 graphically depicts each route by route number.

• • There are no appreciable dose rates in the Emergency Diesel Generator Building.

Page 2 of 2 Rev

TABLE 12.3-4 ACTIVITY INITIATION TIME Time Frame

• Activity o 30 minutes 1 Locally trip the reactor trip breakers and bypass breakers 2 Deleted 4 Powering the process computer 6 Initiate hydrogen monitor 14 Reset MCC breakers for Diesel Generator keep warm systems minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5 Powering the SI accumulator valves 7 Deleted 9 Deleted 12 Deleted ours to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3 Local actions needed to realign Spent Fuel Pool Cooling, RBCCW and Service Water for spent fuel pool cooling 10 Repower porous concrete groundwater pump **

ours to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 10 Monitor porous concrete groundwater system 13 Align Alternate AFW Pump Suction Source or Replenish Demineralized Water Storage Tank (DWST) Inventory ay to 4 days 10 Monitor and maintain the porous concrete groundwater system 11 Open the breakers for the non-safety grade sump pumps in the ESF and Auxiliary Building ays to 30 days 8 Deleted 10 Monitor and maintain the porous concrete groundwater system 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 />.

Page 1 of 1 Rev. 30

s section was applicable during the prestart-up period as it provided estimates of the upational radiological consequences of future operation and the dose to workers during unit struction. Now that Millstone Unit 3 is operational, Annual Reports submitted to the NRC per ulatory Guide 1.16 should be consulted for data on the station occupational person-rem uirements. Information on design dose rates in unrestricted areas has been moved to Section

-1.

12.4-1 Rev. 30

regulatory guides and other references cited in this section were used as basis documentation the development of the radiation protection program. Documented methods and solutions erent from those set out in the guidance have also been incorporated in the radiation protection gram.

.1 ORGANIZATION radiation protection program is established to provide an effective means of radiation ection for permanent and temporary employees and for visitors at the station. To provide an ctive means of radiation protection, the radiation protection program incorporates a osophy from management (Section 12.1.1); employs qualified personnel to supervise and lement the program; provides appropriate equipment and facilities; and utilizes written cedures designed to provide protection of station personnel against exposure to radiation and oactive materials in a manner consistent with Federal and State regulations (Section 13.5).

radiation protection program is developed and implemented through the applicable guidance NPO 05-008, Regulatory Guides 8.2, Revision 1; 8.8, Revision 3; and 8.10, Revision 1-R.

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

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

radiation protection department coordinates with all station, corporate, and contractor anizations to provide radiation protection coverage for all activities that involve radiation or oactive material. The radiation protection department is organized to provide the following ices:

1. preparation and implementation of radiation protection procedures for routine and nonroutine activities associated with the operation, maintenance, inspection, and testing at the station;

2. compliance with regulatory requirements for maximum permissible dose limits and contamination control;

3. maintenance of a personnel radiation dosimetry program and dosimetry records;

4. the surveying of station areas, maintenance of survey records, and the posting of survey results for daily activities within the station; 12.5-1 Rev. 30

6. procurement, maintenance, and calibration 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 receiving 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; and

11. preparation, maintenance, and implementation of the radiological respiratory protection program.

12. Ensure stop work authority when required by actual or potential radiological conditions.

chemistry department is responsible for measuring the radioactive content of all gaseous and id effluents from the site in accordance with the requirements of the Technical Specifications, iological Effluent Monitoring and Offsite Dose Calculation Manual and 10 CFR 20.

a policy of the Millstone Power Station to keep personnel radiation exposure within the licable regulations, and beyond that, to keep it as low as reasonably achievable.

.2 EQUIPMENT, INSTRUMENTATION, FACILITIES criteria for purchasing the various types of portable and laboratory equipment used in the ation protection and chemistry department is based on several factors. Portable survey and ratory radiation detection equipment is selected to provide the appropriate detection abilities, ranges, accuracy and durability required for the expected types and levels of ation anticipated during normal operating or emergency conditions. Selection of respiratory ection equipment such as full-face masks, self-contained breathing apparatus, and respirator rs is made following the guidance of applicable approval regulations.

iation protection equipment, such as portable survey meters, is maintained by radiation ection. Survey equipment for use in emergency situations is stored in emergency kits which located in such areas as the control room and the emergency operation facility. Special able equipment, such as personnel air samplers, is available from radiation protection, and is zed at the discretion of radiation protection supervision. Respiratory protection equipment is marily stored at the respiratory storage and issue facilities.

12.5-2 Rev. 30

mon inventory of hand-held radiation meters, electronic dosimeters, and National Voluntary oratory Accreditation Program (NVLAP) accredited individual monitoring devices. The lstone Station radiation protection group maintains adequate supplies of hand-held radiation ers, secondary dosimeters, and NVLAP accredited dosimeters for normal station activities, tiple unit shutdowns, and/or potential accident conditions. The station will maintain an quate supply of portable radiation protection instrumentation strategically located at the lity to ensure the radiation protection staff is properly equipped to perform their required ctions. These instruments will be calibrated as specified by the manufacturers instructions and cedural requirements or as deemed necessary by radiation protection supervision. Calibration, ration, and maintenance procedures are followed for each specific type of instrument.

ailed records of calibration and maintenance of each instrument are maintained at the station.

ibrations are performed using radiation sources of known activity. These sources are calibrated ertified accurate by the National Institute of Standards and Technology (NIST). Calibration rces are stored by radiation protection. Actual calibration of equipment is performed in the bration laboratories or other appropriate facilities.

radiation protection group and chemistry group maintain appropriate laboratory instruments erform the required radiological evaluations to support the station needs. Radiation protection hemistry personnel check each counting system at regular intervals with standard radioactive rces to determine counting efficiencies, proper voltage settings, and background count rates.

ords are maintained for each instrument or counting system. Repair and maintenance of ratory equipment is performed by station personnel or through vendor repair contracts.

Millstone site contains the following areas:

estricted Area - access to which is neither limited nor controlled by the licensee.

trolled Area - an area, outside a restricted area but inside the site boundary, access to which be limited by the licensee for any reason.

tricted Area - an area, access to which is limited by the licensee for the purpose of protecting viduals against undue risks from exposure to radiation and radioactive materials.

iologically Controlled Area/Radiological Control Area (RCA) - an area, posted with a sign by licensee for the purpose of protecting individuals from exposure to radiation and/or oactive materials. Dosimetry is always required within the RCA.

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

lstone, with three units, has a number of RCAs within the Restricted Area.

hin the RCA, Radiation Areas, High Radiation Areas, Technical Specification Locked High iation Areas, Very High Radiation Areas, Contaminated Areas, Airborne Radioactivity Areas, ioactive Materials Areas, and Hot Particle Areas can be found.

12.5-3 Rev. 30

discretion of radiation protection supervision, a personnel monitor or frisker is placed in cific areas at the station where contamination or the potential for contamination may be ent.

areas where radioactive materials and radiation may result in doses in excess of the dose ts in 10 CFR 20, Section 20.1301 is surveyed, classified, and conspicuously posted with the ropriate radiation caution signs, labels, and signals in accordance with 10 CFR 20, Sections 902 and 20.1903, except as described below.

station employs administrative and physical security measures to prevent unauthorized entry ersonnel into any high or very high radiation areas. The NRC granted the station approval in ordance with 10 CFR 20.1601(c), to use alternative methods for controlling access to high ation areas in place of the controls required by 10 CFR 20.1601 (a) and (b). These alternative hods are described in the administrative section of each units Technical Specifications.

y High Radiation Areas are those areas where an individual could receive in excess of 500 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 etrates. These areas, in addition to the controls specified in the Technical Specifications, have ique key and a specific procedure for entry into the area.

ffic patterns normally discourage or prevent access to radiation or potential radiation areas.

ning signs, audible and visible indicators, barricades, or locked doors are employed to protect onnel from access to high radiation areas that may exist temporarily or semi-permanently as a lt of unit operations and maintenance.

iation protection services and facilities around the site provide all workers the necessary ection and controls for work in radioactive environments.

ministrative radiation protection activities are centered around the radiation protection office.

ndard office equipment, equipment storage areas, records storage, and some personnel imetry equipment are among the items to be included in the radiation protection office.

sonnel decontamination supplies and equipment are stored in the radiation protection ontamination facility. This room contains stainless steel showers and sinks, with drains cted to the wastewater treatment system (Section 9.2.3). A low-background count laboratory sed for counting and/or identifying radioactivity in airborne and liquid samples in formance with 10 CFR 20, and to 10 CFR 50 App. A General Design Criterion 64. The mistry laboratory is used to perform chemical and elemental analyses of environmental uents. All sink and floor drains in this room are directed to the wastewater treatment system; e hood exhaust is directed to the ventilation system. Equipment used to perform routine nting and analyses on all plant radioactivity samples, as required by 10 CFR 20, are acquired, ntained and calibrated as appropriate.

personnel entering contaminated areas are required to wear protective clothing. The nature of work to be done, the contamination level in the area, and the total industrial risks, are the 12.5-4 Rev. 30

uired, to ensure efficient operations and to preclude the spread of contamination. Protective hing available at the station includes the following:

modesty garments shoe covers overshoe rubbers head covers gloves coveralls and lab coats itional items of specialized apparel are available for operations involving high-level tamination, such as:

plastic or rubber suits surgeons masks face shields bubble hoods protective clothing is cleaned and decontaminated at a vendor laundry, on-site laundry facility, isposed of as radioactive waste.

ropriate training and written guidance govern the proper use of protective clothing, where and it is to be worn and removed, the decontamination facilities for personnel and equipment, the areas to be used.

piratory protective equipment is available to qualified station personnel and issued to viduals, as required by actual or potential occupational risk of the work assignment. The iratory protection program follows the guidance of Regulatory Guide 8.15, Revision 1, and plies with 10 CFR 20, Subpart H. Respiratory protection equipment is stored at the iratory storage and issue facilities. Respiratory equipment may include:

pressure demand full-face-piece air line respirators; continuous air flow hoods or suits; pressure demand full-face-piece self-contained breathing apparatus; and 12.5-5 Rev. 30

piratory protective equipment is cleaned, sanitized, repaired and decontaminated at respiratory age and issue facilities, or at vendor facilities.

radiation workers are issued NVLAP accredited dosimeters and are required to wear such imeters at all times while within any RCA. All other individuals who enter an RCA are uired to wear an individual radiation monitoring device.

ctronic dosimeters or direct-reading pocket dosimeters are issued as an additional method for rmining gamma exposure. All individuals are required to examine their dosimeters at frequent rvals while in radiation areas. The use, care, and testing of these direct reading dosimeters will ow applicable guidance of Regulatory Guides 8.4, Revision 1, and 8.28, Revision 0.

cial or additional dosimetry, such as finger ring dosimeters and teledosimetry, are issued under cial conditions at the discretion of radiation protection supervision.

NVLAP accredited individual monitoring devices are processed periodically at the discretion adiation protection personnel. In addition, they can be processed promptly whenever it ears that an overexposure may have occurred.

imeter records furnish the exposure data for the administrative control of radiation exposure.

osure records for each individual are maintained in accordance with the guidance of ulatory Guide 8.7, Revision 2.

.3 PROCEDURES radiation protection procedures and methods of operation for ensuring that occupational ation exposure is as low as reasonably achievable (ALARA) follow the provisions and gestions of Regulatory Guides 8.8, Revision 3; 8.10, Revision 1-R; and 1.33, Revision 2, as licable. Such procedures are implemented by qualified personnel whose qualifications meet requirements of Regulatory Guide 1.8, Revision 3. In addition, all administrative and cedural practices associated with the monitoring of occupational radiation exposure follow the dance of Regulatory Guides 8.2, Revision 1; 8.4, Revision 1; 8.7, Revision 2; 8.9, Revision 1; 8.34, Revision 0.

ny radiation protection procedures at Millstone Nuclear Power Station are common to Units 2

3. Radiation protection procedures are an integral part of the ALARA program at the station.

ess to restricted/radiologically controlled areas is controlled by administrative and physical urity measures as required by 10 CFR 20, Subparts G and J.

ion management assures entry control to high radiation areas through the administration of ation work permits (RWPs) that stipulate purpose of entry, work location, radiological ditions, surveillance and dosimetry requirements, stay time, protective clothing, respiratory 12.5-6 Rev. 30

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 sources (station systems and components), job types and functions, and personnel categories where exposures occur.

Ps are issued for general and specific activities performed in radiation areas, contaminated s, airborne radioactivity areas, and for all activities that require entrance into high radiation s, and very high radiation areas as defined in 10 CFR 20, Section 20.1003. RWPs are also ed prior to maintenance or inspection of contaminated or radioactive equipment with ovable contamination in excess of 1,000 dpm/100 cm2 beta-gamma and/or 100 dpm/100 cm2

a. RWPs are also required prior to entrance into the reactor containment of any unit.

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

iation protection personnel routinely survey selected areas of the station to assess and control osure to radiation and radioactive materials in accordance with 10 CFR 20, Section 20.1501.

ending on the type of survey required and anticipated types and levels of radioactivity, ous portable instruments and techniques are used to perform these surveys. Results of all eys are recorded and kept on file at the radiation protection office on a short term basis.If essary, survey sheets may be posted. Permanent storage is provided by forwarding records to nuclear records facility. Reporting practices for all normal and accident conditions comply h the regulations set forth in 10 CFR 20, Subpart M.

a surveys are performed at scheduled frequencies, based on location, radiation levels, station us, and occupancy. All area survey readings are recorded and filed as required by 10 CFR 20, tion 20.2103 and Regulatory Guide 8.2, Revision 0. Caution placards, describing the ological conditions, are posted to comply with the requirements of 10 CFR 20, Section 902.

12.5-7 Rev. 30

ion, where the potential for the spread of contamination exists. Contamination surveys are e using the smear or swipe technique, or by using an appropriate portable instrument.

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

tamination surveys are performed on personnel, equipment, and in uncontrolled areas to ure that radiological control methods are adequate. Personnel, equipment, and material leaving taminated areas are monitored to prevent the spread of contamination into clean areas. Areas, ipment, and personnel that may be contaminated with radioactive material are decontaminated g applicable methods and techniques, such as those suggested in NCRP65 and IE Circular 07.

els of contamination are also used to judge the potential for airborne radioactive material and need for monitoring air, and the use of engineering controls or respiratory protection.

managements intent to control airborne radioactivity levels as effectively as practicable by per preventive measures, engineering controls, and good housekeeping techniques. In the nt of a radioactive airborne problem, every effort is made to promptly assess the situation.

tion 12.3.4 provides information on the installed airborne radioactivity monitoring rumentation.

trol of airborne radioactivity levels is assured through the use of the stations heating, tilation, and air-conditioning (HVAC) systems and portable air movers and filters. The HVAC ems provide controlled air movement and filtration capability for those areas with a high ntial for airborne radioactivity problems. As required, special control techniques are used to imize airborne exposure arising from special work projects. Respiratory protection equipment vailable for use in those situations where airborne radioactivity hazards exist and where other trol measures are inadequate at the location and time. Respiratory protection equipment use is ssed based upon the principle of keeping the Total Effective Dose Equivalent ALARA, sistent with minimizing total occupational risk.

special control techniques used to minimize airborne exposure include decontamination of component or area prior to performing work, keeping work surfaces damp while work is in gress, and using tents or glove bags in conjunction with appropriate, filtered ventilation ems.

hniques for obtaining breathing zone air samples are grab samples taken in areas esentative of the workers breathing zone and/or lapel air samplers.

e conditions which require special air sampling include lifting the reactor vessel head, ting a contaminated system, and working on an open contaminated system.

egard to reporting practices for airborne contamination surveys, radiation protection ervision is notified when airborne concentrations read 30 percent of DAC and the area uires posting if this condition persists for a sustained period.

12.5-8 Rev. 30

air sampling program provides information on the potential inhalation of radioactive material workers. The information is used to determine what remedial action or protective measures h as respirators, glove boxes, or engineering controls are necessary to protect the worker. Air ples are taken for all work on systems which have the potential for release of airborne oactivity. Surveys are performed on a routine basis, depending on location, station status, and upancy. In addition, surveys are performed whenever work is required on a known or ntially contaminated system that must be opened to the working environment or whenever ding, burning, or grinding is performed on a known or potentially contaminated system.

veys are also performed whenever the continuous air monitor indicates an airborne problem prior to containment entry. Additional surveys are performed as deemed necessary by ation protection supervision.

r to issuance and use of required respiratory protection equipment, each individual must have sfactorily 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 air sampling and respiratory protection programs meet the recommendations and provisions 0 CFR 20, Subpart H, Regulatory Guide 8.15, Revision 1, and NUREG-0041.

cial procedures control the handling or movement of material within and from restricted and ologically controlled areas, such as the shipment and receipt of radioactive materials. These cedures comply with the regulations stipulated in 49 CFR 170-178, 10 CFR 70, 10 CFR 71, 10 CFR 20.1906.

previously discussed in this section, all radiation workers receive a NVLAP accredited vidual monitoring device and direct-reading pocket ion chamber and/or electronic dosimetry onitor personnel exposure. Exposure records are filed and retained for each individual in ordance with the recommendations of Regulatory Guides 8.2, Revision 1 and 8.7, Revision 2, as required by 10 CFR 20, Subpart L. Any reports of overexposures and excessive levels and centrations comply with the regulations of 10 CFR 20, Subpart M. Reports of personnel nitoring, and reports of theft or loss of licensed material are issued in accordance with the ulations required by 10 CFR 20, Subpart M.

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

bioassay program includes:

12.5-9 Rev. 30

selection of measurement techniques, measurement frequency, and program participants; action points and actions to be taken based on measurement results; and interpretation of measurement results in terms of location of radioactive material in the body, the quantity present, the rate of elimination, and the resulting dose commitment and use of personnel contamination monitors, located at RCA exits and the Protected Areas exits, which may serve as passive monitors for detection of internal contamination in lieu of periodic whole body counts for all workers.

hole-body counter is located at the station as needed for in vivo measurement of station onnel, visitors, or support personnel. The whole-body counter provides preliminary kground information, periodic evaluation, and emergency capability for detecting internal osure conditions. Assessment of internal radiation exposure for station personnel may be ormed, for example when:

individuals have a known or suspected intake of four or more DAC hours within a calendar week, incidents involve contamination around the nose or mouth; and accidents involve a potential intake. Excreta samples from suspected individuals may be sent to a qualified laboratory for analysis.

ning in radiation protection principles and procedures is performed by the Nuclear Training artment or by qualified station personnel. New employees, contractors, and other supporting onnel receive validation of prior training and orientation training, as appropriate, before the inning of their work assignments.

permanent station personnel who are required to work in the RCA are required to successfully plete basic training courses and practical exercises to demonstrate their proficiency and petence.

radiological workers participate in the radiological worker training program. The radiological ker training program maintains the proficiency of employees through training and annual aining on selected material. Additional training is given for selected tasks which require eased radiation protection.

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

12.5-10 Rev. 30

ition to the procedures specifically required by Regulatory Guide 1.33, Revision 2.

.4 REFERENCE FOR SECTION 12.5

-1 Nuclear Regulatory Commission, Code of Federal Regulations 10 CFR Part 20, Standards for Protection Against Radiation 5-2 Nuclear 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

-3 Nuclear Regulatory Commission, Code of Federal Regulations 10 CFR Part 70, Domestic Licensing of Special Nuclear Material

-4 Nuclear Regulatory Commission, Code of Federal Regulations 10 CFR Part 71, Packaging and Transportation of Radioactive Material

-5 Department of Transportation, Code of Federal Regulations 49 CFR 170-178, Subchapter C-Hazardous Materials Regulations 5-6 NUREG-0731, Guidelines for Utility Management Structure and Technical Resources, 1980

-7 NUREG-0800, USNRC. Standard Review Plan, Revision 1.

-8 Regulatory Guide 1.8, Rev. 3 Qualification and Training of Personnel for Nuclear Power Plants

-9 Regulatory Guide 1.33, Rev. 2, Quality Assurance Program Requirements 5-10 Regulatory Guide 8.2, Rev. 1, Guide for Administrative Practices in Radiation Monitoring

-11 Regulatory Guide 8.4, Rev. 1, Personnel Monitoring Device - Direct-Reading Pocket Dosimeters

-12 Regulatory Guide 8.7, Rev. 2, Instructions for Recording and Reporting Occupational Radiation Exposure Data

-13 Regulatory Guide 8.8, Rev. 3, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as is Reasonable Achievable 12.5-11 Rev. 30

-15 Regulatory Guide 8.10, Rev. 1-R, Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As is Reasonably Achievable

-16 Regulatory Guide 8.13, Rev. 3. Instruction Concerning Prenatal Radiation Exposure

-17 Regulatory Guide 8.15, Rev. 1, Acceptable Programs for Respiratory Protection

-18 Regulatory Guide 8.27. Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants, Revision 0.

-19 Regulatory Guide 8.28, Rev. 0, Audible-Alarm Dosimeters

-20 Regulatory Guide 8.29. Instructions Concerning Risks from Occupational Radiation Exposure, Revision 1.

-21 Regulatory Guide 8.34, Rev. 0, Monitoring Criteria and Methods To Calculate Occupational Radiation Doses

-22 Regulatory Guide 1.33. Quality Assurance Program Requirements, Revision 2.

-23 INPO 05-008, Guidelines for Radiological Protection at Nuclear Power Stations

-24 IE Circular 81-007, Control of Radioactively Contaminated Material

-25 NCRP Number 65, Management of Persons Accidently Contaminated with Radionuclides

-26 NUREG-0041, Rev. 1, Manual of Respiratory Protection Against Airborne Radioactive Material 12.5-12 Rev. 30

Page 1 of 1 Rev. 30 Page 1 of 1 Rev. 30 MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.2 - 1 ARRANGEMENT - OPERATING PERSONNEL ACCESS AND EGRESS (PLAN ELEVATION 3 FEET 8 INCHES AND ABOVE)

Withheld under 10 CFR 2.390 (d)(1)

Amendment 10 October 1984 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.2-2 ARRANGEMENT - OPERATING PERSONNEL ACCESS AND EGRESS (PLAN ELEVATION 24 FEET 6 INCHES AND ABOVE)

Withheld under 10 CFR 2.390 (d)(1)

August 1999 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.2-3 ARRANGEMENT - OPERATING PERSONNEL ACCESS AND EGRESS (PLAN ELEVATION 38 FEET 6 INCHES AND ABOVE) tIIQI_-'~~ ~

_lJIC.--_IIY 1t1IIIC7_~ ~

_ _N

_ _"-_lUW ....,..

-_f\~ ~

_ .._ . _ .. ~~-_.

-~.....,

e-_ _

. o 60 t20

__ ~

~

. . iltlDi_CC4 SCaLIE-FEET

_ _ r!llliill _ _ ~" __

---.-&_.. - .. ~

AGlJRE 12.2-3 ARRANGEMENT-OPERATING PERSONNEL ACCESS AND EGRESS MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT

!!!!!!::::::!!~~!!!I!l:g~::b==!!=~~"e>CAD FILE: 1223.dgn/1223.cit DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR FIGURE 12.2-4 I-131 CONCENTRATION CONTAINMENT (HISTORICAL)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-1 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

ZONE DOSE RATE ZONE (MREM/HR) CODE I S025 o

II S2.5 m: ~15

~

L2d

~IOO o

>100 L22Th1 FIGURE 12.3-1 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT CAD FILE: 1231.dgn/1231.cl"t DECEMBER 1~7 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-2 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

ZONE DOSE RATE ZONE (MREM/HR) CODE I ~O.25 o

II ~2.5 m ~15 ... ~ .

~

~tOO

>100 NOTE:

DURING RESIN AND EVAPORATOR BOTTOMS TRANSFER OPERATIONS, A MAJOR PORTION OF THE SOLID WASTE DISPOSAL BUILDING WILL BE RADIATION ZONE Y. AT OTHER TIMES THE RADIATION LEVELS WILL BE DETERMINED BY SURVEY.

COHOS POL DEMIN E. FIGURE 12.3-2 REGEN EOPT DESIGN BASIS RADIATION ENCLOSURE UNIT 3 ZONES FOR SHIELDING (NORMAL OPERATIONS)

MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT CAD FILE: 1232.dgn/1232.cit DECEMBER 19'37 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-3 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

ZONE DOSE RATE ZONE (MREM/HR) CODE I SQ25 D

2.5 JlI ~15 rrrmmm lliillllillil

>100 FIGURE 12.:3-:3 DESIGN BASIS RADIATiON ZONES FOR SHIELDlNG (NORMAL OPERATJONS)

MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT CAD FILE: 1233.d<Jn/1233.clt DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-4 DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

ZONE

>100 m

IN T FIGURE 12.3-4 OF DESIGN BASIS RADIATION ZONES FOR SHIELDING (NORMAL OPERATIONS)

MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT CAD FILE: 1234.dgn/1234.clt DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR FIGURE 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 SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-6 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN/REFUELING)

ZONE DOSE RATE ZONE (MREM/HR) CODE I ~0.25 0

rr ~2.5 III ~15

~

$100 ': ' .. :: ....::

0

>100

~

NOTE:

DURING INITIAL COLD SHUTDOWN, THE DOSE RATE IN THE RHR SYSTEM HEAT EXCHANGER CUBICLES MAY EXCEED THE RADIATION ZONE m CRITERiON AND, IN THE ADJACENT AREAS, MAY EXCEED THE RADIATION ZONE I CRITERION UNTIL THE REACTOR COOLANT IS DILUTED BY RWST WATER.

FIGURE 12.3-6 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN/ REFUEL1NG)

MILLSTONE NUCLEAR POWER STATtON UNIT 3 FINAL SAFETY ANALYSIS REPORT CAD FILE: 1236.dgn/1236.clt DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-7 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN/REFUELING)

ZONE DOSE RATE ZONE (MREM/HR) CODE I ~Q25 D

.n: ~2.5 t .. *********

m ~15 .............

~

~IOO

>100 (f£f~~i~j~~~

NOTES:

I. DURING INITIAL COLD SHUTDOWN. THE DOSE RATE IN THE RHR SYSTEM HEAT EXCHANGER CUBICLES MAY EXCEED THE RADIATION ZONE m CRITERION AND, IN THE ADJACENT AREAS. MAY EXCEED THE RADIATION ZONE I CRITERION UNTIL THE REACTOR COOLANT IS DILUTED BY RWST WATER.

2. DURING RESIN AND EVAPORATOR BOTTOMS TRANSFER OPERATIONS. A MAJOR PORTION OF THE SOLID WASTE OISPOSAl BUILDING WILL BE RADIATION ZONE Y. AT OTHER TIMES THE RADIATION LEVELS WILL BE DETERMINED BY SURVEY.

NOS PO\..

EMIN fA FIGURE 12.3-7 GEN EQPT DESIGN BASIS RADIATION CLOSURE NIT 3 ZONES FOR SHIELDING (SHUTDOWN I REFUELING)

MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT CAD FILE: 1237.dgn/1237.cit DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-8 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN/REFUELING)

ZONE DOSE RATE ZONE (MREM/HR) CODE I ~025 o

II: S2.5 m: ~15

~

~

SIOO ~

L:.:.:2J

>100

~

NOTE:

CURING INITIAL COLD SHUTDOWN, THE DOSE RATE IN THE RHA SYSTEM HEAT EXCHANGER CUBICLES MAY EXCEED THE RADIATiON ZONE l'l1 CRITERION AND, IN THE ADJACENT AREAS, MAY EXCEED THE RADIATION ZONE I CRITERION UNTIL THE REACTOR COOLANT IS DILUTED BY RWST WATER.

• FlGURE 12.3-8 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN/ REFUELING)

MILLSTONE NUCLEAR POWER STATION UNIT 3 ANAL SAFETY ANALYSIS REPORT CAD FILE: 1238.dgn/1238.ci-t DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-9 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN/REFUELING)

ZONE DOSE RATE ZONE (MREM/HR) CODE I ~o.25 0

II ~2.5

• • • • • • • • • • 4 *

m: .s15 ...............

~-. -........

~tOO D

>100 BC2]

FIGURE 12.3-9 DESIGN BASIS RADIATION ZONES FOR SHIELDING (SHUTDOWN / REFUELI NG)

MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFEl1' ANALYSIS REPORT CAD FILE: 1239.dgn/1239.cit DECEMBER 1997 Withheld under 10 CFR 2.390 (d)(1)

December 1997 Rev. 20.3

SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 MPS-3 FSAR FIGURE 12.3-10 SHEET 1 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 1 TO AUXILIARY BUILDING Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 2 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 2 TO THE ESF BUILDING l _

Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 3 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 3 TO THE HYDROGEN RECOMBINER BUILDING Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 4 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 4 TO THE HYDROGEN RECOMBINER BUILDING AND TO UNIT 2 CHEMISTRY LABS Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 5 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 5 TO FUEL BUILDING Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 6 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 6 TO TURBINE BUILDING Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 7 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 7 TO DIESEL GENERATOR Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 8 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 8 TO ESF BUILDING Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR SECURITY-RELATED-INFORMATIONWithhold under 10 CFR 2.390 FIGURE 12.3-10 SHEET 9 ROUTES TO POST-ACCIDENT VITAL AREAS - ROUTE 9 TO AUXILIARY BUILDING Withheld under 10 CFR 2.390 (d)(1)

Rev. 21.3

MPS-3 FSAR FIGURE 12.3 - 11 FUEL TRANSFER TUBE SHIELDING Amendment 6 January 1984 Rev. 20.3

MPS-3 FSAR FIGURE 12.3 - 12 UPPER REACTOR CAVITY NEUTRON SHIELD Amendment 8 May 1984 Rev. 20.3