Text

6 to Updated Final Safety Analysis Report, Chapter 12, Radiation Protection
	ML20309A598
Person / Time
Site:	North Anna
Issue date:	09/30/2020
From:	; Virginia Electric & Power Co (VEPCO)
To:	; Office of Nuclear Reactor Regulation
References
	20-324
	Download: ML20309A598 (104)
	v • d • e

North Anna Power Station Updated Final Safety Analysis Report Chapter 12

Intentionally Blank Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-i Chapter 12: Radiation Protection Table of Contents Section Title Page 12.1 SHIELDING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-1 12.1.1 Design Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-1 12.1.2 Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-2 12.1.2.1 Primary Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-2 12.1.2.2 Secondary Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-3 12.1.2.3 Reactor Coolant Loop Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-4 12.1.2.4 Containment Structure Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-4 12.1.2.5 Fuel-Handling Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-4 12.1.2.6 Auxiliary Equipment Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-5 12.1.2.7 Waste Storage Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-5 12.1.2.8 Accident Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-6 12.1.2.9 Boron Recovery Tank Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-6 12.1.2.10 Main Control Room Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-6 12.1.2.11 Shielding Review for NUREG-0578 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-7 12.1.3 Source Terms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-7 12.1.4 Area Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-9 12.1.4.1 Normal Plant Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-9 12.1.4.2 Post-Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-10 12.1.5 Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-11 12.1.6 Dose Rate Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-11 12.1.6.1 Sample Sink Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-11 12.1.6.2 Valve-Operating Area Outside Demineralizer Cubicle. . . . . . . . . . . . . . . . . . 12.1-12 12.1.6.3 GAMTRAN Computer Code. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-13 12.1.7 Estimates of Exposure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-13 12.1.7.1 Considerations for Dose Predictions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-13 12.1.7.2 Reports From Other Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-15 12.1.7.3 Dose From Stored Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-16 12.1.7.4 Health Physics Area Dose Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-16 12.1 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-17 12.1 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-17 12.2 VENTILATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-1 12.2.1 Design Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-1 12.2.2 Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-1 12.2.2.1 Auxiliary Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-2

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-ii Chapter 12: Radiation Protection Table of Contents (continued)

Section Title Page 12.2.2.2 Containment Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-2 12.2.2.3 Turbine Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-3 12.2.2.4 Fuel Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-3 12.2.3 Source Terms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-3 12.2.4 Airborne Radioactivity Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-3 12.2.5 Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-6 12.2.5.1 Filter Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-6 12.2.5.2 Temporary Air Ducting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-6 12.2.6 Estimates of Inhalation Doses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-7 12.2 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-9 12.2 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-9 12.3 HEALTH PHYSICS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3-1 12.3.1 Program Objectives and Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3-1 12.3.2 Facilities and Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3-2 12.3.3 Personnel Dosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3-3 12.3 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3-3 12.4 RADIOACTIVE MATERIALS SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4-1 12.4.1 Materials Safety Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4-1 12.4.2 Facilities and Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4-2 12.4.3 Personnel and Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4-2 12.4.4 Required Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4-2 Appendix 12A Description of Neutron Supplementary Shield . . . . . . . . . . . . . . . . . . . . 12A-i 12A.1 INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-2 12A.2 NEUTRON SHIELD DESIGN CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-3 12A.3 EFFECTIVENESS OF THE SUPPLEMENTARY NEUTRON SHIELD . . . . . . 12A-3 12A.4 SHIELD DESIGN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-4 12A.4.1 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-4 12A.4.2 Location. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-5 12A.4.3 Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-5

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-iii Chapter 12: Radiation Protection Table of Contents (continued)

Section Title Page 12A.4.4 Supports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-6 12A.4.5 Missile Effects. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-6 12A.4.6 Effect on Containment Sump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-6 12A.5 REACTOR PRESSURE VESSEL SUPPORT INTEGRITY REVIEWS . . . . . . . 12A-6 12A References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-8

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-iv Chapter 12: Radiation Protection List of Tables Table Title Page Table 12.1-1 Radiation Zone Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-18 Table 12.1-2 Containment Shielding Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-19 Table 12.1-3 N-16 and Activated Corrosion Product Activity . . . . . . . . . . . . . . . . . . 12.1-21 Table 12.1-4 Area Radiation Monitoring Locations, Number and Ranges. . . . . . . . . 12.1-22 Table 12.1-5 Materials Used for Source and Dose Rate Calculations . . . . . . . . . . . . 12.1-23 Table 12.2-1 Equilibrium Activities in Different Plant Buildings (Ci/cm3) . . . . . . . . 12.2-10 Table 12.2-2 Estimate of Annual Inhalation Doses to Plant Personnela . . . . . . . . . . . 12.2-11 Table 12A-1 Comparison of Calculated Neutron Dose Rates with Measurements Made at North Anna Unit 1, Adjusted to 100% Power . . . . . . . . . . . . . . . . . . . . 12A-9 Table 12A-2 Calculated Neutron Dose Rates with Supplementary Neutron Shielding 12A-10 Table 12A-3 Reactor Pressure Vessel Support and Neutron Shield Tank Loads Phase12A-11 Table 12A-4 Reactor Pressure Vessel Nozzle Support Loads Phase, Including Reactor Pressure Vessel Internals Movement, Asymmetric Pressure, Deadweight, and Seismic . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-12 Table 12A-5 Relative Displacement Between Top and Bottom of Nozzle Support a 12A-13 Table 12A-6 Survey Results of Unit 1 Reactor Containment at the 291 ft. Elevation on 11/10/10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-14 Table 12A-7 Survey Results of Unit 2 Reactor Containment at the 291 ft. Elevation on 10/20/10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-15 Table 12A-8 Survey Results of Unit 1 Reactor Containment at the 291 ft. Elevation on 01/04/17 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-15 Table 12A-9 Survey Results of Unit 2 Reactor Containment at the 291 Ft. Elevation on 10/18/17 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-16

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-v Chapter 12: Radiation Protection List of Figures Figure Title Page Figure 12.1-1 Radiation Zones Containment Structure . . . . . . . . . . . . . . . . . . . . . . . 12.1-24 Figure 12.1-2 Radiation Zones Auxiliary Building . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-32 Figure 12.1-3 Radiation Zones Fuel Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-35 Figure 12.1-4 Radiation Zones Decontamination Building . . . . . . . . . . . . . . . . . . . . 12.1-37 Figure 12.1-5 Radiation Zones Waste Disposal Building . . . . . . . . . . . . . . . . . . . . . 12.1-39 Figure 12.1-6 Shield ArrangementPlan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-40 Figure 12.1-7 Permali Locations Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-41 Figure 12.1-8 Permali Locations Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-42 Figure 12.1-9 Shield Arrangement Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-43 Figure 12.1-10 Shield Arrangement Plan Operating Floor. . . . . . . . . . . . . . . . . . . . . . 12.1-44 Figure 12.1-11 Dose Rate Per Curie of Co-60 Equivalent vs. Distance from Low Level Contaminated Storage Area . . . . . . . . . 12.1-45 Figure 12A-1 Plan View of Operating Floor Showing Detector Locations . . . . . . . . 12A-17 Figure 12A-2 Collar Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-18 Figure 12A-3 Plan View of Unit 2 Containment for Survey Points. . . . . . . . . . . . . . 12A-19 Figure 12A-4 Shield Dust Cover Blocks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-20 Figure 12A-5 Crane Wall Openings With Permali Elevation 291 ft. 10 in. Unit 2 . . 12A-21 Figure 12A-6 Crane Wall Openings With Permali Elevation 291 ft. 10 in. Unit 1 . . 12A-22 Figure 12A-7 Location of Supplementary Neutron Shields . . . . . . . . . . . . . . . . . . . . 12A-23 Figure 12A-8 RPV Nozzle Support Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12A-24 Figure 12A-9 Plan View of Unit 1 Containment for Survey Points. . . . . . . . . . . . . . 12A-25 Figure 12A-10 Plan View of Unit 1 Containment for Survey Points. . . . . . . . . . . . . . 12A-26 Figure 12A-11 Plan View of Unit 2 Containment for Survey Points. . . . . . . . . . . . . . 12A-27

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-vi Intentionally Blank

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-1 CHAPTER 12 RADIATION PROTECTION At the North Anna Power Station, entrance to the station proper is controlled by station security. Inside the station proper, there is a protected area (inner barrier) consisting of fences and/or walls of structures. The containment building, turbine building, auxiliary building, service building, fuel building and other miscellaneous buildings are within the protected area. From a radiological access standpoint, the area within the protected area is the primary restricted area.

Other secondary restricted areas exist within the station proper but outside the protected area, such as the Old Steam Generator Storage Facility. Individuals entering restricted areas must have satisfactorily completed a basic Health Physics training course or possess the equivalent Health Physics knowledge, or be escorted by an individual who has those qualifications.

Within the restricted areas, Health Physics procedures are implemented as detailed in Sections 12.1.5 and 12.3. It is anticipated that, during normal station operation, areas outside the established restricted areas will not experience radiation levels sufficient to classify them as restricted areas in the context of 10 CFR 20. However, if such radiation levels were to occur, they would be detected by periodic radiation surveys and appropriate radiation protection measures would be established for such areas in accordance with Section 12.3.

The policy and objectives of VEPCO are to ensure that the exposure of personnel to radiation is maintained as low as is reasonably achievable (ALARA) at its nuclear power stations.

Maintaining individual exposure ALARA is a requirement of 10 CFR 20 and a management commitment. Management assumes the responsibility for ensuring the implementation of this policy by its incorporation into all aspects of station planning, design, construction, operation, maintenance, and decommissioning. This policy applies not only to controlling the maximum dose to individuals but also maintaining the collective dose to personnel, i.e., total man-rem exposure, as low as is reasonably achievable.

To attain the goal of this commitment, system, station, and contractual personnel shall integrate their efforts as necessary to perform their functions in such a manner that exposure(s) to radiation will be maintained ALARA. As applicable, new procedures shall be formulated while existing procedures and practices shall be reviewed and modified, if necessary, to ensure their conformance to the principle of maintaining exposures ALARA.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12-2 Intentionally Blank

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-1 12.1 SHIELDING 12.1.1 Design Objectives Radiation protection, including radiation shielding, is designed to ensure that the criteria specified in 10 CFR 20 and 10 CFR 50 are met during normal operation and that the guidelines suggested in 10 CFR 50.67 and Regulatory Guide 1.183 would be met in the event of the design basis accident (Section 15.4.2).

Virginia Power implemented the revised 10 CFR 20 January 1, 1994. The criteria used for design basis accidents based on the old 10 CFR 20 retain their same definitions and therefore the design basis accident (DBA) analyses do not require recalculation using criteria of the revised 10 CFR 20 rule. (

Reference:

First set of NRC Question/Answer #14.)

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

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

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

The original design of this radiation shielding was based upon radiation zone criteria which were established in support of the expected access requirements and durations of occupancy during normal operations and during refueling outages. Descriptions of the zone criteria are presented in Table 12.1-1, and the detailed radiation zone criteria for normal and shutdown '

operations are illustrated on Figures 12.1-1 through 12.1-5. These figures do not represent operational requirements and should be considered HISTORICAL.

Design dose rates are based on the expected frequency and duration of occupancy. Values of design dose rates are upper limits and are based on conservative assumptions. Representative operating dose rates are expected to be much lower than the design dose rates reported.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-2 Occupancy time is such that individual radiation doses will be within the requirements of 10 CFR 20.

Radiation zones are shown on Figure 12.1-1 through 12.1-5 for the containment building, auxiliary building, fuel building, decontamination building, and waste disposal building. The zones are defined in Table 12.1-1.

The service building and onsite environs are Zone 1 throughout. During special operations, local areas within the service building or near the contaminated storage pad or spent-fuel-cask-handling area may temporarily exceed these normal limits; during such times the area will be defined in accordance with health physics procedures.

The average dose rate at the exclusion boundary is such that the exposure of an individual would not be greater than 5 mrem/yr. from all sources of direct radiation at the site. All shielding dose rate calculations are based on 1% failed fuel elements.

Maximum accident doses shall not exceed the following:

Accident or Case Control exclusion area boundary (EAB)

Room & low population zone (LPZ)

Design Basis Loss-of-Coolant Accident 5 rem TEDE 25 rem TEDE (LOCA)

Steam Generator Tube Rupture Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Coincident Iodine Spike 5 rem TEDE 2.5 rem TEDE Main Steam Line Break Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Coincident Iodine Spike 5 rem TEDE 2.5 rem TEDE Locked Rotor Accident 5 rem TEDE 2.5 rem TEDE Rod Ejection Accident 5 rem TEDE 6.3 rem TEDE Fuel Handling Accident 5 rem TEDE 6.3 rem TEDE 12.1.2 Design Description Building arrangements and machine location drawings of Units 1 and 2 structures, showing plan and sectional views, are given in Section 1.2.2. The plot plan and site plan are shown on Reference Drawings 3 and 4.

12.1.2.1 Primary Shielding Primary shielding is provided to limit radiation emanating from the reactor vessel. Such radiation consists of neutrons diffusing from the core, prompt fission gammas, fission product

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-3 gammas, and gammas resulting from the slowing down and capture of neutrons. The primary shielding is designed to:

1. Attenuate neutron flux to prevent excessive activation of components and structures.

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

3. Reduce the contribution of radiation from the reactor to optimize the thickness of the secondary shields.

The primary shield consists of a water-filled neutron shield tank and a concrete shield. The neutron shield tank has a radial thickness of approximately 3 feet, and it is surrounded by 4.5 feet of reinforced concrete. The shield tank prevents the overheating and dehydration of the primary shield wall concrete and minimizes the activation of the plant components within the reactor containment. A cooling system is provided for the water in the neutron shield tank. (The neutron shield tank cooling water subsystem is discussed in Section 9.2.2.)

A 15 ft. 8 in. high x 2 inch thick cylindrical lead shield located beneath the neutron shield tank protects station personnel servicing the neutron detectors during reactor shutdown.

Appendix 12A contains a detailed description of supplementary neutron shielding. The manway in the upper part of the primary shield is plugged during reactor operation. The control-rod drive concrete missile shield located above the reactor vessel is designed to provide some additional neutron shielding. The primary shield arrangement is shown on Figure 12.1-6.

The shield materials and thicknesses are listed in Table 12.1-2. The application of Permali material for supplementary neutron shielding is shown on Figure 12.1-7 for Unit 1.

A 3-1/2 inch thick stainless steel radiation shield is provided at the 12-inch diameter Incore Sump Room drain to protect station personnel during normal power operation and during refueling outages.

12.1.2.2 Secondary Shielding Secondary shielding consists of the shielding for the reactor coolant, the reactor containment, fuel handling equipment, auxiliary equipment, the waste storage area, and the yard, as well as accident shielding.

Nitrogen-16 is the major source of radioactivity in the reactor coolant during normal operation, and its shielding requirements control the combined thickness of the crane and containment walls. In areas such as the auxiliary building, where N-16 is not the major source of activity, activated corrosion and fission products from the reactor coolant system control the secondary shielding. Activated corrosion and fission products in the reactor coolant system also result in the shutdown radiation levels in the reactor coolant loop areas. Tables 11.1-6 and 12.1-3

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-4 list the activities used in designing the containment secondary shielding. Table 11.1-6 lists the fission product activities and activated corrosion products in the reactor coolant system with 1%

failed fuel. Table 12.1-3 lists the activated corrosion product activities and the N-16 activity at the reactor vessel outlet nozzle.

12.1.2.3 Reactor Coolant Loop Shielding Interior shield walls separate the reactor coolant loop, pressurizer, incore instrumentation, and containment access sectors. This shielding allows access to the incore instrument sector during normal operation and facilitates maintenance in all sectors during shutdown. The crane support wall provides limited access protection in the annulus between the crane wall and the reactor containment wall and provides part of the exterior shielding required during power operation. Shield walls are provided around each steam generator above the operating floor to a height required for personnel protection. Shielding beams below the operating floor are strategically positioned around the steam generators and reactor coolant pumps. The shielding beams provide protection for personnel in the wall annulus from gamma streaming up through the relief openings in the operating floor. The shielding arrangement is shown in Figures 12.1-6, 12.1-9, and 12.1-10.

12.1.2.4 Containment Structure Shielding The containment shielding consists of the steel-lined, steel-reinforced concrete cylinder and hemispherical dome as described in Section 3.8.2. This shielding, together with the crane support wall, attenuates radiation during full-power operation and during the assumed design basis accident to or below design levels at the outside surface of the containment and at the site boundaries.

12.1.2.5 Fuel-Handling Shielding Fuel-handling shielding is designed to facilitate the removal and transfer of spent-fuel assemblies from the reactor vessel to the spent-fuel pit. It is designed to protect personnel against the radiation emitted from the spent-fuel and control-rod assemblies.

The refueling cavity above the reactor vessel is flooded to approximately Elevation 290 to provide a temporary water shield above the components being withdrawn from the reactor vessel.

The water height is thus approximately 26 feet above the reactor vessel flange. This height ensures approximately 7 feet of water above the active portion of a withdrawn fuel assembly at its highest point of travel. Under these conditions, the dose rate is less than 50 mRem/hr at the water surface.

After removal of the fuel from the reactor vessel, it is moved to the spent-fuel pit by the fuel transfer mechanism via the fuel transfer canal. The fuel transfer canal is a passageway connected to the reactor cavity and extending to the inside wall of the containment structure. The canal is

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-5 formed by two shield walls extending upward to the same height as the reactor cavity. During refueling, the canal and the reactor cavity are flooded with water to the same height.

The spent-fuel pit in the fuel building is permanently flooded to provide approximately 7 feet of water above a fuel assembly when it is being withdrawn from the fuel transfer system.

Water height above stored fuel assemblies is a minimum of 23 feet. The sides of the spent-fuel pit, three of which also form part of the fuel building exterior walls, are 6-foot-thick concrete to ensure a dose rate of no more than 2.5 mRem/hr outside the building.

Approximately 3 feet of concrete shielding is provided above and on each side of the fuel transfer tubes in the area between the reactor containment wall and the fuel building wall, and in the area between the reactor containment wall and the fuel transfer canal.

12.1.2.6 Auxiliary Equipment Shielding The auxiliary components exhibit varying degrees of radioactive contamination due to the handling of various fluids. The auxiliary shielding protects operating and maintenance personnel working near the various auxiliary system components, such as those in the Chemical and Volume Control System, the boron recovery system, the waste disposal system, and the sampling system.

Controlled access to the auxiliary building is allowed during reactor operation. Major components of systems are individually shielded so that compartments may be entered without having to shut down and possibly decontaminate the entire system. Ilmenite concrete is used in certain shields.

Potentially highly contaminated ion exchangers and filters are located in the ion-exchange structure along the south wall of the auxiliary building. Each ion exchanger or filter is enclosed in a separate, shielded compartment. The concrete thicknesses provided around the shielded compartments are sufficient to reduce the dose rate in the surrounding area to less than 2.5 mRem/hr and the dose rate to any adjacent cubicle to less than 100 mRem/hr. The shielding thicknesses around the mixed-bed demineralizers are based upon a saturation activity that gives a contact radiation level of nearly 12,000 rem/hr.

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

12.1.2.7 Waste Storage Shielding The waste storage and processing facilities in the auxiliary building, decontamination building, and clarifier building are shielded to protect operating personnel in accordance with the radiation protection design bases set forth in Section 12.1.1.

Boron recovery tanks, which are used to store letdown before recycling to the station or processing as waste, are shielded to reduce dose rates to 2.5 mRem/hr in accessible areas. Boric acid storage tanks are located in the auxiliary building so that shielding may be installed if necessary during station operation.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-6 The waste gas decay tanks are located in shielded cubicles, which are buried for missile protection. The resulting dose rate at the ground surface above the tanks is less than 0.75 mRem/hr.

Periodic surveys by Health Physics personnel using portable radiation detectors ensure that radiation levels outside the shield walls meet design specifications, and they establish access limitations within the shielded cubicles. In addition, continuous surveillance is provided in the waste solidification area of the decontamination building and in the control board area by area radiation monitors.

12.1.2.8 Accident Shielding Accident shielding is provided by the reactor containment, which is a reinforced-concrete structure lined with steel. For structural reasons, the thicknesses of the cylindrical walls and dome are 54 inches and 30 inches, respectively. These thicknesses are more than adequate to meet the guideline limits of 10 CFR 50.67 at the exclusion boundary.

Additional shielding is provided for the main control room. This, together with the shielding afforded by its physical separation from the containment structure, ensures that an operator would be able to remain in the main control room for 30 days after an accident and not receive a dose in excess of 5 rem TEDE.

12.1.2.9 Boron Recovery Tank Shielding The boron recovery tanks (see Section 12.1.2.7), are shielded to the height required for personnel protection on the site and to ensure that the dose rate at the exclusion boundary from direct radiation does not exceed the design dose rates as specified in Table 12.1-1.

12.1.2.10 Main Control Room Shielding The main control room is shown in Figure 1.2-3 and on Reference Drawing 5.

The design basis for the control room envelope is that the radiation dose to personnel inside the control room envelope (from sources both internal and external to the control room envelope) be less than or equal to 5 rem TEDE for the 30 day duration of the design basis accident. The control room northern, western, and eastern walls are 2' thick concrete. The southern wall of the control room is 18" thick concrete. The southern wall of the cable vault is 2' thick concrete to bring the total concrete shielding on the side of the control room facing the containment to 42".

The ceiling for the control room is 2' thick concrete. The doorways to the control room are on the northern wall of the control room facing away from the containment structure and can be covered with radiation shielding doors. Based on NUREG-0800, Section 6.4 (Reference 8), this level of shielding allows the dose in the control room from containment shine and cloud shine to be treated as negligible.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-7 Special consideration has been given to the design of penetrations and structural details of the main control room to establish an acceptable condition of leaktightness.

The air conditioning systems are installed within the spaces served and designed to provide uninterrupted service under accident conditions. On an emergency signal, the control room normal replenishment air and exhaust systems are isolated automatically by tight closures in the ductwork. Breathing-quality air is discharged from high-pressure storage bottles to the MCR/ESGR envelope. The MCR/ESGR envelope is also provided with an emergency ventilation system fitted with particulate and impregnated charcoal filters to introduce cleaned outside air into the protected spaces within an hour after an accident. This can continue indefinitely to supply breathable quality air to the MCR/ESGR envelope. Fan/filter units also start in recirculation during bottled air discharge to account for inleakage during MCR/ESGR envelope access.

The radiation level in the main control room is measured by a fixed monitor to verify safe operating conditions. Portable monitors are available to provide backup to the fixed monitors.

As an additional precaution, personnel air packs are available in the control area.

12.1.2.11 Shielding Review for NUREG-0578 In response to the requirements of NUREG-0578, a design review was conducted using the Stone & Webster Engineering Corporation GAMTRAN1 computer code with inputs from the ACTIVITY-2 and RADIOISOTOPIC computer codes. The NRC-specified source terms were used. All systems designed to function after an accident were considered as sources, including safety injection, recirculation spray, hydrogen recombiner, sampling, auxiliary building sump, and drain lines. The letdown portion of the chemical and volume main control system was excluded because it is isolated and because its use in the post-accident situation would be unacceptable. All vital areas were identified and evaluated. Areas where continuous occupancy is required are the main control room, the technical support center, the counting room, the operational support center, and the security control center. Limited access is needed to such places as emergency power supplies and sampling stations.

All the NUREG-0578 Category A requirements have been satisfied at North Anna Units 1 and 2, as indicated by letter, A. Schwencer, NRC, to J. H. Ferguson, VEPCO, dated April 23, 1980.

12.1.3 Source Terms The total quantity of the principle nuclides in process equipment that contains or transports radioactivity is identified as a function of operating history in Chapter 11. Design and expected values of the radioisotopic inventory for both the reactor coolant and main steam systems are listed in Section 11.1. Design and expected values of the radioisotopic inventory for each portion of the radioactive liquid waste system are listed in Section 11.2.5 and for the waste gas decay tank in the gaseous waste disposal system in Section 11.3.5.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-8 Table 11.1-11 lists the activities in the volume control tank using the assumptions summarized in Table 11.1-5. The activities in the pressurizer (both the liquid and vapor phases) are given in Table 11.1-13 using the assumptions summarized in Table 11.1-5. Saturation activities for demineralizer resins are listed in Table 11.1-13. Spent-fuel activities are listed in Table 11.1-4.

Process piping designated to carry significant amounts of radioactive materials is located behind shielding to minimize the radiation exposure of plant personnel. Pipe tunnels, chases, or shafts are provided as required to properly segregate radioactive piping behind shields. Where necessary, extension-stem-operated valves are used.

Concrete, exposed carbon steel, and galvanized carbon steel surfaces within the fuel, auxiliary, decontamination, and waste disposal buildings that require protective coatings and may be subject to decontamination are typically finished with epoxy, silicone alkyd, or urethane enamel protective coatings or approved equal. Stainless steel surfaces are not painted. Stainless steel is used extensively in the fuel, decontamination, and waste disposal buildings.

Tanks such as the high- and low-level waste tanks, evaporator bottoms tanks, fluid waste treating tank, and contaminated drain collecting tank have been designed to allow for cleaning and to minimize the buildup of radioactive material using the following factors:

1. These tanks are vertical cylindrical tanks with flanged and dished heads to allow complete draining.

2. The tank outlet lines are at the lowest point of the tank to aid in complete draining.

3. The tanks are of stainless steel construction to minimize corrosion and the buildup of activity and to facilitate cleaning.

4. The tanks are provided with inspection openings or manholes that can be used during cleaning.

Drip pan bedplates are provided under pumps. Individual equipment cubicles and pipe chases containing radioactive fluid system components and equipment have floor drains that are piped to and processed by the waste disposal system.

The sampling system uses small line sizes to maintain high velocity to keep particles in suspension in the fluid stream. The sample lines to the central sample points connect to recirculation lines to permit multivolume flushes of sample lines so that representative samples are drawn. Local check samples are available from the recirculation lines if needed.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-9 12.1.4 Area Monitoring 12.1.4.1 Normal Plant Operations The area radiation monitoring system reads out and records the radiation levels in selected areas throughout the station, and alarms (audibly and visually) if these levels exceed a preset value or if the detector malfunctions. Each detector reads out and alarms both in the main control room and locally. Each channel is equipped with a check source remotely operated from the main control room. Recorders produce a continuous, permanent record of radiation levels while the detectors are functioning. Area-radiation-monitoring channels for Unit 1 are powered from the Unit 1 480V emergency buses; channel monitoring systems or areas common to both units are powered from the emergency bus for either Unit 1 or Unit 2.

The area radiation monitors are designed for continuous operation. Continuous, as used to describe the operation of an area radiation monitor, means that the monitor provides the required information at all times with the following exceptions: (1) the monitor is not required to be in operation because of specified plant conditions given in the Technical Requirements Manual, or (2) the monitor is out of service for testing or maintenance and approved alternate monitoring methods are in place.

The monitor locations, shown on Reference Drawings 1, 2, and 6, give an early warning of high radiation levels when plant personnel enter various portions of the plant. To perform this function they are generally located near the main entrance pathway for a given building or portion thereof. In some areas they are located at the major work area involved. In all cases they provide a representative indication of the radiation level in that vicinity of the plant and not necessarily the maximum that might be measured against one of the nearby shield walls. The audio and visual alarm provides adequate warning to personnel in the event of an abnormally high radiation level.

These monitors have remote displays in the main control room indicating the radiation levels throughout the plant, and they may be monitored before entry into potentially high radiation fields. When radioactive material is being handled within a given area, such as the decontamination building, the monitors provide a representative reading based on planned work areas for handling such material.

In addition, if the dose rate at the manipulator crane area monitor exceeds a preset value, the alarm automatically trips the containments purge air supply and exhaust fan and closes the purge system butterfly valves, thus isolating the containment from the environment.

The alarm setpoint of each area monitor is variable, and it is set at a radiation level slightly above that of normal background radiation in the respective area. The monitoring equipment consists of fixed-position gamma detectors and associated electronic equipment. These channels warn of any increase in radiation level at locations where personnel may be expected to remain for extended periods of time. The instruments and their ranges and locations are listed in Table 12.1-4.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-10 Tests and calibrations of the radiation monitors are performed at intervals specified in the applicable Technical Procedures. Special restrictions, as specified in the Technical Requirements Manual, are imposed on plant operators or maintenance activities if the area monitors are not functional. The manipulator crane monitor is a control function and is part of a redundant alarm system with the containment gaseous and particulate monitors. If the manipulator crane monitor is not functional, the containment gaseous and particulate monitors can still function and can be backed up by local portable equipment. This portable equipment, together with Health Physics surveys during maintenance activities, will allow these activities to continue if a normal fixed area monitor is not functional.

The radiation monitors in the Fuel Building also provide a control function. When a Hi-Hi radiation condition is sensed by either of these monitors, during a fuel handling condition, the control room bottled air system will discharge, the control room normal ventilation will isolate, and the control room/emergency switchgear room emergency ventilation system will start automatically to recirculate and filter control room air.

12.1.4.2 Post-Accident Conditions The containment high-range radiation monitoring system (CHRRMS) provides indication in the control room of containment radiation level as required by NUREG-0578, Section 2.1.8.b, and subsequent clarification contained in the NRC letter dated October 30, 1979.

Each containment has two redundant Class I monitor systems consisting of a high range detector (100 - 107 R/hr), a control room readout unit and associated interconnecting cable. The detectors are located approximately 155 degrees apart for Unit 1 and 130 degrees apart for Unit 2 on the inside crane wall to provide physical separation. The location also facilitates the periodic calibration of the detectors since they are close to the operating floor.

The CHRRMS components are qualified to IEEE-323-1974, IEEE-344-1975 and meet the requirements of Regulatory Guide 1.97, proposed Revision 2. The high range monitors are powered from diverse Class 1E vital buses. The indicators in the control room are installed in racks designed per the separation and seismic requirements of Regulatory Guide 1.75, Revision 1, and IEEE-344-1975 respectively.

The addition of the high-range containment radiation monitors is for indication purposes only and does not affect the logic schemes of any safety-related systems.

The Technical Support Center (TSC) radiation monitoring system is localized and satisfies the guidelines established in NUREG-0696. The radiation monitoring system components consist of a particulate, iodine, and noble gas monitor and two area monitors.

This monitoring system provides continuous indication of the dose rate and airborne activity in the TSC during an emergency, as well as alerting personnel of adverse conditions. This

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-11 system is totally contained within the TSC and is in no way connected to the control room or any safety-related systems.

12.1.5 Operating Procedures A radiation protection program consistent with the requirements of 10 CFR 20 and designed to ensure that doses are kept ALARA is maintained. Applicable HP procedures, (i.e., RWPs), are used to control access to all radiation and contaminated areas.

The station auxiliary systems containing radioactive fluids are designed for remote operation by the use of extensive instrumentation for monitoring, remotely operated pneumatic or electrical control valves, and manually operated valves with extension stems that allow the operator to operate the valves while behind shield walls.

Special tools are used extensively for fuel handling. These tools and processes are described in Section 9.1.4.

The operation of the filter transfer shield, which is used for the handling of spent filter cartridges, is described in Section 11.5.3. This transfer shield is of lead and steel construction and functions only as a transfer and temporary storage device.

A lead shield beneath the neutron shield tank in the containment protects personnel during the servicing of the neutron detectors. This shield is described in Section 12.1.2.1.

A neutron detector carriage provides both distance and material shielding during the changing of the neutron detectors.

Persons or groups entering areas of high radiation are equipped with radiation-monitoring devices. A person entering an area in which the radiation is greater than a predetermined level is accompanied by, or is in constant communication with, at least one other person.

12.1.6 Dose Rate Calculations To indicate the methods used to determine dose rates, two sets of calculations are described below.

12.1.6.1 Sample Sink Area The receptors for the sampling sink are located just off the surface of the concrete wall behind the sinks. Two sources of radiation are considered to be significant in this area: the sample piping, located in a pipe space behind the wall at which the sampling sinks are located; and the volume control tanks, located in individual cubicles behind the pipe space, as shown in Figure 12.1-2 Sh. 3.

The volume control tanks are separated by a 2-foot-thick concrete wall. Concrete density of this and other concrete walls is 146 lb/ft3. On the sampling sink side of the volume control tank,

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-12 the cubicle wall is 2.5-foot-thick concrete. The distance from the axial centerline of a volume control tank to the surface of the sampling sink wall is approximately 18.5 feet.

Each volume control tank was approximated as a source by two right circular cylinders 84 inch in diameter with 0.25-inch steel walls, with liquid volume of 120 ft3 and gaseous volume of 180 ft3.

The sample piping primarily consists of 3/4-inch or smaller tubing containing process fluids. The piping is located behind an 18-inch concrete wall. For the purpose of this analysis, the maze of pipes was approximated by four disks side-by-side along the wall behind the sampling sinks, each 0.75 inch thick and 6 feet in diameter. Each disk was assumed to be covered by a steel plate of minimal thickness to represent the pipe wall thickness.

A reduction factor was applied to the source intensity to account for the piping density.

Although the fluid in the pipes comes from many different process streams, the conservative assumption was made that all pipes contained primary coolant samples drawn from the hot leg of the coolant loop. Primary coolant activities are listed in Table 11.1-6.

The computer code GAMTRAN described in Section 12.1.6.3 was used to calculate the dose rate from each source. At a receptor located on the line passing through the center of the disk representing the sample pipes and coincident with the disk axis and intersecting the cylindrical axis of one of the volume control tanks, the dose rate was calculated to be 4.1 mRem/hr. Of the total, the sample piping contributed approximately 97%.

12.1.6.2 Valve-Operating Area Outside Demineralizer Cubicle In the valve-operating area outside the demineralizer cubicle on the 244-foot level of the auxiliary building, typical receptor locations were chosen at 3- and 6-foot heights above the 244-foot level, lying on a plane perpendicular to the vertical shield wall, passing through the cylindrical axis of the mixed-bed demineralizer, and at the outside surface of the shield wall.

The mixed-bed demineralizer was chosen as the source because it is the most radioactive source in the area and because the concrete shielding between the mixed-bed demineralizer and the receptors is the same thickness as that between other demineralizers.

The mixed-bed demineralizer is assumed to be a right circular cylinder source inside a 5/16-inch mild steel shield with source strengths based on Surry Power Station source data corrected to North Anna power level.

The volume of the demineralizer resin is assumed to be 39 ft3 with a height of 7.13 feet.

A 2-foot-thick concrete wall extends vertically from Elevation 244 to the floor below the demineralizer cubicle. Above the floor, the wall is 4-foot-thick concrete. The floor of the demineralizer cubicle is 2-foot-thick concrete. Concrete density in all cases is taken as 146 lb/ft3.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-13 The computer code GAMTRAN, described below, was used to calculate the dose rates at the receptors. Calculated dose rates at each receptor were less than 1 mRem/hr from the mixed-bed demineralizer.

12.1.6.3 GAMTRAN Computer Code The GAMTRAN code is a Stone & Webster developed point kernel code for shield design analysis. The gamma ray attenuation coefficients used in GAMTRAN are generated using the OGRE (Reference 1) pair production and photoelectric cross sections. The Compton scattering component is calculated by the Klein-Nishina equation.

Gamma ray buildup factors are generated by a two-parameter formula based on the work of Berger (Reference 2) and Chilton (Reference 3). The parameters used for the buildup factors are based on data from the Weapons Radiation Shielding Handbook (Reference 4). Flux-to-dose conversion factors were based on curves in the Reactor Shield Design Manual (Reference 5).

12.1.7 Estimates of Exposure Radiation shielding is provided on the basis of maximum concentrations of radioactive materials within each shielded region (e.g., 1% failed fuel) rather than the annual average values.

For batch processes, as an example, the point of the highest radionuclide concentration in the batching process (e.g., just before draining a tank) is assumed. The shielding designs are therefore intentionally conservative in that the dose rates reflect maximum rather than average sources to be shielded.

The design objectives of the plant shielding for normal operation in terms of maximum dose rates allowed at in-plant locations are given in Table 12.1-1. It is expected that the average dose rates would be less than 20% of these values.

Shielding thicknesses were calculated using the Stone & Webster code GAMTRAN described in Section 12.1.6.3. Table 12.1-5 lists the densities of the materials used for shielding calculations. Care was taken to ensure that the material actually used for construction was at least as dense as that used for analyses. Figures 12.1-6, 12.1-9, and 12.1-10 show the shielding arrangement for the containment. Arrangements for the other buildings are shown in Section 1.2.

Supplementary neutron shielding is discussed in detail in Appendix 12A.

12.1.7.1 Considerations for Dose Predictions It is general practice to arrive at the radiation zoning by taking liberal estimates of the time to be spent in each zone and dividing this into 100 mrem/week to arrive at a design value in terms of mRem/hr that will not be exceeded in that zone, even under worst-case conditions. The shielding is then designed assuming maximum conditions to ensure that these exposure values are never exceeded under normal operating conditions. (Higher doses may result from specific repair jobs when shielding is not possible.)

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-14 The radiation zone designations are shown in Figures 12.1-1 through 12.1-5. These delineate the maximum dose rates at all locations within the major buildings of North Anna Units 1 and 2.

Because of the conservatism employed in performing the worst-case dose rate calculations, the shielding is conservatively designed, thus ensuring that the average exposures in each zone will be far less than the maximum.

To compute the expected man-rem values per zone and throughout the plant, the following items should be considered:

1. Time-and-motion study data must be obtained to allocate time spent in each zone in the plant such that the sum of these times equals the total time the employee is at the station in an average year.

2. An average employee concept would not apply because some employees never go in some zones, whereas others frequently spend time in these zones.

3. Once in a zone, movement within the zone must be considered.

4. The innumerable large and small components in each zone that act as object shields would have to be factored into the dose assessment. This would complicate the analytical models and require several times the man-months required presently to perform the worst-case type of analysis in which such component object shielding is conservatively ignored.

5. Similarly, a number of components located in the regions being shielded would also have to be included in the modeling to compute expected values. Most of them are conservatively left out of the worst-case analysis.

6. Conservatism in sources (e.g., 1% failed fuel design defect versus 0.2% expected) would have to be eliminated to predict expected dose rates.

7. Explicit margins in other source terms would have to be factored out of the analysis.

8. In the worst-case model, each source is assumed to be at maximum levels. This assumes all other sources in that system are at minimum levels. Viewed plant wide, however, an activities balance would have to be used for average expected conditions.

9. Much more complicated mathematical models of large components would have to be developed to replace the few region models which are presently used to intentionally overestimate the emanation of radiation from these large sources.

A man-rem analysis cannot be computed with sufficient accuracy to obtain good data of a predictive nature. However, sufficient operating data on similar plants do exist to provide estimates of man-rem doses for the station as a whole. This operating experience is demonstrative of the fact that the radiation shielding is conservatively designed. This is a direct result of the

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-15 design of shields for worst-case conditions, conservative dose rate calculations, and implicit and explicit designers margins.

12.1.7.2 Reports From Other Plants Relative to the estimations of exposure levels during maintenance, refueling, and inservice inspection activities, such estimates do not lend themselves to prediction analysis based on an analytical modeling. Reliance should be placed on operating experience at other stations as the most reasonable source of such data. In this connection, VEPCOs engineers participated in the efforts of the Atomic Industrial Forums Task Force on Occupational Exposures.

One survey reported by Charlesworth (Reference 6) at the April 1971 American Power Conference covered data obtained at seven operating water-cooled reactor plants with a total plant worker dose of 1700 man-rem during the previous year for an average of 244 man-rem/yr per plant. In this survey, it was found that on an average 75% of these exposures were estimated to have been received during shutdown operations.

Another survey by Goldman (Reference 7) summarizes the results of 27 plant-years of operation from operating reports. This survey indicated a range of 0.5 to 2.3 rem/yr with limited data on the number distribution of staff in several exposure categories. From these data, Goldman concluded that 19 plant-years of operating data resulted in an in-plant population average of 238 man-rem per plant-year. These results are close to the 244 man-rem per plant-year reported by Charlesworth.

The average dose rate level in the visitors center will be less than 0.01 mRem/hr above natural background based on the worst-case assumption. Assuming that a visitor will spend 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at the visitors center four times per year, he would receive a dose of less than 0.16 mrem/yr.

The expected annual doses to onsite personnel are governed by the controls imposed by the station supervision and/or Health Physics personnel. However, dose estimates for in-station personnel for routine operation are expected to parallel those reported from operating plant experience as discussed above.

Extensive radiation shielding is provided based on the maximum concentration of radioactive materials within each shielded region rather than on annual average values. The shielding and occupancy zones for normal operation are intentionally very conservative so that the normally received dose rates should be less than 10% of the limits specified in 10 CFR 20.

The highest level of personnel exposure is expected to occur during shutdown and maintenance periods on systems containing items such as coolant purification filters, cleanup and radwaste demineralizers, ion-exchange resins, charcoal adsorber units, and solid-radwaste-handling components. Since this is the case, the plant shielding and machinery

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-16 locations have been designed to provide maximum laydown space, maximum working room, and minimum time required to perform operations consistent with the reasonable operation of the plant. Experience gained in the operation of nuclear plants has been factored into these designs with the objective of minimizing the total man-rem exposure to plant personnel.

12.1.7.3 Dose From Stored Waste For the purpose of a conservative analysis, it is assumed that 1 Ci of cobalt-60 equivalent is stored in the low-level contaminated storage area (Reference Drawing 4). The dose rates at the various distances, including the site boundary, per curie of cobalt-60 equivalent, are presented in Figure 12.1-11. No credit is taken for the drum shielding and self-shielding of the waste stored outside the building.

12.1.7.4 Health Physics Area Dose Evaluation The Health Physics office, counting room, and monitoring area complex in the service building is, under normal operating conditions, a continuous access area. The only anticipated radioactive sources in this area are radioactive samples brought in for analysis and radioisotopes used in analytical equipment such as radiation monitoring equipment. Therefore, any radiation doses received while in this area will be controlled by adherence to standard health physics practices for handling radioactive material. Shielding design for the station as a whole ensures that contributions from other station areas do not exceed the design levels for their respective areas and make no significant contribution to the service building dose rate.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-17

12.1 REFERENCES

1. Oak Ridge National Laboratory, OGRE - General Purpose Monte Carlo Gamma Ray Transport Code System, RSIC Code Package CCC-46, Oak Ridge, Tennessee, 1967.

2. M. J. Berger, in Proceedings of Shielding Symposium, U.S. Naval Radiological Defense Laboratory, Reviews and Lectures No. 29, p. 47.

3. A. B. Chilton, D. Holoviak, and L. K. Donovan, Interior Report Determination of Parameters in an Empirical Function for Buildup Factors for Various Photon Energies.

4. P. N. Stevens and D. K. Trubey, Weapons Radiation Shielding Handbook: Chapter 3 -

Methods for Calculating Neutron and Gamma Ray Attenuation, DNA-1892-3, Defense Nuclear Agency, Washington, D. C., March 1972.

5. T. Rockwell, III, ed., Reactor Shield Design Manual, TID-7004, United States Atomic Energy Commission, March 1956.

6. D. G. Charlesworth, Water Reactor Plant Contamination and Decontamination Requirements, survey conducted by the Subcommittee on Nuclear Systems, ASME Research Committee on Boiler Feedwater Studies, presented at the 33rd Annual Meeting of the American Power Conference, Chicago, April 1971.

7. M. I. Goldman, Radioactive Waste Management and Radiation Exposure, Nuclear Technology, Vol. 14, May 1972.

8. Standard Review Plan 6.4, Control Room Habitability System, 1981.

12.1 REFERENCE DRAWINGS The list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of Station Drawings are controlled by station procedure.

Drawing Number Description

1. 11715-FK-9B Instrument Piping, Radiation Monitoring, Sheet 2, Units 1 & 2

2. 11715-FK-9A Instrument Piping, Radiation Monitoring, Sheet 1, Units 1 & 2

3. 11715-FY-1B Site Plan, Units 1 & 2

4. 11715-FY-1A Plot Plan, Units 1 & 2

5. 11715-FE-27B Arrangement: Main Control Room, Elevation 276'- 9", Units 1 & 2

6. 11715-FK-9C Instrument Piping, Radiation Monitoring, Sheet 3, Units 1 & 2

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-18 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Table 12.1-1 RADIATION ZONE CRITERIA Maximum Dose Zone Access Typical Locations Rate (mRem/hr)

Full-Power Operation Main control room, outside surface of containment, and all turbine plant and I Continuous 0.75 administration areas Passageways of auxiliary and fuel buildings, in general, and inside reactor containment II Periodic 2.5 personnel lock III Limited 15 Outside surface of shielded tank cubicles Annulus between crane wall and containment IV Controlled 100 wall V Restricted Over 100 Inside shielded equipment compartments Hot Shutdown (after 15-min decay)

Reactor containment above operating floor; III Limited 15 outside of crane wall V Restricted Over 100 Inside shielded equipment compartments Cold Shutdown for Maintenance (after 8-hr decay)

Reactor containment above operating floor and II Periodic 2.5 outside of crane wall V Restricted Over 100 Inside shielded equipment compartments Cold Shutdown for Refueling Reactor containment above operating floor, outside of crane wall, and adjacent to fuel transfer canal near incore instrumentation II Periodic 2.5 devices V Restricted Over 100 Inside shielded equipment compartments Surface of water over Above fuel assembly when over upender or raised fuel assembly 50 racks

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-19 Table 12.1-2 CONTAINMENT SHIELDING

SUMMARY

Symbol Figure Shield Description Materiala Thickness (in)

A 12.1-9 Neutron shield tank Water 34 Steel 3 B 12.1-9 Primary shield Concrete 54 12.1-7 Supplementary neutron Permali shield 6 E 12.1-9 Neutron shield tank Steel 1.5 support Lead 2 F 12.1-6 and Cubicle - crane support Concrete 12.1-9 wall 33 F 12.1-9 Shielding beams Concrete 24 G 12.1-9 Crane support wall Concrete 24 H 12.1-6 and Containment wall Concrete 12.1-9 54 I 12.1-9 Containment dome Concrete 30 J 12.1-9 Floor elevation 243 ft Concrete 42 - 48 K 12.1-9 Operating floor Concrete 24 L 12.1-6 and Refueling cavity wall Concrete 12.1-9 42 M 12.1-9 and Control-rod drive Concrete 12.1-10 missile shield 24 N 12.1-9 Refueling cavity water Water 108 O 12.1-9 and Removable block wall 12.1-10 Facing personnel hatch Concrete 18 All others Concrete 12 P 12.1-6 Fuel transfer canal wall Concrete (containment structure) 54 Q 12.1-6 Fuel transfer canal wall Concrete (containment structure) 72 R 12.1-6 Fuel transfer tube Concrete shielding 36 (min)

S 12.1-6 Fuel transfer canal wall Concrete (fuel building) 72 T 12.1-6 Incore instrumentation Concrete cubicle wall 42

a. All poured concrete is reinforced with steel.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-20 Table 12.1-2 (continued)

CONTAINMENT SHIELDING

SUMMARY

Symbol Figure Shield Description Materiala Thickness (in)

U 12.1-6 Cubicle wall Concrete 36 V 12.1-6 Regenerator heat Concrete exchanger wall 24 W 12.1-6 Cable vault wall Concrete 24 X 12.1-6 Auxiliary feed pump Concrete wall 36 Y 12.1-6 Safeguards area wall Concrete 12 Unit 2 only Z 12.1-9 Incore sump room drain Stainless Steel 3 1/2

a. All poured concrete is reinforced with steel.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-21 Table 12.1-3 N-16 AND ACTIVATED CORROSION PRODUCT ACTIVITY Isotope Activity ( µCi/cc @ 577°F)

Mn-54 5.6 x 10-4 Mn-56 2.1 x 10-2 Fe-59 7.5 x 10-4 Co-58 1.8 x 10-2 Co-60 5.4 x 10-4 N-16 a 73.3

a. At the reactor vessel outlet nozzle at 2910 MWt.

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-22 Table 12.1-4 AREA RADIATION MONITORING LOCATIONS, NUMBER AND RANGES Channel Location (number) Range (mRem/hr)

Reactor containment area - low range (2)

(1/2-RM-RMS-163/263) 10-1104 Personnel hatch area (2)

(1/2-RM-RMS-161/261) 10-1104 Manipulator crane (2)

(1/2-RM-RMS-162/262) 10-1104 Incore instrumentation transfer area (2)

(1/2-RM-RMS-164/264) 10-1104 Decontamination area (1)

(1-RM-RMS-151) 10-1104 New fuel storage area (1)

(1-RM-RMS-152) 10-1104 Fuel pit bridge (1)

(1-RM-RMS-153) 10-1104 Auxiliary building area (1)

(1-RM-RMS-154) 10-1104 Waste solidification area (1)

(1-RM-RMS-155) 10-1104 Sample room (1)

(1-RM-RMS-156) 10-1104 Main control room (1)

(1-RM-RMS-157) 10-1104 Laboratory (1)

(1-RM-RMS-158) 10-1104 Technical Support Center (2)

(1-RM-RMS-185/186) 0.05 - 106 Local Emergency Operations Facility (2)

(1-RM-RMS-187/188/189) 10-1104

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-23 Table 12.1-5 MATERIALS USED FOR SOURCE AND DOSE RATE CALCULATIONS Material Density (lb/ft3)

Ilmenite concrete 240 Ordinary concrete 146 Steel 490.5 Lead 707.6 Air, steam, or vapor 0.075 Water Pressurized reactor coolant 46 All other 62.4 Core 273.4

Revision 56--Updated Online 09/30/20 NAPS UFSAR 12.1-24 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Figure 12.1-1 (SHEET 1 OF 8)