ML17033B570

Redacted Updated Final Safety Analysis Report Chapter 12

ML17033B570
Person / Time
Site:	North Anna
Issue date:	09/29/2016
From:	V Sreenivas Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Sreenivas V, NRR/DORL/LPL2-1, 415-2597
Shared Package
ML17033B477	List: ML17033B480 ML17033B492 ML17033B500 ML17033B509 ML17033B515 ML17033B518 ML17033B525 ML17033B531 ML17033B533 ML17033B540 ML17033B545 ML17033B550 ML17033B570 ML17033B573 ML17033B575 ML17033B583 ML17033B585 ML17033B590 ML17033B591
References
Download: ML17033B570 (96)

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North Anna Power Station Updated Final Safety Analysis Report Chapter 12 Intentionally Blank

Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 12-i12.1SHIELDING. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-112.1.1Design Objectives. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-112.1.2Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-212.1.2.1Primary Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-2 12.1.2.2Secondary Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-312.1.2.3Reactor Coolant Loop Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-412.1.2.4Containment Structure Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-412.1.2.5Fuel-Handling Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-4 12.1.2.6Auxiliary Equipment Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-512.1.2.7Waste Storage Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-5 12.1.2.8Accident Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-612.1.2.9Boron Recovery Tank Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-612.1.2.10Main Control Room Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-612.1.2.11Shielding Review for NUREG-0578. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-712.1.3Source Terms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-7 12.1.4Area Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-9 12.1.4.1Normal Plant Operations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-912.1.4.2Post-Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1012.1.5Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1112.1.6Dose Rate Calculations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1112.1.6.1Sample Sink Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1112.1.6.2Valve-Operating Area Outside Demineralizer Cubicle. . . . . . . . . . . . . . . . . .12.1-1212.1.6.3GAMTRAN Computer Code. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1312.1.7Estimates of Exposure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1312.1.7.1Considerations for Dose Predictions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-13 12.1.7.2Reports From Other Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1512.1.7.3Dose From Stored Waste. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1612.1.7.4Health Physics Area Dose Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1612.1References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12

.1-1712.1Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-1712.2VENTILATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-112.2.1Design Objectives. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-1 12.2.2Design Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-112.2.2.1Auxiliary Building. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-2Chapter 12: Radiation ProtectionTable of ContentsSectionTitle Page Revision 52-09/29/2016 NAPS UFSAR 12-iiChapter 12: Radiation ProtectionTable of Contents (continued)SectionTitle Page12.2.2.2Containment Structure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-212.2.2.3Turbine Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-312.2.2.4Fuel Building. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-312.2.3Source Terms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-3 12.2.4Airborne Radioactivity Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-312.2.5Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-6 12.2.5.1Filter Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-6 12.2.5.2Temporary Air Ducting. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-612.2.6Estimates of Inhalation Doses. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-712.2References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12

.2-912.2Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.2-912.3HEALTH PHYSICS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.3-112.3.1Program Objectives and Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.3-1 12.3.2Facilities and Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.3-212.3.3Personnel Dosimetry. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.3-312.3Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.3-312.4RADIOACTIVE MATERIALS SAFETY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.4-112.4.1Materials Safety Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.4-112.4.2Facilities and Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.4-212.4.3Personnel and Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.4-212.4.4Required Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.4-2Appendix12ADescription of Neutron Supplementary Shield . . . . . . . . . . . . . . . . . . . .12A-i12A.1INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-2 12A.2NEUTRON SHIELD DESIGN CRITERIA. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-212A.3EFFECTIVENESS OF THE SUPPLEMENTARY NEUTRON SHIELD . . . . . .12A-3 12A.4SHIELD DESIGN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-412A.4.1Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A

-412A.4.2Location. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 2A-512A.4.3Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 2A-5 Revision 52-09/29/2016 NAPS UFSAR 12-iiiChapter 12: Radiation ProtectionTable of Contents (continued)SectionTitle Page12A.4.4Supports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 2A-512A.4.5Missile Effects. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-512A.4.6Effect on Containment Sump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-612A.5REACTOR PRESSURE VESSEL SUPPORT INTEGRITY REVIEWS . . . . . . .12A-612AReferences. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A

-7 Revision 52-09/29/2016 NAPS UFSAR 12-ivChapter 12: Radiation ProtectionList of TablesTableTitle PageTable12.1-1Radiation Zone Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-18Table12.1-2Containment Shielding Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-19Table12.1-3N-16 and Activated Corrosion Product Activity . . . . . . . . . . . . . . . . . .12.1-21Table12.1-4Area Radiation Monitoring Locations, Number and Ranges. . . . . . . . .12.1-22Table12.1-5Materials Used for Source and Dose Rate Calculations . . . . . . . . . . . .12.1-23Table12.2-1Equilibrium Activities in Different Plant Buildings (Ci/cm 3). . . . . . . .12.2-10Table12.2-2Estimate of Annual Inha lation Doses to Plant Personnel

a. . . . . . . . . . .12.2-11Table 12A-1Comparison of Calculated Neutron Dose Rates with Measurements Made at NorthAnna Unit1, Adjusted to 100% Power. . . . . . . . . . . . . . . . . . . .12A-8Table 12A-2Calculated Neutron Dose Rates with Supplementary Neutron Shielding12A-9Table 12A-3Reactor Pressure Vessel Support and Neutron Shield Tank Loads Phase12A-10Table 12A-4Reactor Pressure Vessel Nozzle Support Loads Phase, Including Reactor Pressure Vessel Internals Movement, As ymmetric Pressure, Deadweight, and Seismic. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-11Table 12A-5Relative Disp lacement Between Top and Bottom of Nozzle Support a 12A-12Table 12A-6Survey Results of Unit1 Reactor Containment at the 291ft. Elevation on 11/10/10. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-13Table 12A-7Survey Results of Unit2 Reactor Containment at the 291ft. Elevation on 10/20/10. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-14 Revision 52-09/29/2016 NAPS UFSAR 12-vChapter 12: Radiation ProtectionList of Figures FigureTitle PageFigure 12.1-1Radiation Zones Containment Structure . . . . . . . . . . . . . . . . . . . . . . .12.1-24Figure 12.1-2Radiation Zones Auxiliary Building . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-32Figure 12.1-3Radiation Zones Fuel Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-35Figure 12.1-4Radiation Zones Decontamination Building . . . . . . . . . . . . . . . . . . . .12.1-37Figure 12.1-5Radiation Zones Waste Disposal Building . . . . . . . . . . . . . . . . . . . . .12.1-39Figure 12.1-6Shield Arrangement-Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-40Figure 12.1-7Permali Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-41 Figure 12.1-8Shield Arrangement Elevation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12.1-42Figure 12.1-9Shield Arrangement Plan Operating Floor. . . . . . . . . . . . . . . . . . . . . .12.1-43Figure 12.1-10Dose Rate Per Curie of Co-60 Equivalent vs. Distance from Low Level Contaminated Storage Area. . . . . . . . .12.1-44Figure 12A-1Plan View of Operating Floor Showing Detector Locations. . . . . . . .12A-15Figure 12A-2Collar Details. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-16Figure 12A-3Plan View of Unit2 Containment for Survey Points. . . . . . . . . . . . . .12A-17Figure 12A-4Shield Dust Cover Blocks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-18Figure 12A-5Crane Wall Openings With Permali Elevation 291ft. 10 in.. . . . . . . .12A-19Figure 12A-6Location of Supplementary Neutron Shields. . . . . . . . . . . . . . . . . . . .12A-20Figure 12A-7RPV Nozzle Support Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-21Figure 12A-8Plan View of Unit1 Containment for Survey Points. . . . . . . . . . . . . .12A-22 Revision 52-09/29/2016 NAPS UFSAR 12-vi Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 12-1CHAPTER 12RADIATION PROTECTIONAt the NorthAnna Power Station, entrance to the station proper is controlled by stationsecurity. 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 with in the protecte d area. From aradiological access standpoint , the area within the pr otected area is the pr imary restricted area.

Other secondary restricted areas exist within the station proper but outsi de 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 Physic s procedures are imple mented as detailed inSections12.1.5 and12.3. It is anticipated that, dur ing normal station opera tion, areas outside the established restricted areas will not experience radiation levels sufficient to classify them asrestricted areas in the context of 10CFR20. However, if such radiation levels were to occur, they would be detected by periodic radiation survey s and appropriate radiation protection measureswould be established for such areas in accordance with Section12.3.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 10CFR20 and a managementcommitment. Management assumes the responsibil ity for ensuring the implementation of thispolicy 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 co llective dose to personnel, i.e., total man-rem exposure, as low as is reasonably achievable.To attain the goal of this commitment, system, st ation, and contractual personnel shallintegrate their efforts as necessary to perform their func tions in such a manner that exposure(s) to radiation will be maintained ALARA. As appli cable, new procedures shall be formulated whileexisting procedures and practices shall be reviewed and modified, if necessary, to ensure their conformance to the principle of maintaining exposures ALARA.

Revision 52-09/29/2016 NAPS UFSAR 12-2 Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 12.1-112.1SHIELDING12.1.1Design ObjectivesRadiation protection, including radiation shield ing, is designed to ensure that the criteriaspecified in 10CFR20 and 10CFR50 are met dur ing normal operation and that the guidelinessuggested in 10CFR50.67 and Regulatory Guide1.183 would be met in the event of the designbasis accident (Section15.4.2).Virginia Power implemented the revised 10CFR20 January1,1994. The criteria used fordesign basis accidents based on the old 10CFR20 re tain their same defin itions and therefore the design basis accident (DBA) analyses do not requ ire recalculati on using criteria of the revised10CFR20 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 crit eria based on zone criteria. Although these criteria were used to establish the original shie ld design, they were neve r intended to establish requirements for the radiation pr otection implementation during pl ant 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-sp ecific and are established through the implementation of the Radiation Protection Program and ALARA Program. These programs evaluate the need for a combinat ion of exposure saving principals such as reduced source term, decreasing occupancy time, or in creased shielding. These program s use shielding as one methodto help ensure compliance with 10CFR20.

This section provides the basi s for the original plant sh ielding design. Although currentdose rates may not be consistent with the zone maps in this chapter, these maps are not being changed to be current, as that w ould make them inconsis tent with the original design basis criteria for the shielding. Recent Heath Physics surveys should be cons ulted for information on current station radiological conditions.

The original design of this radiation shielding was based upon radiation zone criteria whichwere established in support of the expected access requirements and dur ations of occupancy during normal operations and during refueling outages. Descriptions of the zone criteria arepresented in Table12.1-1, and the detailed ra diation zone criteria fo r normal and shutdown '

operations are illustrated on Figures12.1-1 through 12.1-5. Thes e 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 ofdesign dose rates are upper limits and are based on conservative assumptions. Representativeoperating dose rates are expected to be much lower than the design dose rates reported.

Revision 52-09/29/2016 NAPS UFSAR 12.1-2 Occupancy time is such that individual radiati on doses will be within the requirements of10CFR20.Radiation zones are shown on Figure12.1-1 through12.1-5 for the containment building, auxiliary building, fuel building, decontaminat ion building, and waste disposal building. Thezones are defined in Table12.1-1.

The service building and onsite environs are Zone1 throughout. During special operations,local areas within the service building or near the contamin ated storage pad or spent-fuel-cask-handling area may temporarily ex ceed these normal limits; during such times thearea will be defined in accordance with health physics procedures.

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

Maximum accident doses shall not exceed the following:12.1.2Design Description Building arrangements and machine location drawings of Units1 and2 structures, showingplan and sectional views, are given in Section1.2.2. The plot plan and site plan are shown onReference Drawings3 and4.

12.1.2.1Primary Shielding Primary shielding is provided to limit radiation emanati ng from the reactor vessel. Such radiation consists of neutrons diffusing from the core, prompt fission gammas, fission product Accident or Case Control Room exclusion area boundary (EAB)

& low population zone (LPZ)

Design Basis Loss-of-Coolant Accident (LOCA)5 rem TEDE25 rem TEDE Steam Generator Tube Rupture Fuel Damage or Pre-accident Spike5 rem TEDE25 rem TEDE Coincident Iodine Spike5 rem TEDE2.5 rem TEDE Main Steam Line Break Fuel Damage or Pre-accident Spike5 rem TEDE25 rem TEDE Coincident Iodine Spike5 rem TEDE2.5 rem TEDE Locked Rotor Accident5 rem TEDE2.5 rem TEDE Rod Ejection Accident5 rem TEDE6.3 rem TEDE Fuel Handling Accident5 rem TEDE6.3 rem TEDE Revision 52-09/29/2016 NAPS UFSAR 12.1-3 gammas, and gammas resulting from the slowing down and captu re of neutrons. The primary shielding is designed to:1.Attenuate neutron flux to prevent excessive activation of compone nts 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 3feet, and it is surrounded by 4.5feetof reinforced concrete. The shield tank prevents the overheating and deh ydration of the primary shield wall concrete and minimizes the activat ion of the plant compone nts within the reactor containment. A cooling system is provided for the water in the ne utron shield tank. (The neutron shield tank cooling water subsystem is discussed in Section9.2.2.)A 15ft. 8in. high x 2inch thick cylindrical lead shield located beneath the neutron shield tank protects station personnel servicing the neutron detectors during reactor shutdown.Appendix12A contains a detail ed description of supplemen tary neutron shielding. The manway in the upper part of the primary shield is plugge d during reactor operation. The control-rod drive concrete missile shield located above the reac tor vessel is designed to provide some additional neutron shie lding. The primary shield arrangement is shown on Figure12.1-6.

The shield materials and thicknesses are listed in Table12.1-

2. The applicati on of Permalimaterial for supplementary neutron shielding is shown on Figure12.1-7 for Unit1.A 3-1/2inch thick stainless stee l radiation shield is provided at the 12-inch diameter Incore Sump Room drain to protect station personnel during norma l power operation and during refueling outages.

12.1.2.2 Secondary Shielding Secondary shielding consists of the shield ing for the reactor coolant, the reactor containment, fuel handling equipm ent, auxiliary equipment, the wa ste 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 build ing, where N-16 is not the major source ofactivity, activated corrosion a nd fission products from the reactor coolant system control the secondary shielding. Activated co rrosion and fission products in th e reactor coolant system alsoresult in the shutdown radiation levels in the reactor coolant loop areas. Tables11.1-6 and12.1-3 Revision 52-09/29/2016 NAPS UFSAR 12.1-4list the activities used in designing the containment secondary shielding. Table11.1-6 lists the fission product activities and activated corrosion products in the reactor coolant system with 1%failed fuel. Table12.1-3 lists the activated corrosi on product activities and the N-16 activity at thereactor 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 fac ilitates maintenance in all sect ors during shutdo wn. The crane support wall provid es limited access protecti on in the annulus between the crane wall and thereactor containment wall and pr ovides part of the exterior sh ielding required during power operation. Shield walls are provided around each steam generator a bove the operating floor to a height required for personnel protection. Shielding beams be low the operati ng floor are strategically positioned around the steam generators and r eactor coolant pumps. The shielding beams provide protection for pers onnel in the wall annulus from gamma streaming up through the relief openings in the operating floor. The shielding arrangement is shown in Figures12.1-6, 12.1-8, and12.1-9.

12.1.2.4Containment Structure Shielding The containment shielding consists of the steel-lined, steel-reinforced concrete cylinder and hemispherical dome as described in Section3.8.2. This shieldi ng, together with the crane supportwall, attenuates radiation during full-power operation and during the assumed design basis accident to or below design le vels at the outside surface of the containment and at the site boundaries.

12.1.2.5Fuel-Handling Shielding Fuel-handling shielding is desi gned to facilitate the removal and transfer of spent-fuel assemblies from the reactor vessel to the spent-fu el pit. It is designed to protect personnel against the radiation emitted from the spen t-fuel and control-rod assemblies.The refueling cavity above the reactor vessel is flooded to approximately Elevation290 to provide a temporary water shield above the components being wit hdrawn from the reactor vessel.The water height is thus approximately 26feet above the reactor vessel flange. This heightensures approximately 7feet of wa ter above the active portion of a withdrawn fuel assembly at its highest point of travel.

Under these conditions, the dose rate is less than 50mRem/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 52-09/29/2016 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 ar e flooded with water to the same height.

The spent-fuel pit in the fuel building is permanently flooded to provide approximately7feet of water above a fuel asse mbly when it is being withdrawn from the fuel transfer system.Water height above stored fuel assemblies is a minimum of 23feet.

The sides of the spent-fuel pit, three of which also form part of the fuel building exterior walls, are 6-foot-thick concrete toensure a dose rate of no more than 2.5mRem/hr outside the building.Approximately 3feet of concrete shielding is provided above and on each side of the fueltransfer tubes in the area between the reactor contai nment wall and the fuel building wall, and inthe area between the reactor containment wall and the fuel transfer canal.

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

Controlled access to the auxiliary building is allowed during reactor ope ration. Major components of systems are individually shie lded 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 wa ll of the auxiliary building. Each ion exchanger or filter is enclosed in a separate, shielded compartment. The conc rete thicknesses provided around the shieldedcompartments are sufficient to reduce the dos e rate in the surrounding area to less than2.5mRem/hr and the dose rate to any adjacent cubicle to less than 100mRem/hr. The shielding thicknesses around the mixed-bed demineralizers are based upon a saturation activity that gives acontact radiation level of nearly 12,000rem/hr.

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

12.1.2.7Waste Storage Shielding The waste storage and processing facilities in the auxiliary building, decontamination building, and clarifier building are shielded to pr otect operating personnel in accordance with the radiation protection design bases set forth in Section12.1.1.Boron recovery tanks, which are used to stor e letdown before recycling to the station orprocessing as waste, are shielded to reduce dose rates to 2.5mRem/hr in accessible areas. Boricacid storage tanks are located in the auxiliary building so that shielding may be installed if necessary during station operation.

Revision 52-09/29/2016 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 th e tanks is less than0.75mRem/hr.Periodic surveys by Health Physics personnel using portable radi ation detectors ensure that radiation levels outside the shield walls meet design specificat ions, and they establish access limitations within the shielded c ubicles. In addition, continuous su rveillance is provided in thewaste 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 domeare 54inches and 30inches, respectively. These thicknesses are more than adequate to meet theguideline limits of 10CFR50.67 at the exclusion boundary.Additional shielding is provided for the ma in control room. This , together with theshielding afforded by its physical separation from the cont ainment structure, ensures that an operator would be able to remain in the main control room for 30days after an accident and notreceive a dose in excess of 5rem TEDE.

12.1.2.9Boron Recovery Tank ShieldingThe boron recovery tanks (see Section12.1.2.7), are shielded to the height required for personnel protection on the site an d to ensure that the dose rate at the exclusion boundary from direct radiation does not exceed the design dose rates as specified in Table12.1-1.

12.1.2.10Main Control Room ShieldingThe main control room is shown in Figure1.2-3 and on Reference Drawing5.

The design basis for the control room envelope is that the ra diation dose to personnel inside the control room envelope (from sources both intern al and external to the control room envelope)be less than or equal to 5rem TEDE for the 30day duration of the design basis accident. The control room northern, western, and eastern walls are 2' thick conc rete. The southern wall of the control room is 18" thick concrete. The southern wall of the cable vault is 2' thick concrete tobring the total concrete shieldi ng 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 struct ure and can be covered with radiation shielding doors. Based on NUREG-0800, Section6.4 (Reference8), this level of shielding allows the dose in the control room from containmen t shine and cloud shine to betreated as negligible.

Revision 52-09/29/2016 NAPS UFSAR 12.1-7 Special consideration has been given to the de sign of penetrations a nd structural details ofthe main control room to establish an acceptable condition of leaktightness.

The air conditioning systems are installed within the spaces served and designed to provideuninterrupted service under accident conditions. On an emergency signal, the control room normal replenishment air and exhaust systems are is olated automatically by tight closures in the ductwork. Breathing-quality air is discharged from high-pr essure storage bottles to the MCR/ESGR envelope. The MCR/ESGR envelope is also provided with an emergency ventilation system fitted with particulate an d impregnated charcoal filters to introduce cleaned outside air into the protected spaces within an hour after an accident. This can continue indefinitely to supplybreathable quality air to the MCR/ESGR envelope. Fan/filter units also start in recirculationduring 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 mon itors 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 theStone & Webster Engineering Corporation GAMT RAN1 computer code with inputs from theACTIVITY-2 and RADIOISOTOPIC computer codes. The NRC-specified source terms were used. All systems designed to func tion after an accident were c onsidered as sources, including safety injection, recirculation spray, hydrogen recombiner, samp ling, 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 situ ation would be unacceptable. All vital areas were identified an d evaluated. Areas where continuous occupancy is required are the main control room, the technical support center, the c ounting room, the operational supportcenter, and the security control center. Limited access is needed to such places as emergency power supplies and sampling stations.All the NUREG-0578 CategoryA requirements have been satisfied at NorthAnna Units1and2, as indicated by letter, A. Schwencer, NRC, to J. H. Ferguson, VEPCO, datedApril23,1980.12.1.3Source 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 Chapter11. Design and expected values of the radioisotopic i nventory for both the reactor coolant and main steam systems arelisted in Section11.1. Design and expected values of the radioiso topic inventory for each portionof the radioactive liquid waste system are listed in Section11.2.5 and for the waste gas decay tankin the gaseous waste disposal system in Section11.3.5.

Revision 52-09/29/2016 NAPS UFSAR 12.1-8Table11.1-11 lists the activities in the volume control tank using the assumptionssummarized in Table11.1-5. The activities in the pressurizer (both the liquid and vapor phases)are given in Table11.1-13 using the assumptions summarized in Table11.1-5. Saturation activities for demineralizer resins are listed in Table11.1-13.

Spent-fuel activities are listed inTable11.1-4.

Process piping designated to carry significant amounts of ra dioactive materials is located behind shielding to minimize the radiation exposure of plant pers onnel. Pipe tunnels, chases, or shafts are provided as required to properly segr egate radioactive piping behind shields. Wherenecessary, 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, silic one alkyd, or urethaneenamel protective coatings or approved equal. Stainless steel surfaces are not painted. Stainless steel is used extensively in the fuel, dec ontamination, 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 us ing the following factors:1.These tanks are vertical cyli ndrical 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 st eel construction to mi nimize corrosion and th e 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 equipmen t cubicles and pipe chases containing radioactive flui d system components and equipmen t 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 insuspension in the fluid stream. The sample lines to the centr al sample points connect to recirculation lines to permit multivolume flushes of sample lines so that representative samplesare drawn. Local check samples are available from the recirculation lines if needed.

Revision 52-09/29/2016 NAPS UFSAR 12.1-912.1.4Area Monitoring 12.1.4.1 Normal Plant Operations The area radiation monitoring syst em reads out and records the radiation levels in selected areas throughout the sta tion, and alarms (audibly and visual ly) if these levels exceed a preset value or if the detector malfunctions. Each detector reads out and alarms both in the main controlroom and locally. Each channel is equipped with a check source remote ly operated from the main control room. Recorders produce a continuous, permanent record of radiation levels while the detectors are functioning. Area-ra diation-monitoring channels for Unit1 are powered from the480V emergency bus1H; channel monitoring system s or areas common to both units are poweredfrom the emergency bus for either Unit1 or Unit2.

The area radiation monitors are designed fo r continuous operation. C ontinuous, 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 monitoringmethods are in place.The monitor locations, shown on Reference Drawings1, 2, and6, give an early warning ofhigh radiation levels when plant personnel enter various portions of the plant. To perform this function they are generally locate d near the main entrance pathwa y for a given building or portion thereof. In some areas they are located at the major work area i nvolved. In all cases they provide a representative indication of the ra diation level in that vicinity of the plant and not necessarily the maximum that might be measured against one of the nearby sh ield 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 ma in control room indicating the radiation levels throughout the plant, and they may be monitored before entry into potentially high radiation fields. When radioactive mate rial is being handled within a given area, such as the decontamination building, the moni tors 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 containment's purge air supply and exhaust fan and closes the purgesystem butterfly valves, thus isolating the containment from the environment.The alarm setpoint of each area m onitor is variable, and it is se t at a radiation level slightly above that of normal b ackground radiation in the respectiv e area. The monito ring equipment consists of fixed-position gamma detectors and a ssociated electronic equipment. These channels warn of any increase in radiation level at locations where person nel may be expected to remainfor extended periods of time. The instruments and their ranges and lo cations are listed inTable12.1-4.

Revision 52-09/29/2016 NAPS UFSAR 12.1-10Tests and calibrations of the radiation monitors are performed at intervals specified in theapplicable 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 a nd is part of a redundant alarm system with the containment gase ous and particulate monitors. If the manipulator crane monitor is not functional, the cont ainment gaseous and particulate monitors can still function and can be backed up by local portable equipm ent. 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 fu el 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.

12.1.4.2Post-Accident Conditions The containment high-range radi ation monitoring system (CHRRMS) provides indication in the control room of cont ainment radiation level as required by NUREG-0578, Section2.1.8.b, and subsequent clarification contained in the NRC letter dated October30,1979.

Each containment has two redundant ClassI monitor systems consisting of a high rangedetector (100 - 107R/hr), a cont rol room readout unit and associated interconnecting cable. The detectors are located approximately 155degrees apart for Unit1 and 130degrees apart for Unit2 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 ar e qualified to IEEE-323-1974, IE EE-344-1975 and meet therequirements of Regulatory Guide1.97, proposed Revision2. The high range monitors arepowered from diverse Class1E vital buses. The i ndicators in the contro l room are installed in racks designed per the separati on and seismic requirements of Regulatory Guide1.75, Revision1,and IEEE-344-1975 respectively.The addition of the high-ra nge containment radiation mon itors is for indication purposesonly and does not affect the logic sche mes of any safety-related systems.The Technical Support Center (TSC) and Local Emergency Opera tions Facility (LEOF)radiation monitoring systems are localized systems and satisfies the gu idelines established in NUREG-0696. The radiation monitoring system components consist of a particulate, iodine, and noble gas monitor and two area monitors.

Revision 52-09/29/2016 NAPS UFSAR 12.1-11 These monitoring systems provi de continuous indi cation of the dose rate and airborneactivity in the TSC and LEOF during an emergency, as well as alerting personnel of adverse conditions. These systems are totally contained within the TSC and LEOF and are in no wayconnected to the control room or any safety-related systems.12.1.5Operating Procedures A radiation protection program consistent with the requirements of 10CFR20 and designed to ensure that doses are kept ALARA is maintained. Appli cable HP procedures,(i.e.,RWPs), are used to control access to all radiation and contaminated areas.The station auxiliary system s containing radioactive flui ds are designed for remote operation by the use of extensive instrumentation for monitoring, remotely operated pneumatic orelectrical 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 t ools and processe s are describedin Section9.1.4.

The operation of the filter tran sfer shield, which is used fo r the handling of spent filter cartridges, is described in Section11.5.3. 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 prot ects personnel during the servicing of the neutron detectors. This shield is described in Section12.1.2.1.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 equippe d with radiation-monitoring devices. A person entering an area in which the radiation is greate r than a predetermined level isaccompanied by, or is in constant communication with, at least one other person.12.1.6Dose Rate CalculationsTo indicate the methods used to determine dose rates, two sets of ca lculations are describedbelow.

12.1.6.1Sample Sink Area The receptors for the sampling sink are located just off the surface of the concrete wallbehind the sinks. Two sources of radi ation are considered to be significant in this area: the sample piping, located in a pipe space be hind the wall at which the samp ling sinks are located; and the volume control tanks, located in individual cubicles behind the pipe space, as shown inFigure12.1-2 Sh.3.

Revision 52-09/29/2016 NAPS UFSAR 12.1-12 The volume control tanks are separated by a 2-foot-thick concrete wall. Concrete density ofthis and other concrete walls is 146lb/ft

3. On the sampling sink side of the volume control tank,the cubicle wall is 2.5-foot-thick concrete. The distance from the axial centerline of a volumecontrol tank to the surface of the sampling sink wall is approximately 18.5feet.Each volume control tank wa s approximated as a source by two right circular cylinders84inch in diameter with 0.25-inch st eel walls, with liquid volume of 120ft 3 and gaseous volumeof 180ft 3.The sample piping primarily c onsists of 3/4-inch or smal ler tubing cont aining process fluids. The piping is located behind an 18-inch c oncrete wall. For the purpose of this analysis, the maze of pipes was approximated by four disks si de-by-side along the wall behind the samplingsinks, each 0.75inch thick and 6feet in diameter.

Each disk was assumed to be covered by a steel plate of minimal thickness to re present 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 proc ess streams, th e conservative assumption was made that all pi pes contained primary coolant sa mples drawn from the hot leg of the coolant loop. Primary coolant activities are listed in Table11.1-6.

The computer code GAMTRAN described in Section12.1.6.3 was used to calculate thedose 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 inte rsecting the cylindrical axis of one of the volume contro l tanks, the dose rate was calculated to be 4.1mRem/hr. Of the total, the sample piping contributed approximately 97%.

12.1.6.2Valve-Operating Area Outs ide Demineralizer Cubicle In the valve-operating area outsi de the demineralizer cubicle on the 244-foot level of the auxiliary building, typical receptor locations were chosen at 3- and 6-foot heights above the244-foot level, lying on a plane perpendicular to the vertical shield wall, passing through thecylindrical axis of the mixed-bed demineralizer, and at the outside surface of the shield wall.The mixed-bed demineralizer was chosen as th e source because it is the most radioactive source in the area and because the concrete shie lding between the mixed-bed demineralizer andthe receptors is the same thickness as that between other demineralizers.

The mixed-bed demineralizer is assumed to be a right circular cyli nder source inside a5/16-inch mild steel shield with source strengths based on Surry Power Station source datacorrected to NorthAnna power level.The volume of the demineralizer resin is assumed to be 39ft 3 with a height of 7.13feet.

Revision 52-09/29/2016 NAPS UFSAR 12.1-13A 2-foot-thick concrete wall extends vertically from Elevation244 to the floor below thedemineralizer cubicle. Above the floor, the wall is 4-foot-thick conc rete. The floor of thedemineralizer cubicle is 2-foot-thick concrete. Concrete density in all cases is taken as 146lb/ft 3.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 1mRem/hr from themixed-bed demineralizer.

12.1.6.3GAMTRAN Computer CodeThe GAMTRAN code is a Stone & Webster devel oped point kernel code for shield designanalysis. The gamma ray attenuation coefficients used in GAMTRAN ar e generated using theOGRE (Reference1) pair producti on and photoelectric cross sec tions. 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 ofBerger (Reference2) and Chilton (Reference3). The parameters used for the buildup factors arebased on data from the Weapons Radiation Shielding Handbook (Reference4). Flux-to-dose conversion factors were based on curves in the Reactor Shield Design Manual (Reference5).12.1.7Estimates of Exposure Radiation shielding is provide d on the basis of maximum concentrations of radioactive materials within each shielded re gion (e.g., 1% failed fuel) rather than the annual average values.

For batch processes, as an exam ple, the point of the highest ra dionuclide 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 ra ther than average sources to be shielded.The design objectives of the plant shielding for normal operation in terms of maximum doserates allowed at in-plant locations are given in Table12.1-1. 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 GAMTRANdescribed in Section12.1.6.3. Table 12.1-5 lists the densities of the materials used for shielding calculations. Care was taken to ensure that the material actu ally used for constr uction was at leastas dense as that used for analyses. Figures12.1-6, 12.1-8, and12.1-9 show the shieldingarrangement for the containment. Arrangements for the other buildings are shown in Section1.2.

Supplementary neutron shielding is discussed in detail in Appendix12A.

12.1.7.1Considerations 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 100mrem/week to arrive at a design value in terms of mRem/hr that will not be exceeded in that zone, ev en under worst-case conditions. The Revision 52-09/29/2016 NAPS UFSAR 12.1-14 shielding is then designed assuming maximum condi tions to ensure that th ese exposure values arenever exceeded under normal operating conditions. (Higher doses may result from specific repair jobs when shielding is not possible.)The radiation zone designations are shown in Figures12.1-1 through12.1-5. These delineate the maximu m dose rates at all loca tions within the major buildings of NorthAnnaUnits1 and2.

Because of the conservatism em ployed in performing the worst-case dose rate calculations, the shielding is conserva tively designed, thus ensuring that the average exposures in each zonewill 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 spen t in each zone in the plantsuch that the sum of these times equals the total time the em ployee is at the station in anaverage year.2.An "average employee" con cept would not apply because so me employees never go in some zones, whereas others frequent ly 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 modelsand 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% e xpected) would have to be eliminated to predict expected dose rates.7.Explicit margins in other source terms woul d 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 component s would have to be developed to replace the few region models which are pres ently used to intentionally overestimate the emanation of radiation from these large sources.

Revision 52-09/29/2016 NAPS UFSAR 12.1-15 A man-rem analysis cannot be computed with sufficient accuracy to obtain good data of apredictive nature. However, suff icient operating data on simila r plants do exist to provideestimates of man-rem doses for the station as a w hole. This operating experi ence is demonstrative of the fact that the radiation sh ielding is conservatively designed.

This is a direct result of the design of shields for worst-case conditions, conser vative dose rate calculat ions, and implicit andexplicit designer's margins.

12.1.7.2Reports From Other Plants Relative to the estimations of exposure levels during maintena nce, refueling, and inservice inspection activities, such estimate s do not lend themselves to prediction analysis based on an analytical modeling. Reliance should be placed on operating experience at other st ations as the most reasonable source of such data. In this connection, VEPCO's engineer s participated in theefforts of the Atomic Industrial Forum's Task Force on Occupational Exposures.

One survey reported by Charlesworth (Reference6) at the April1971 American Power Conference covered data obtained at seven operating water-cooled reactor plants with a total plantworker dose of 1700man-rem duri ng the previous year for an average of 244man-rem/yr perplant. In this survey, it was found that on an average 75% of th ese exposures were estimated to have been received during shutdown operations.Another survey by Goldman (Reference7) summarizes the results of 27plant-years of operation from operating reports. Th is survey indicated a range of 0.5 to 2.3rem/yr with limiteddata on the number distribution of staff in seve ral exposure categories. Fr om these data, Goldmanconcluded that 19plant-years of operating data resulted in an in-pla nt populatio n average of238man-rem per plant-year. These results are close to the 244man-rem per plant-year reported by Charlesworth.The average dose rate level in the visitor's center will be less than 0.01mRem/hr above natural background based on the worst-case assu mption. Assuming that a visitor will spend4hours at the visitor's center four times per year, he woul d receive a dose of less than0.16mrem/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-stationpersonnel for routine operation are expected to parallel those reporte d from operating plant experience as discussed above.

Extensive radiation shielding is provided based on the max imum concentration of radioactive materials within each shielded region rather than on annual average values. The shielding and occupancy zones fo r normal operation are intentionally very conser vative so that the normally received dose rates should be less than 10% of the limits specified in 10CFR20.

Revision 52-09/29/2016 NAPS UFSAR 12.1-16The highest level of personnel exposure is expected to occur during shutdown andmaintenance periods on systems containing items such as coolant purification filters, cleanup andradwaste demineralizers, ion-exchange resins, charcoal adsorber units, and solid-radwaste-handling components.

Since this is the case, the plant shielding and machinery locations have been designed to provide maximum laydown spac e, maximum working room, andminimum time required to perform operations consistent with the reasonable operation of the plant. Experience gained in the operation of nuc lear plants has been factored into these designs with the objective of minimizing the tota l man-rem exposure to plant personnel.

12.1.7.3Dose From Stored Waste For the purpose of a conservative analysis, it is assumed that 1Ci of cobalt-60 equivalent is stored in the low-level contaminated storage area (Reference Drawing4). The dose rates at the various distances, including the site boundary, per curie of cobalt-60 equivale nt, are presented inFigure12.1-10. No credit is taken for the drum shielding and self-shielding of the waste stored outside the building.

12.1.7.4Health Physics Area Dose EvaluationThe Health Physics office, counting room, and monitoring area complex in the servicebuilding is, under normal operating conditions, a continuous acc ess area. The only anticipated radioactive sources in this area are radioactive samples brought in for analysis and radioisotopes used in analytical equipment such as radiatio n monitoring equipment. Therefore, any radiationdoses received while in this area will be controlled by a dherence to standard health physics practices for handling radioactiv e 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 significa nt contribution to the se rvice building dose rate.

Revision 52-09/29/2016 NAPS UFSAR 12.1-17

12.1REFERENCES

1.Oak Ridge National Laboratory, OGRE - General Purpose Monte Carlo Gamma RayTransport Code System, RSIC Code Package CCC-46, Oak Ridge, Tennessee, 1967.2.M. J. Berger, in Proceedings of Shi elding Symposium , U.S. Naval Radiological DefenseLaboratory, 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: Chapter3 -Methods for Calculating Neutron and Gamma Ray Attenuation, DNA-1892-3, DefenseNuclear Agency, Washington, D. C., March1972.5.T. Rockwell, III, ed., Reactor Shield Design Manual, TID-7004, United States AtomicEnergy Commission, March1956.6.D. G. Charlesworth, Water Reactor Plant Contami nation and DecontaminationRequirements , survey conducted by the Subcommittee on Nuclear Systems, ASME ResearchCommittee on Boiler Feedwater Studies, presented at the 33rd Annual Meeting of theAmerican Power Conference, Chicago, April1971.7.M. I. Goldman, Radioactive Waste Management and Radiation Exposure , NuclearTechnology, Vol.14, May1972.8.Standard Review Plan6.4, Control Room Habitability System , 1981.12.1REFERENCE 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 St ation Drawings are controlled by station procedure.

Drawing Number Description1.11715-FK-9BInstrument Piping, Radia tion Monitoring, Sheet 2, Units 1 & 22.11715-FK-9AInstrument Piping, Radia tion Monitoring, Sheet 1, Units 1 & 23.11715-FY-1BSite Plan, Units 1 & 24.11715-FY-1APlot Plan, Units 1 & 25.11715-FE-27BArrangement: Main Control Room, Elevation 276'- 9", Units 1 & 26.11715-FK-9CInstrument Piping, Radia tion Monitoring, Sheet 3, Units 1 & 2 Revision 52-09/29/2016 NAPS UFSAR 12.1-18The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.Table12.1-1RADIATION ZONE CRITERIAZoneAccess Maximum Dose Rate (mRem/hr)TypicalLocations Full-Power OperationIContinuous0.75 Main control room, outside surface of containment, and all turbine plant and administration areasIIPeriodic2.5 Passageways of auxiliary and fuel buildings, in general, and inside reactor containment personnel lockIIILimited15 Outside surface of shielded tank cubiclesIVControlled100 Annulus between crane wall and containment wallVRestrictedOver 100 Inside shielded equipment compartments Hot Shutdown (after 15-min decay)IIILimited15 Reactor containment above operating floor; outside of crane wallVRestrictedOver 100 Inside shielded equipment compartments Cold Shutdown for Maintenance (after 8-hr decay)IIPeriodic2.5 Reactor containment above operating floor and outside of crane wallVRestrictedOver 100 Inside shielded equipment compartments Cold Shutdown for RefuelingIIPeriodic2.5 Reactor containment above operating floor, outside of crane wall, and adjacent to fuel transfer canal near incore instrumentation

devicesVRestrictedOver 100 Inside shielded equipment compartments Surface of water over raised fuel assembly50 Above fuel assembly when over upender or

racks Revision 52-09/29/2016 NAPS UFSAR 12.1-19Table12.1-2CONTAINMENT SHIELDING

SUMMARY

SymbolFigureShield DescriptionMaterial a Thickness (in)A12.1-8Neutron shield tankWater Steel 34 3B12.1-8Primary shieldConcrete5412.1-7Supplementary neutron shield Permali 6E12.1-8Neutron shield tank support Steel Lead 1.5 2F12.1-6 and 12.1-8 Cubicle - crane support wall Concrete 33F12.1-8Shielding beamsConcrete24G12.1-8Crane support wallConcrete24 H12.1-6 and 12.1-8Containment wallConcrete 54I12.1-8Containment domeConcrete30 J12.1-8Floor elevation 243ftConcrete42 - 48 K12.1-8Operating floorConcrete24 L12.1-6 and 12.1-8Refueling cavity wallConcrete 42M12.1-8 and 12.1-9 Control-rod drive missile shield Concrete 24N12.1-8Refueling cavity waterWater108O12.1-8 and 12.1-9 Removable block wall Facing personnel hatch

All others Concrete Concrete 18 12P12.1-6Fuel transfer canal wall (containment structure)

Concrete 54Q12.1-6Fuel transfer canal wall (containment structure)

Concrete 72R12.1-6Fuel transfer tube shielding Concrete 36 (min)S12.1-6Fuel transfer canal wall (fuel building)

Concrete 72T12.1-6Incore instrumentation cubicle wall Concrete 42a.All poured concrete is reinforced with steel.

Revision 52-09/29/2016 NAPS UFSAR 12.1-20U12.1-6Cubicle wallConcrete36V12.1-6Regenerator heat exchanger wall Concrete 24W12.1-6Cable vault wallConcrete24X12.1-6Auxiliary feed pump wall Concrete 36Y12.1-6Safeguards area wallConcrete12 Unit2 only Z12.1-8Incore sump room drainStainless Steel31/2Table12.1-2(continued)CONTAINMENT SHIELDING

SUMMARY

SymbolFigureShield DescriptionMaterial a Thickness (in)a.All poured concrete is reinforced with steel.

Revision 52-09/29/2016 NAPS UFSAR 12.1-21Table12.1-3N-16 AND ACTIVATED CORROSION PRODUCT ACTIVITYIsotopeActivity (µCi/cc@577°F)

Mn-54 5.6x10-4 Mn-56 2.1x10-2 Fe-59 7.5x10-4 Co-58 1.8x10-2 Co-60 5.4x10-4 N-16 a 73.3a.At the reactor vessel outlet nozzle at 2910 MWt.

Revision 52-09/29/2016 NAPS UFSAR 12.1-22Table12.1-4AREA RADIATION MONITORING LOCATIONS, NUMBER AND RANGESChannelLocation(number)

Range (mRem/hr)Reactor containment area - low range (2)

(1/2-RM-RMS-163/263) 10-1-10 4 Personnel hatch area (2)

(1/2-RM-RMS-161/261)10 10 4 Manipulator crane (2)(1/2-RM-RMS-162/262)10 10 4Incore instrumentation transfer area (2)(1/2-RM-RMS-164/264)10 10 4 Decontamination area (1)

(1-RM-RMS-151)10 10 4 New fuel storage area (1)(1-RM-RMS-152)10 10 4 Fuel pit bridge (1)(1-RM-RMS-153)10 10 4 Auxiliary building area (1)

(1-RM-RMS-154)10 10 4Waste solidification area (1)(1-RM-RMS-155)10 10 4 Sample room (1)(1-RM-RMS-156)10 10 4 Main control room (1)

(1-RM-RMS-157)10 10 4 Laboratory (1)(1-RM-RMS-158)10 10 4 Technical Support Center (2)(1-RM-RMS-184/185/186)10 10 4 Local Emergency Operations Facility (2)

(1-RM-RMS-187/188/189)10 10 4 Revision 52-09/29/2016 NAPS UFSAR 12.1-23Table12.1-5MATERIALS USED FOR SOURCE AND DOSE RATE CALCULATIONS Material Density (lb/ft 3)Ilmenite concrete240Ordinary concrete146 Steel490.5 Lead707.6 Air, steam, or vapor0.075

WaterPressurized reactor coolant46 All other62.4Core273.4 Revision 52-09/29/2016 NAPS UFSAR 12.1-24The 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)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-25The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 2 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-26The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 3 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-27The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 4 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-28The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 5 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-29The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 6 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-30The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 7 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-31The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-1(SHEET 8 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016 NAPS UFSAR 12.1-32The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-2(SHEET 1 OF 3)RADIATION ZONES AUXILIARY BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-33The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-2(SHEET 2 OF 3)RADIATION ZONES AUXILIARY BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-34The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-2(SHEET 3 OF 3)RADIATION ZONES AUXILIARY BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-35The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-3(SHEET 1 OF 2)

RADIATION ZONES FUEL BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-36The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-3(SHEET 2 OF 2)

RADIATION ZONES FUEL BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-37The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-4(SHEET 1 OF 2)RADIATION ZONES DECONTAMINATION BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-38The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Figure 12.1-4(SHEET 2 OF 2)RADIATION ZONES WASTE DECONTAMINATION BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-39The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Figure 12.1-5RADIATION ZONES WASTE DISPOSAL BUILDING Revision 52-09/29/2016 NAPS UFSAR 12.1-40 Figure 12.1-6SHIELD ARRANGEMENT-PLAN Revision 52-09/29/2016 NAPS UFSAR 12.1-41 Figure 12.1-7PERMALI LOCATIONS Revision 52-09/29/2016 NAPS UFSAR 12.1-42 Figure 12.1-8 SHIELD ARRANGEMENT ELEVATION Revision 52-09/29/2016 NAPS UFSAR 12.1-43 Figure 12.1-9 SHIELD ARRANGEMENT PLAN OPERATING FLOORSee Appendix12A for a discussion of supplementary neutron shielding.

Revision 52-09/29/2016 NAPS UFSAR 12.1-44 Figure 12.1-10DOSE RATE PER CURIE OF CO-60 EQUIVALENTVS. DISTANCE FROM LOW LEVEL CONTAMINATED STORAGE AREA Revision 52-09/29/2016 NAPS UFSAR 12.2-112.2VENTILATION12.2.1Design Objectives One of the objectives of the ventilation system is to ensure that the airborne radioactivityconcentration in different locations inside th e station buildings duri ng normal operation, includinganticipated operational occurrences, are less than those allowed in Table1, Column3, ofAppendixB of 10CFR20, except in the contai nment structures. Conc entrations in areasaccessible to plant administrative personnel and public visitors areas at the site will be less than 1% of the above.

The design and expected airborne radioactivity levels, including anticipated operationaloccurrences, for different buildings are listed in Table12.2-1. The design and expected annual inhalation dose rates for plant personnel in each building are listed in Section12.2.6.

The calculational methodology used to pe rform the design an d expected airborne radioactivity levels, which are based on the criteria of the old 10CFR20, are valid analyses and do not require recalculation according to the revised 10CFR20 limits.

The containment internal cleanup system described in Section9.4.9 and the high-efficiencyparticulate air (HEPA) and charcoal filters described in Section9.4.8 are not required to reduce the radioiodine in the containmen t to the derived air concentrat ion (DAC) before personnel entry.

Personnel entry will be under admi nistrative control only and will be allowed only in accordance with standard health physics pract ices, factoring in act ivity levels, occupancy times, and approved breathing equipment, as discussed in Sections12.1.5 and12.2.5.12.2.2Design Description Detailed descriptions of ventilation systems for differen t buildings are given in the following sections of this report:SectionSection Title9.4.1Main Control Room and Relay Rooms 9.4.2Auxiliary Building 9.4.3Decontamination and Wast e Solidification Building9.4.4Turbine Building 9.4.5Fuel Building 9.4.6Engineered Safety Features Areas 9.4.7Service Building 9.4.8Auxiliary Building HEPA/Charcoal Filter Loops 9.4.9Containment Structure Revision 52-09/29/2016 NAPS UFSAR 12.2-2 12.2.2.1Auxiliary BuildingThe equilibrium airborne activities in the auxiliary build ing result from the leakage of primary coolant from pump seals and valve stem s and from small, miscellaneous leaks. In addition, a small amount of iodine is released to the auxiliary build ing atmosphere from the sampling sink drains, but this is negligible compared to the ot her assumed leaks. All of the iodines and noble gases associated with these leaks are assumed to be released to the auxiliary buildingair and exhausted through the auxiliary build ing ventilation, which exhausts a minimum of10building volumes per hour.

In the auxiliary building, the primary coolant letdown to the Chemical and Volume Control System passes through a mixed-bed demineralizer with a decont amination factor of 10 for allisotopes except Cs, Mo, Y, and the noble gases, fo r which the decontaminat ion factor is 1, which reduces the ionic activity in the coolant.There is a small potential for leakage upstream of the demineralizer. However, in the analysis, one-third of the leakag e is assumed to occur before the demineralizers; the remaining two-thirds is assumed to occur after the deminerali zers. The release of radioactive material in thisarea is considered unlikely because:1.All the piping is welded.2.All valves are of the diaphragm type, which precludes stem leakage.3.No pumps having seals or other equipment with moving parts that might leak are located in this area.4.Demineralizer and filter vents are contained by a piping system that discharges via a charcoalfilter and radiation monitor.

The radioactive demineralizers ar e all in individual shielding cubicles along the south wall of the auxiliary building. These cubicles are not connected to the ventila tion supply or exhaustsystem (Reference Drawings1 &2). The only air normally passing thro ugh these cubicles is slight leakage past valve stem extension or pipe penetrati on sleeves caused by any minordifference in air pressure between floors of the auxiliary building. Therefore, it is not deemed necessary to provide an exhaust sy stem directly from this area.

12.2.2.2Containment Structure The equilibrium airborne activities in the c ontainment structure have as their source the leakage of primary coolant within the containment for up to 18months prior to purging. No dilution of the containment atmosphere is assumed during the 6-month period before the purge.

Revision 52-09/29/2016 NAPS UFSAR 12.2-3 12.2.2.3Turbine Building Airborne activity ente rs the turbine building atmosphe re via the main steam leakagespecified in Section11.1. The turbin e building ventilation rate is 7 x10 5 scfm and the building volume is 4 x10 6 ft 3.12.2.2.4Fuel BuildingAirborne activity is assumed to occur in the fuel building atmosphere from activity releasedfrom failed fuel assemblies in the spent-fuel pit.

For the design case, one-third of a core from eachunit, operated at 100% power for 3years, 365days/year, with 1% fail ed fuel, is assumed to be in the spent-fuel pit. For the exp ected case, one-third of a core from each unit, operated at 100%power for 3years, 300days/year, with 0.2% failed fuel, is assumed to be in the spent-fuel pit.

The fuel in the spent-fuel pit is assumed to have decayed for 100hours, the minimum time before fuel can be transferred from the core to the spent-fuel pit.Escape rate coefficients for both design and expected cases for th e failed fuel in thespent-fuel pit are assumed to be 10

-5 of the escape rate coefficients of the failed fuel in the core,which are listed in Table11.1-5.The spent-fuel pool is assumed to have an effective decont amination factor of 200 for iodines, the same decontaminati on factor used in the analysis of the fuel-handling accident inSection15.4.5.

The fuel building has a vent ilation exhaust rate of 35,000scfm and a volume of 160,000ft 3.12.2.3Source TermsThe activities listed in Table 12.2-1 are based on failed fuel and leakage assumptions givenin Section11.1 and the additional assumptions given in Section12.2.2.12.2.4Airborne Radioactivity Monitoring Radioactivity may become airborne through opera tions such as the welding or grinding of acontaminated componen t, the decontamination of such components, leak age from a systemcontaining radioactive fluids or gases, or the disturbance of the deposited activity in various areas of the plant. An airborne samp ling location is selected on the basis of the potential for airborneactivity within the work area as determined by engineering evaluation.

This system is capable of mo nitoring any of eight possib le ventilation paths but can be programmed as to the sequence a nd duration of monitoring. Seven of these sample points lie in probable maintenance or fuel-handling areas. Th e eighth sample point is a spare. The pointssampled are (1)the fuel building, (2)the safeguards area of Unit1, (3)the safeguards area of Unit2, (4)the central area of the auxiliary building, (5)the general area of the au xiliary building,(6)the containment purge, and (7)the decontamin ation building. The ventil ation vent multi-port Revision 52-09/29/2016 NAPS UFSAR 12.2-4 sampler particulate monitor and the ventilation ve nt sample gas monitor which are described inSection11.4.2.6 has a manual override which allows the continuous sampling of a chosen area.

The containment gas and particulate monitors (Sections11.4.2.17 and11.4.2.18) sample from the containment recirculation duct.

In the event that concurrent operations are being performed in different work areas, the multisample particulate monitor can be placed on manual and alternated at selected intervals between the work areas. Additionally, process radiation monitors continuously monitor selected ventilation lines containing or possibly containing radioactivity.

Each monitor has a readout withan audible/visual alarm in the main control room. Local audible and visual alarms for the processand ventilation vents are provided by the post-accident radiation normal range monitors. The multisample monitor does not have a local readout and alarm. The above system can be supplemented with a portable moving or fixed filter paper continuous monitoring unit to provideadditional monitoring for major maintenance, with a potential for high airborne radioactivity.

Such equipment would be calibrated and operated in accordance with established procedures.

Low-volume air samplers are fixed filter (either paper, glass fiber, or charcoal cartridge, or acombination of these) vacuum pump-type samplers. High-volume air samplers are fixed filter, generally paper or cloth.

When either of the above sample rs is used, it is op erated for a known am ount of time at a known flow rate. The filters are removed for counting with appr opriate instruments. Depending on the analysis desired, filters can be counted fo r beta-gamma, alpha, iodi nes, or gamma isotopic.

Theÿ concentrations are then calc ulated from these data. If requi red, portable counting equipment (beta-gamma or gross gamma) is available for counting filters at or near the location of the airsampler.For the conditions given above, other than rou tine surveys, if personnel duties in the area are of a routine or fixed nature and other indicators (i.e., re lated systems level or pressure indicators, the radiation monito ring system, etc.) show no abnormal conditions , the samplers will be continuously operated and the filters changed and counted r outinely at varying intervals.

On occasions when it is expected that conditi ons could change rapidly or vary considerably, the filters will be changed and count ed routinely at varying intervals.The air-sampling program is in addition to or supplements a ny protective equipment that isauthorized or required by 10CFR20.The sensitivity of the particulate monitor is such that th e monitor can detect airborne particulate levels as low as one-third of the permissible 10CFR20 values. Because theparticulates are collected on a movi ng filter tape, equilibrium is essentially reached in a collectiontime of 5hours.

Revision 52-09/29/2016 NAPS UFSAR 12.2-5 The sensitivity of the gas monitor is such that the permissible 10CFR20 values for Xe-133and one-tenth the permissible 10CFR20 values for Kr-85 are detectable.

Sampling time is not significant.

The total general area ventilation system flow rate is 74,100cfm. The lowest exhaust flow rate from any building area that exhausts to the general area ventilation system and that is normally occupied by operating personnel is 12,400cfm. Airborne con centrations in this area are therefore diluted by a factor of approximately six between the point of intake and the sampling point. The sensiti vity of the monitors is such that as low as six-tenths of the permissible10CFR20 level for Kr-85 and I-131 is detectab le by the ventilation vent sample gas and particulate monitors. The central air ventilation system flow rate is 60,600cfm. This system exhausts air from cubicles not normally occupied by operating personnel. The lowest rate of exhaust flow from an area that exhausts to the central area ventilation system is 150cfm. Thisresults in a dilution factor of approximately 400. Airborne activity levels above 10CFR20permissible levels may not be detectable in the cubicles by the ventilation vent sample monitor.However, airborne levels throughout the auxiliary building, including the cubicles, are monitored as part of the routine health physics surveys as described in Section12.3.1. The portable monitoring equipment used in these surveys is described above.

The primary function of the centr al area ventilation vent samp le is to warn of abnormalreleases indicative of gross equipment malfunctio

n. In addition, the poss ible radiation sources within the cubicle areas are li mited by design, as discussed in 12.2.2.1. Therefore, the ventilationvent sample monitor, in conj unction with the routine health physics airborne sampling program,provides adequate protection for operating personnel.

Background radiation leve ls and other factors that affect the sensitivity were difficult to quantify until after the station was in operation. To minimize the backgr ound contribution, the monitors were located on the upper level of the a uxiliary building where the radiation levels were expected to be the lowest. Lead shielding redu ces the background radiation to a level that does not interfere with the detector sensitivity. Stainless steel sample lines minimize deposition and plateout losses.

The post-accident air monitoring may be performed with portable air samplers, and in compliance with the TMI-2 Le ssons Learned requirements. Ca rtridges are removed and counted in the shielded counting room with a multichannel analyzer. To reduce noble gas interference, silver zeolite cartridges have been obtained. To ensure the timely analysis of the cartridges in anemergency, several multi-channe l analyzers are availa ble for use in air monitoring. The required procedures are in effect. Thus, the capability exists for accurately m onitoring iodine in the presence of noble gases.To comply with the NRC's directive to provide the ability to monitor the post-accident release of potentially hi gh levels of radioactivity via the ventilation system, as expressed in Revision 52-09/29/2016 NAPS UFSAR 12.2-6 NUREG-0578 and clarified in NUREG-0737, high-range effluent m onitors have been installed in various release paths of the plant. They are described in Section11.4.3.12.2.5Operating Procedures Air sampling and bioassays are us ed to identify hazards, to evaluate individual exposures,and to assess protection afforded.

When the use of respirators is considered necessary, their use is in accordance with written proce dures for personnel training and fo r the selection, fitting, testing, and maintenance of the equipment.Respiratory equipment approved by the National Institute for Occupational Safety and Health/Mine Safety and Health Administration (NIOSH/MSHA) is use

d. Equipment not tested and certified by NIOSH/MSHA requires an authorization a nd exemption be approved by the USNRC before use.

Authorization has been received to use MSA Model401 (brass or aluminum parts),Ultralite, and Custom4500 Dual-Purpose SCBA charged with 35% oxygen and 65% nitrogen.

All units are to be equipped w ith silicone face-pieces. Regulator use is not to be initiated attemperatures greater than 135°F. Units may be us ed in areas where temperatures exceed 135°F if regulator use is initiated prior to entry into the areas. Authorization has been received to use MSA Model Firehawk M7 SCBA char ged with 35% oxygen and 65%

nitrogen. All units are to beequipped with rubber face-pi eces. Breathing gas quality and composition, including hydrocarbonexclusion, are ensured by strict controls and main tained in accordance with the latest revision of Compressed Gas Association (CGA) specification4.3, GradeE for Oxygen and CGAspecification10.1, GradeB for Nitrogen.

12.2.5.1 Filter ChangesBefore a filter change, all filter casings are isolated to prev ent the flow of air through thecontaminated filters. Fi lters are removed from th eir frames and placed directly into a plastic bag.All filter assemblies are pr ovided with adequate worki ng space to permit two men toreplace the filters. To facilitate filter handling, no bank is more than three filter units high.

12.2.5.2Temporary Air DuctingIn the reactor containment, connections for flexible duct, from the discharge side of portable ventilation units, are provided at the lower level in the ventilation purge exhaust duct to allow removal of radioactive gases from the steam generato rs or other area s of maintenance.

These connections are capped during normal containment operation and the caps are removed when necessary to connect flexible duct.

In the decontamination building spent-fuel cask area, a flexible hose connection is permanently installed on the exha ust duct to permit the removal of airborne radioactivity during Revision 52-09/29/2016 NAPS UFSAR 12.2-7 maintenance and repair activities. The hot laboratory in the service building has a permanentflexible hose for use in capturing airborne radioactivity.12.2.6Estimates of Inhalation Doses The design and expected inhalation dose rates within the following areas are negligible.

The calculational methodology used to perform the estimated annual inhalation dosesreported in Table12.2-2 is base d on the criteria of the old 10CFR20. These analyses remainvalid and do not require recalculation according to the revised 10CFR20 criteria.1.Main control room and relay room.2.Decontamination building.

3.Engineered safety features area.4.Service building.

Estimates of inhalation doses to plant personnel in the c ontainment structure, turbine building, auxiliary building, and fuel building are listed in Table12.2-2. Airborne concentrations used for inhalation dose estimates ar e based on the following assumptions:1.Containment structureEntry to the containment structure can and will be made during power operation; however, ifduring such entries, levels of airborne radioactivity significant to inhalation doseaccumulation were present, suitable protective air-breathin g equipment normally would beused. After plant shutdown and containment purge, as done in preparation for refueling operations, there would be no signi ficant levels of airborne radioactivity in the containment.However, for conservatism in calculating inhalation doses attributable to containment entry,the following was assumed:a.Iodine-131 in the containmen t at the maximum permissible concentration before entry.b.52hours/year occupancy factor.c.No protective air-breathing equipment.2.Turbine buildinga.0.2% failed fuel.b.20gallons/day per unit primary system to secondary system leak rate.c.1.2x10 7lb/hr per unit steam flow.d.22gpm per unit steam generator blowdown.e.10lb/hr per unit main steam leakage into the turbine building.

Revision 52-09/29/2016 NAPS UFSAR 12.2-8f.0.1 partition factor for iodines from liquid to steam in the steam generator.g.4.0x10 6 ft 3 per unit free volume of the turbine building.h.No credit taken for plateout or dec ontamination inside the turbine building.i.700,000scfm per unit ventilation rate.j.750hours/year occupancy factor.3.Auxiliary buildinga.0.2% failed fuel.b.0.003gpm per unit (at 120°F) tota l primary system to auxiliar y building leakage, divided as follows:1)50% from sampling purges, wi th a partition factor of 10 3 for iodines released to the building atmosphere.2)16.7% upstream from the mixed-bed demineralizers, with a partition fact or of 10 for iodines released to the building atmosphere.3)33.3% downstream from the mixed-bed deminer alizers, with a de contamination factor of 10 and a partition factor of 10 3 for iodines released to the building atmosphere.c.8.1x10 5 ft 3 free volume of the auxiliary building.d.750hours/year occupancy factor.4.Fuel buildinga.0.2% failed fuel.b.2900MWt per unit reactor power.c.Stored spent fuel has been in the reactor for 3years of power operation.d.Average thermal neutron flux in the reactor core of 5.45 x10 13/cm 2-sec.e.157 fuel assemblies per core.f.One-third of a core from e ach unit in the spent-fuel pit in the fuel bui lding (105 fuelassemblies).g.A decontamination factor of 100 fo r iodine in the spent-fuel pit.h.Escape rate coefficients for the spent-fuel pit of 6.5 x10-13sec-1 for noble gases and 1.3x10-13sec-1 for iodines.i.1.85x10 5 ft 3 free volume of the fuel building.j.3.5x10 4 scfm ventilation rate.

Revision 52-09/29/2016 NAPS UFSAR 12.2-9k.250hours/year occupancy factor.

The above occupancy factors are based on operating data from the Connecticut YankeeAtomic Power Plant.

The inhalation dose is then calculated by the following method:

x

12.2REFERENCES

1.Letter from N. Kalyanam, NRC, to J. P. O'Hanlon, Virginia Power, July31,1998,NorthAnna Power Station, Units1 and2 - Exemption from 10CFR20.1703(a)(1),10CFR20.1703(c), and 10CFR20, AppendixA, Protection Factors for Respirators,Footnoted.2(d), and Authorizati on to Use Certain Respirators for Worker Protection InsideContainment (Tac Nos.M98384 andM98385), Serial No.98-473.2.Letter from Karen Cotton, NRC, to David A. Heacock, Virginia Electric Power Company,May28,2010, NorthAnna Power Station, Unit Nos. 1 and 2 and Surry Power Station, Unit Nos. 1 and 2, Exemption From Certain Requirements of 10CFRPart20 (TAC Nos.ME2835, ME2836, ME2828 and ME2829), Serial No.10-363.12.2REFERENCE DRAWINGSThe 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 St ation Drawings are controlled by station procedure.

Drawing Number Description1.11715-FM-2AArrangement: Auxiliary Building, Plan, Elevation 244'- 6"2.11715-FM-2FArrangement: Auxiliary Building; Sections 3-3, 4-4, & 5-5Direm ()Occupancy Factor (hr)Airborne Concentration (µCi/cc)xMPC iµCi/cc ()-------------------------------------------------------------------------------------------------------------------------------

---

-=30 Rem yr----------

-1 yr 2000 hr------------------

x Revision 52-09/29/2016 NAPS UFSAR 12.2-10Table12.2-1 EQUILIBRIUM ACTIVITIES IN DIFFERENT PLANT BUILDINGS (CI/CM 3)Auxiliary BuildingTurbine BuildingContainment StructureFuel BuildingIsotopeDesignExpectedDesignExpectedDesignExpectedDesignExpectedKr-85m1.3x10-08 1.3x10-09----1.4x10-06 1.5x10-07 1.2x10-15 2.3x10-16Kr-853.1x10-08 3.1x10-09----2.5x10-03 2.0x10-04 2.9x10-10 4.9x10-11Kr-877.1x10-09 7.1x10-10----2.5x10-07 2.5x10-08----Kr-882.2x10-08 2.2x10-09----1.6x10-06 1.6x10-07 4.0x10-19 7.9x10-20Xe-131m1.5 x10-12 1.5x10-13----7.4x10-05 7.4x10-06 3.5x10-09 7.0x10-10Xe-133m1.9 x10-08 1.9x10-09----2.7x10-05 2.7x10-06 4.9x10-10 9.7x10-11Xe-1331.7x10-06 1.7x10-07----5.7x10-03 5.7x10-04 3.0x10-08 6.1x10-10Xe-135m9.1 x10-10 9.1x10-11----6.8x10-07 6.8x10-08 3.3x10-13 6.5x10-14Xe-1353.7x10-08 3.7x10-09----1.1x10-05 1.1x10-06 5.3x10-11 1.1x10-11Xe-1383.3x10-09 3.3x10-10----3.0x10-08 3.0x10-09----I-1313.0x10-09 3.0x10-10 2.1x10-11 1.4x10-12 2.2x10-06 2.0x10-07 2.7x10-11 5.4x10-12I-1321.1x10-09 1.1x10-10 3.0x10-12 1.7x10-13 3.9x10-07 3.7x10-08 2.3x10-11 4.7x10-12I-1334.9x10-09 4.9x10-10 2.3x10-11 1.3x10-12 2.9x10-06 2.7x10-07 3.1x10-12 6.2x10-13I-1346.3x10-10 6.3x10-11 3.4x10-13 1.4x10-14 6.8x10-08 6.7x10-09----I-1352.6x10-09 2.6x10-10 7.1x10-12 3.3x10-13 1.1x10-06 9.9x10-08 2.5x10-15 5.1x10-16µ Revision 52-09/29/2016 NAPS UFSAR 12.2-11Table12.2-2 ESTIMATE OF ANNUAL INHALATION DOSES TO PLANT PERSONNEL aLocationEstimated Annual Dose (rem)Containment structure, Unit10.78Containment structure, Unit2 0.78 Turbine building 0.0023 Auxiliary building 0.060 Fuel building 0.0024 ba.Personnel whose work areas are normally in the locations designated above. Other plant personnel, such as administrative personnel, are expe cted to receive a small fraction of the doses listed above, if they receive any inhalation dose at all.b.The impact of discharging a full core from each unit would be to increase the estimated annual dose received in the fuel building by a factor of three.

Revision 52-09/29/2016 NAPS UFSAR 12.2-12 Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 12.3-112.3HEALTH PHYSICS PROGRAM12.3.1Program Objectives and Procedures The Radiological Protection pr ogram provides the guidance a nd technical support required with the handling and evaluation of radiological hazards associated with the operation and maintenance of the station. The administration of the program is the responsibility of the Manager Radiological Protection.The Radiological Protection program consist of administra tive and technical proceduresand other associated Health P hysics documents. This program a nd its revisions are approved bythe Facility Safety Review Committee and is available for onsite review by the NRC. Each station employee receives training in basi c radiation protection as described in Section13.2. A RadiationWork Permit system is included in the Radia tion Protection program and is described in theapplicable Health Physics proce dures. Protective clothing and ot her requirements are listed on or referenced by the permit.Operating guidelines and rules to ensure that Total Effective Dose Equivalent (TEDE) willbe ALARA during operation an d maintenance are provided in the Radiological Protectionprogram. Each station employee will be oriented as to its contents and us ually quizzed to ensure his/her competence. Indivi duals deliberately violat ing procedures set forth in the program will be subject to administrative action.Periodic radiation and conta mination surveys by health physics personnel ensure that current radiological conditions are known. Results of these survey s are posted at the entrance tothe radiological control area, the station's main health physics control point. Station personnel therefore have access to information regarding current radiological conditions in the area they intend to visit.Station personnel will be issued dosimetry e quipment, including indi cating dosimeters, foractivities within the radiological controlled areas. A system has been devised whereby theindividual's accumulated exposure, after performing a job within the radiol ogical control areas, is logged, thus allowing Health Physics to estimate his total exposure for the current month. If anindividual's dose is excessively higher than others in his section for the same time span, Health Physics will inform his/her supervisor and request that another person be assigned the requiredtask. Estimates of work completion time will be made, and the use of stay-time and the rotation of individuals will minimize exposure.

Personnel doses will be limited to 10CFR20.1201 limits. Admi nistrative controls will be implemented to assure personnel doses do not exceed 10CFR20.1201 limits.

The routine monitoring program consists of air samples; contamination surveys (smears);

gamma, beta-gamma, or neutron surveys; and bot h general area and contac t dose rate readings.

Revision 52-09/29/2016 NAPS UFSAR 12.3-2 The In-Plant Radiation Monitoring Program ensures the capability to accurately determine the airborne iodine concentrat ion in vital areas under accident conditions. Th is program includes(1)training of personnel, (2)procedures for monitoring, and (3)provisions for maintenance of sampling and analysis equipment.

Health physics personnel perfor m regular in-plant surveys in all areas where personnel access is required. The frequency depends on the area in question and on current plant conditions, and is defined in the Radiolog ical Protection Program. Appropriate general area readings andsmears are taken, in addition to selected air samples. Other areas of the station are surveyed asappropriate for general area, beta-gamma, contamination, and airborne activity.12.3.2Facilities and Equipment The health physics facility is located in the se rvice building corridor l eading to the auxiliary building and thus is convenient to all personnel entering and exiting the RCA. The facilitiesinclude office space, briefing room, labs, a co unt room, change rooms, dosimetry issue area, instrument issue, laundry area a nd a personnel decontamination ar ea. These facili ties are shownon Reference Drawing1.

Locker rooms are provided for personnel entering the RCA. A change out area is located in the RCA for the donning and storag e of protective clothing. An am ple supply of c overalls, lab coats, hoods, shoe covers, rubber gloves, plas tic suits, etc. are available as required.

The personnel decontamination area is located at the exit to the RCA and is used formonitoring personnel for contamination and performing any decontamination of personnel as required. Showers and sinks are provided to ai d in any personnel decontamination effort.

Fixed and portable instru mentation is available for countin g and/or detecting and indicating radiation levels from al l radiation sources at the station. A sufficient number are on hand to ensurecontinued availability. Calibration/recalibration is performed in accordance with applicable technical procedures.

Respiratory protection devices ar e available to protect personnel from airborne radioactivityand are issued in accordance with the applicable RWP.

Radiation areas are clearly pos ted and warning signs, barric ades and locked doors are usedin accordance with the Radiati on Protection program to protect personnel from inadvertent access to high radiation areas.

Additional shielding material is available as needed and can be used on either a permanent or temporary basis. The material consist of l ead blankets, steel sheets and concrete blocks. Aspecial transfer cask is available for handling hi ghly radioactive filters.

Remote-handling tools areavailable for handling small li ghtweight objects or remotely operating valves or other Revision 52-09/29/2016 NAPS UFSAR 12.3-3 components, while cranes and monorails can afford the distance required for handling heavier objects.Personnel exiting any RCA are m onitored for radioactive cont amination in accordance with the Radiation Protection program.

Additional monitoring is performed for personnel exiting theprimary restricted area.12.3.3Personnel Dosimetry External dosimetry is provided for all personne l who enter any radiological controlled area or radioactive material storage area at the st ation. Thermoluminescent dosimetry (TLD) badges are used to determine lens dose equivalent, shallow dose equivalent, effect ive dose equivalent and deep dose equivalent as required by 10CFR20. Indicating dosimeter s are used to estimate doses in the periods between badge readings. Extremity dosimetry is worn in accordance with theapplicable RWP.

TLD dosimeters will be calibr ated according to methods and standards established by themanufacturer of the equipment and in accordance with applicable technical procedures.

The Bioassay program is in accordance with the requirements of 10CFR20. The Bioassay program quantifies the amount of radioactive material present in workers and converts the results to calculated dose and estimated intakes of radioactive material. The program also offers amethod to aid in evaluating the effectiveness of Station programs to control and minimize airborne radioactive mate rial. Frequencies, procedures and t ypes of analyses are defined in the Radiation Protection program.

Whole-body counts of all station employees are taken as soon as practicable after theirassignment to the station. Nonemployee personnel assigned duties at the station are whole-body counted as required by radiation protection.Standard lab equipment is available to prepare samples as required fo r counting. Distillingapparatus and ion-exchange columns are availabl e for preparing liquids for tritium analysis.12.3REFERENCE DRAWINGSThe 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 St ation Drawings are controlled by station procedure.

Drawing Number Description1.11715-FM-5AArrangement: Service Building, Sheet 1 Revision 52-09/29/2016 NAPS UFSAR 12.3-4 Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 12.4-112.4RADIOACTIVE MATERIALS SAFETY12.4.1Materials Safety Programs Established health physi cs procedures require the notific ation of the Radi ation Protection Department of the arrival of radioactive materials at the st ation. Appropriate surveys and inventory are then taken and the material is taken to a designated area fo r storage and/or use.

High-activity sources, such as reactor start-up sources, are nor mally stored in their shipping containers, in other appropriate c ontainers, or under water until thei r use is required, at which time Health Physics coverage will be provided. Sources such as those required for calibrating high-range gamma survey meters are obtained from manufacturers in shielded devices designedso that the sources cannot be r eadily removed and so that doses to those using the sources can be kept ALARA. Other calibration sources will be stored in locked ar eas and/or shielded containers,and their removal will be by authorized personnel only.

The use of unsealed by-product materi al received at the site is essentially limited to that of health physics or chemistry pers onnel in the preparati on of low-level calibra tion sources for count room equipment. It is not expected that any unsealed, special nucl ear material will be received at the site.The Radiological Protection Pl an requires that no radioac tive material or suspectedradioactive material be carrie d or removed from a restricted area without Health Physics'notification and approval. Within the restricted area, all unattend ed tools, loose components, or equipment containing or contaminated with radioa ctive material must be identified by tagging orplaced behind barriers.Tool kits are available for work in contaminated areas only, thereby eliminating the need totransfer a large number of tool s back and forth between clean and radiological controlled areas.

These tools are periodically checked and decontaminated as required. When special tools are required and used, they must be surveyed by Health Physics before leaving the radiological controlled areas for storage or us e in other areas of the station.Hot storage areas are provided to contain and control radioactive mate rial. These areas are equipped with locks to preclude unauthorized entrance and will provide storage for contaminated items and highly radioactive items such as incore detectors until they are used elsewhere orshipped off the site. The Old Steam Generator Storag e Facility is a hot storage area and stores thesteam generators lower assemblies removed from containment. In addition to the hot storageareas, other areas are designated as radioactive material storage areas, used to store radioactive tools and equipment.

Revision 52-09/29/2016 NAPS UFSAR 12.4-212.4.2Facilities and Equipment The facilities available for handling radioactive material that is considered waste are described in Chapter11. A decontamination facility is de scribed in Section9.5.9. A tool andequipment storage facility, is mentioned in Section12.4.1. The exhausts for the hot-lab hoods and laundry are described in Section9.4.7.2. Additional information pertaining to facilities andequipment is contained in Sections12.1.5 and12.3.2.12.4.3Personnel and Procedures The Manager Radiological Protect ion is responsible for the station Radiation Protectionprogram. His duties, experience and qualificati ons are described in Dominion Nuclear Facility Quality Assurance Program Description, Topical Report DO M-QA-1. Reporting to the Manager Radiological Protection are supervis ors, health physicists and techni cians. There are at least five persons assigned to the Health Physics Department at the st ation, meeting the qualifications astechnicians described in ANSI3.1.

12.4Property "ANSI code" (as page type) with input value "ANSI3.1.</br></br>12.4" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..4Required Materials The following by-product, source, and special nuclear materials exceed the amounts inTable1, Regulatory Guide1.70.3, Additional Informa tion, Radioactive Materials Safety for Nuclear Power Plants, dated February1974:*Cs-137 - sealed source for instrument calibration.*Am-Be - sealed neutron source for instrument calibration.

Revision 52-09/29/2016 NAPS UFSAR 12A-iAppendix12A 1Description of Neutron Supplementary Shield1.Appendix12A was submitted as AppendixQ in the original FSAR.

Revision 52-09/29/2016 NAPS UFSAR 12A-ii Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 12A-1APPENDIX12ADESCRIPTION OF NEUTRO N SUPPLEMENTARY SHIELDIn compliance with 10CFR50.55(e), NRC RegionII was notified on April28,1978, that the maximum dose rates on the operating floor of NorthAnna Unit2 could exceed the valuespresented in Chapter12 of the FSAR. By letter dated May25,1978, NRC RegionII was informed that VEPCO was investigating severa l methods of reducing the radiation levels.A final report was submitted on January31,19 79, describing the sh ielding design thatreduces the dose rates to within the Chapter12 lim its. As part of this shielding design effort, a comprehensive re-evaluation of the reactor pr essure vessel (RPV) suppor t system was conducted.

Details of these analyses were provided in the report.By letter, Serial No.300B, dated February22,1979, the report was supplemented withadditional information. With the ne utron shielding in place, the fu el assembly impact loads haveincreased by approximately 10%. This change alone would reduce the margins previouslyreported; however, the loads are still less than th e allowable values. Recent testing on fuel grid impact strength has resulted in Westinghouse's increasing the allowable loads by approximately 25% above those in the report. Th ese new allowables have been previously reported to the NRC on the Diablo Canyon docket (Docket Nos.50-275 and50-323). When using the new allowable loads along with the revised impact loads, the revised margin is higher than in the report. The "better estimate" factor of safe ty of 1.76 would now be approximately 1

.97. In addition, thelimiting stress on the reactor vessel internals at the core barrel girth weld has decreased from that reported. This is a result of the time phasing of the component forces.

The original supplementary neutron shield rest ored expected dose rates inside containment to the original UFSAR Chapter12 limits, and it did not chan ge the conclusions previouslyestablished at the time. Section 12.3, Health Physics Program, now controls personal exposure through ALARA for dose rate concerns, not the original UFSAR Chapter12 limits Table12.1-1 which is considered historical.

In October 2010, the supplementary neutron sh ielding saddle assembli es were observed to be installed over microtherm insu lation. The saddle asse mblies had to be removed, except for the encased metal piece screwed to the supplementary neutron shield collar, to remove the microtherm from the reactor pressure vessel nozzl es to meet the analys is of GSI-191. The saddleassemblies were in such degraded co ndition they could not be reinstalled.

Revision 52-09/29/2016 NAPS UFSAR 12A-212A.1INTRODUCTION The radiation levels inside the reactor containment, determined by radiation surveys(Reference1) on Unit1, were greater than the design levels presented in Chapter12 at two locations:1.The annulus area between the crane wall an d the containment wall on the operating floor(Elevation291ft. 10in.) at crane wall openings.2.Inside the personnel airlock.

The survey results indicated dose rates on the operating fl oor in the annulus area at openings in the crane wall on the order of 2500mRem/hr neutron and 200mRem/hr gamma. The gamma radiation levels were primarily attributable to neutron capture reactions in the containment concrete and steel structures. This conclusion was co nsistent with thermal neutron flux measurements on the order of 3 x10 4 n/cm 2-sec using thermoluminescent dosimetry. The survey results indicated dose rates in the pers onnel airlock on the order of 40mR/hr neutron and2mR/hr gamma.Based on the higher-than-anticipated radiati on levels inside the containment, additionalneutron shielding was designed and installed in both units.The neutron attenuation effectiveness of the shield was conservatively calculated, and the safety analysis demonstrated that the installation of the proposed shielding had no effect on the safety of the plant or the integrity of the reactor vessel support system, and that it substantially reduced the combined neutron and gamma dose rates in the pe rsonnel airlock and in areas required for general containment access.

In October 2010, the supplementary neutron sh ielding saddle assembli es were observed to be installed over microtherm insu lation. The saddle asse mblies had to be removed, except for the encased metal piece screwed to the supplementary neutron shield collar, to remove the microtherm from the reactor pressure vessel nozzl es to meet the analys is of GSI-191. The saddle assemblies were in such degrad ed condition they could not be reinstalled. Following themodification, Health Physics surveys of the Unit 1 an d 2 containments while at power verified that the remaining supplementary neutron shield was still able to meet the design criteria to reduce gamma and neutron radiation in the outer crane wall annulus area.12A.2NEUTRON SHIELD DESIGN CRITERIA The neutron shield is designed to:1.Reduce radiation levels both in the portion of th e annulus area between the crane wall and the containment wall on the operating floor that is required for general containment access and inthe personnel airlock to the levels presented in Chapter12.

Revision 52-09/29/2016 NAPS UFSAR 12A-32.Be a structure that does not require removal dur ing refueling and c oncurrent personnel radiation exposure.3.Have negligible effect on the safety of the plant or the integrity of the reactor vessel support system and reactor coolant system. The effects of the shie ld on reactor pressure vessel internals response and cavity pres sure will not impair the safety of the pl ant or the integrity of the RPV supports.4.Be a structure incapable of becoming a potential missile that could adversely affect any safety-related equipment.5.Permit the required inservice inspection of reactor vessel nozzle and piping welds.12A.3EFFECTIVENESS OF THE SUPPLEMENTARY NEUTRON SHIELDThe effectiveness of the original collar/saddle shield in reducing neutron streaming from the reactor cavity was assessed by two distinctly different calculational methods. The first methodinvolved the use of the COHORT-II Monte Carlo program (Reference2) in an analog mode,starting with an isotropic surface source at the outside surface of the reactor pressure vessel. The second method involved the use of the MORSE M onte Carlo program (Reference3) with neutronalbedo representations of surface scattering and an isotropic source at the outer surface of thereactor pressure vessel.The dose rates in the crane wall openings we re calculated using bot h Monte Carlo programs without the collar/saddle shield in place and compared to measurements at NorthAnna Unit1.The results of these calculations are tabulated in Table12A-1.The neutron dose rates were th en calculated for the same de tector locations with thecollar/saddle shield in place, using both Monte Carlo computer programs. Table12A-2 shows the neutron dose rates for the two calculational methods.The assessment of the effectiveness of the co llar/saddle shield was concentrated at the openings in the crane wall above the operating floor. The eff ect of the crane wall is such that the dose rates in the annular region be tween the crane wall and containm ent wall will be a fraction ofthose levels predicted for the openings. Similarly, the dose rates in the personnel air lock areexpected to be well within the 2.5mRem/hr cr iterion at that locati on as a result of theeffectiveness of the collar/saddle shield.It is also expected, as noted previously, that the actual neutron dose rate s will fall within the range predicted by the two analyses. For the highest neutron radiat ion area in the annular regionon the operating floor (Detector Location5, as shown on Figure12A-1), this would indicatevalues ranging from 25 to 96mRem/hr. Since the gamma dose ra tes on the operating floor areprimarily attributable to (neutron-gamma) reactions with the containment concrete and liner, we Revision 52-09/29/2016 NAPS UFSAR 12A-4 expect the combined neutron-gamma dose rates in the annular region betw een the crane wall andcontainment wall to be below the 100mRem/hr criterion. To reduce even further the potentialexposure rates, openings in the crane wall between the personnel lock and the elevator will beblocked with 3inches of Permali, TypeJN. Th e opening opposite the pe rsonnel lock will beblocked with 6inches of Permali, TypeJN.With the saddle assemblies removed from the supplementary neutron shield design, theoriginal calculations do not represent the current neutron shielding. In order to document the impact of removing the saddle sh ields on the supplementary neutron shield effectiveness, HealthPhysics performed surveys of the 291ft. elevati on of containment at 100%

power in both units.Results of the surveys are in Tables 12A-6 and 12A-7. In Unit1 outer crane wall annulus area, themax neutron dose rates were 95mRem/hr and the max gamma dose rate was 60mRem/hr. In theUnit2 outer crane wall annulus area, the max neutron dose rates was 112.5mRem/hr and the maxgamma dose rate was 30mRem/hr. Both units' pe rsonnel airlocks have dose rates within theoriginal 2.5mRem/hr criterion.12A.4SHIELD DESIGN12A.4.1Description The supplementary neutron shield is composed of these main components:

1.Collar Assembly: As shown in Figure12A-2, the cylindrical collar assembly is composed of six segments, each with an ex tended base and centering tabs.

The segments rest on the top of the neutron shield tank and ar e fastened together by a metal strap to form the collar. The collar fits around the reactor pr essure vessel over the insulati on and extends to the spaces between the nozzles. Each collar segment consists of an outer steel casing, and is filled with a silicon-based neutr on-attenuating material.

2.Saddle Assembly

This was removed in October 2010.

The saddle assembly was removed,except for the encased metal piece. The encased metal piece is now considered part of the collar assembly as it is screwed to the supplementary neutron shield collar.

3.Dust Cover Blocks

The dust cover blocks are silicone-based neutron-attenuating material blocks encased in stainless steel sheet metal.

The blocks are shaped to cover the dust covers on the RPV nozzle support structure and to partia lly fill the space between the dust cover and the collar base underneath each nozzle, as shown in Figures12A-2 and12A-4.

4.Crane Wall Area Shielding

Neutron-attenuating shield material will be placed in the crane wall openings extending from di rectly opposite the personnel hatc h to the elevator entrance and over the portion of the fuel transfer canal behind the crane wall, as shown inFigure12A-5.

Revision 52-09/29/2016 NAPS UFSAR 12A-512A.4.2Location The neutron-shielding co mponents, with the ex ception of the shield ing in the crane wall openings, are all located inside the upper reactor cavity. The bases of the six collar segments rest on the top of the neutron shield ta nk. The collar segments are strappe d together in contact with the RPV insulation. In this position, the collar segmen ts are placed directly in the path of escapingneutrons emanating from the annulus between th e reactor pressure vesse l and the neutron shield tank.The dust cover blocks, shown in Figures12A-2 and12A-4, ar e positioned on top of the neutron shield tank around the dust covers underneath the nozzles.Shielding is located in those crane wall openings shown in Figure12A-5.

The layout arrangement of the supplementary neutron shield is shown in Figure12A-6.12A.4.3Materials The neutron-attenuating material used in the collar and dus t cover blocks is a silicon-based elastomer with a hydrogen density of approximately 0.06gm/cm 3 (4.3% by weight). The shield material will be impregnated with boron carbide (B 4 C) to 2.0% by weight, with the resultanteffective boron density of 0.02gm/cm 3 (1.5% by weight).The material used for attenuati ng neutrons in the crane wall openings is Permali, TypeJN, a densified beechwood laminate that in corporates 6% hydrogen and 3% boron.

The outer wall of the collar segments is constr ucted of 3/8-inch carbon steel, and the innerwall is 10-gauge stainless steel. The dust cover blocks are enca psulated with stainless steel.12A.4.4Supports The entire extended base of the collar rest s on top of the neutron shield tank. The innercylindrical surface rests against the RPV insulation. Additionally, coll ar segments are held together by a metal belt wrapped around the collars at the top.

The dust cover blocks rest on top of the NST and RPV nozzle support stru cture dust coversand are laterally restrained by the collar base.

Shielding sections are supporte d in the crane wall openings by a steel framework attachedto the crane wall.

12A.4.5Missile EffectsThe only credible missiles were the saddle strips on the nozzle of a postulated brokenreactor coolant pipe. With their removal, there are no credible missiles.

Revision 52-09/29/2016 NAPS UFSAR 12A-6The collar segments are not expected to be potential missiles for the following reasons:1.The collar is located so that it is not subjected to direct jet impingement forces from thepostulated limited-displacement breaks.2.The pressurization of the reactor cavity due to the mass and energy released from the break would force the collar segments down against the neutron shield tank, against each other, andagainst the RPV insulation.3.The metal belt around the collar, together with centering tabs at the base of each segment, will keep the collar assembly in place.

Under LOCA conditions, the dust cover blocks will not become missiles because they are not exposed to lifting forces on any surface.12A.4.6Effect on Containment SumpOriginally the saddle strips of the saddle assembly were the only postulated piece of the supplementary neutron shield that was analyzed for effect on the containment sump. With the removal of the saddle strips, the other pieces of the supplementa ry neutron shield do not requirean analysis for effects on the containment su mp due to their composition, size, and shape.12A.5REACTOR PRESSURE VESSEL SUPPORT INTEGRITY REVIEWS A 27-node model was used to calculate the pressure-time hi story in the reactor cavity following a pos tulated 150-in 2 , cold-leg, limited-displacement rupture. The computer code RELAP4/MOD5 8 (with air) was used to calculate the pressure-time transients.

The pressure transients were then transfor med into asymmetric force-time histories andmoment-time histories for application to both the reactor pressure vessel and internal structures.In this regard, the unbalanced forces on the reactor pressure vessel an d the primary shield wall (PSW) were higher than previous ly determined. Peak horizontal RPV force increased from 1540to 1660kips and peak moment increased from 26 x10 3 to 49.5x10 3in-kips.A recalculated RPV support stiffness, using additional flexibil ity in the sliding block, was used in the development of RP V and PSW motion in response to forces on the reactor pressure vessel.The most important changes involved the so-called Case 1 (maximum horizontal RPV displacement). The maximum horiz ontal displacement in fact was relatively unchanged (from0.072 to 0.071i nch), but it had to be combined with RPV rocking (0.00038 vs. 0.000517rad)present at this new, slightly shifted time point (from 0.070 to 0.0737second).

These new displacements were combined with revised PSW asymmetric pressure response data. New loads for the RPV s upport and the neutron shield tank were developed and are Revision 52-09/29/2016 NAPS UFSAR 12A-7presented in Tables12A-3 and12A-4. The RPV no zzle support loads are s hown to be higher thanpreviously reported. It is concluded, however, that none exceed the integrit y definition inherent inFigure12A-7. This figure shows that the new load data remain within the structural integrity limit envelope.Revised relative displacement data are presented in Table12A-5. While these data againshow differences, these values ar e shown to have little effect when compared with the allowable displacement envelope.

It is therefore concluded th at fundamental conclusions relating to the integrity ofRPVsupports and the extent of permissi ble local plasticity are unchanged.

The re-evaluation of the system included the assessment of ch anges in load effects in the steam generator and reactor coolant pump supports. No design-basis loads were affected and no changes to data reported in Section5.5.9 are required.

The analysis of the neutron shield tank and primary shield wall showed that the applied loads are within the material capability of these components.The emergency core cooling system (ECCS) branch piping for Unit2 was stress analyzed.

This evaluation showed that the ECCS branch piping remains integral.12AREFERENCES1.E. A. Warman et al., Radiation Survey in Reactor Containment Building NorthAnna Unit1

,Report RP-30, Stone & Webster Engineering Corporation, July21,1978.2.L. Soffer and L. Clemons, Jr., Cohort-II - A Monte Carlo General Purpose Shielding Computer Code, Report No.NASA TN D-6170, National Aeronautics and SpaceAdministration, April1971.3.E. A. Straker et al., The MORSE Code with Combinatorial Geometry, Report DNA-286 OT,Defense Nuclear Agency, May1972.

Revision 52-09/29/2016 NAPS UFSAR 12A-8Table 12A-1COMPARISON OF CALCULATED NEUTRON DOSE RATES WITH MEASUREMENTS MADE AT NORTHANNA UNIT1, ADJUSTED TO 100% POWER Neutron Dose Rate (mRem/hr)Type of Data Analytical Approach Flux-to-Dose Response Function Detector Location a3456Calculated doseCOHORT IIANSI/ANS-6.1.1-19771920257029302410Equivalent rateMORSESnyder-Neufeld2260330024202300 Measurement (uncorrected for

instrument overres-ponse)2090264028601430a.Refer to Figure12A-1.

Revision 52-09/29/2016 NAPS UFSAR 12A-9Table 12A-2CALCULATED NEUTRON DOSE RATES WITH SUPPLEMENTARY NEUTRON SHIELDING Expected Neutron Dose Rate as Measured with PNR-4 Detector (mRem/hr)AnalyticalApproachDetector Location a 12 b 3456COHORT II method-19082779666MORSE method2854517252519a.Refer to Figure12A-1.b.Detector location 2 is on the inside of the crane wall (i.e., surface of Permali Shield, Type JN).

Revision 52-09/29/2016 NAPS UFSAR 12A-10Table 12A-3REACTOR PRESSURE VESSEL SUPPORT AND NEUTRON SHIELD TANK LOADS PHASELoadTypeF H kipsF VkipsV SWkipsM SWin-kipsPkipsV BkipsM Bin-kipsTin-kips Pipe rupture a125312493509370,268 106726831,8977296 Seismic+/-121 +/-81 +/-259+/-32,467 +/-316+/-278 +/-84,658+/-3883Total13741330 3768402,7351383546116,55511,179 Design capability of NST/RPV support844100025,748 b 617,993 b10,4336260545,964745,955a.Includes internals due to break number 2 plus deadweight plus asymmetric pressurization loading on the primary shield wall, reactor pressure vessel, and neutron shield tank.b.Based on weighted average of mill test reports.

Revision 52-09/29/2016 NAPS UFSAR 12A-11Table 12A-4REACTOR PRESSURE VESSEL NOZZLE SUPPORT LOADS PHASE, INCLUDING REACTOR PRESSURE VESSEL INTERNALS MOVEMENT, ASYMMETRIC PRESSURE, DEADWEIGHT, AND SEISMIC Loads at Nozzle Supports (kips)123456 CommentTime (sec)F H F V F H F V F H F V F H F V F H F V F H F V Maximum horizontal0.073731253291-1224910-3211249-123892598658239-1647 Maximum vertical - up0.1650517551488380-171302-509403-403610-158666 Maximum vertical -

down0.1400 974-1275 925 -597-252-76-962-549-749-1508264-2120 Maximum relative hori-zontal0.13501139-9731090-886-284-663-1126-811-880-1004232-1382 Maximum rotation0.08001233-3181197702-3121151-1216841-962746291-2643 Revision 52-09/29/2016 NAPS UFSAR 12A-12Table 12A-5RELATIVE DISPLACEMENT BETWEEN TO P AND BOTTOM OF NOZZLE SUPPORT a Nozzle Support b Maximum Horizontal at RPV (time=0.07373sec)Maximum Vertical - Up at RPV (time=0.165sec)Maximum Vertical -

Down at RPV (time=0.140sec)

Maximum Relative Horizontal Between RPV and PSW (time=0.135sec)Maximum Rotational at RPV (time=0.080sec) 1D H D V 0.040100 0.009822 0.018163 0.025877 0.030514

-0.008919 0.036592

-0.006801 0.039348

-0.001930 2D H D V 0.040000 0.040737 0.014012 0.016620 0.029768

-0.004422 0.035851

-0.006620 0.038988 0.027734 3D H D V -0.008066 0.057320 -0.002929

0.011255-0.005692

-0.000949-0.006815

-0.005331-0.007745 0.047834 4D H D V -0.039262

0.041790 -0.013612

0.017877-0.029809

-0.004280-0.035804

-0.006277-0.038403 0.034381 5D H D V -0.031263

-0.000081 -0.010673

0.030184-0.023157

-0.010780-0.028006

-0.007147-0.030362

-0.005164 6D H D V 0.003300-0.010485 0.000854 0.035061 0.003795

-0.014138 0.003067

-0.008331 0.004534

-0.017848a.Key: RPV = reactor pressure vessel; PSW = primary shield wall.b.Negative value for D v means nozzle support in compression, and pos itive value means nozzl e support in tension.

Revision 52-09/29/2016 NAPS UFSAR 12A-13Table 12A-6SURVEY RESULTS OF UNIT1 REACTOR CONTAINMENT AT THE 291FT. ELEVATION ON 11/10/10 Survey Point aGamma Dose Rates (mRem/hr)Neutron Dose Rates (mRem/hr)10.290.5024.951.75 314.5530.00 437.5055.50 526.1095.00 660.0055.00 71.003.75 8102.00600.00 929.00100.0010380.00950.00 11274.001350.00 12273.00850.00 1391.50775.00 147.703.00a.Refer to Figure12A-8.

Revision 52-09/29/2016 NAPS UFSAR 12A-14Table 12A-7SURVEY RESULTS OF UNIT2 REACTOR CONTAINMENT AT THE 291FT. ELEVATION ON 10/20/10 Survey Point aGamma Dose Rates (mRem/hr)Neutron Dose Rates (mRem/hr)10.501.027.03.0 330.0112.5 420.085.0 51.503.5 625.047.5 720.042.5 812.03.0 9390.01350.01050.0250.0 11125.0950.0 12325.0775.0 1360.0237.5 14127.5725.0 1514.04.5a.Refer to Figure12A-3.

Revision 52-09/29/2016 NAPS UFSAR 12A-15 Figure 12A-1PLAN VIEW OF OPERATING FLOOR SHOWING DETECTOR LOCATIONS Revision 52-09/29/2016 NAPS UFSAR 12A-16 Figure 12A-2 COLLAR DETAILS Revision 52-09/29/2016 NAPS UFSAR 12A-17 Figure 12A-3PLAN VIEW OF UNIT2 CONTAINMENT FOR SURVEY POINTS Revision 52-09/29/2016 NAPS UFSAR 12A-18 Figure 12A-4SHIELD DUST COVER BLOCKS Revision 52-09/29/2016 NAPS UFSAR 12A-19 Figure 12A-5CRANE WALL OPENINGS WITH PERMALI ELEVATION 291FT. 10 IN.

Revision 52-09/29/2016 NAPS UFSAR 12A-20 Figure 12A-6 LOCATION OF SUPPLEMEN TARY NEUTRON SHIELDS Revision 52-09/29/2016 NAPS UFSAR 12A-21 Figure 12A-7 RPV NOZZLE SUPPORT LOADS Revision 52-09/29/2016 NAPS UFSAR 12A-22 Figure 12A-8PLAN VIEW OF UNIT1 CONTAINMENT FOR SURVEY POINTS