ML17033B570
| ML17033B570 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/29/2016 |
| From: | V Sreenivas Plant Licensing Branch II |
| To: | Heacock D A Virginia Electric & Power Co (VEPCO) |
| Sreenivas V, NRR/DORL/LPL2-1, 415-2597 | |
| Shared Package | |
| ML17033B477 | List: |
| References | |
| Download: ML17033B570 (96) | |
Text
North Anna Power Station Updated Final Safety Analysis ReportChapter 12 Intentionally Blank Intentionally Blank Revision 52-09/29/2016NAPS UFSAR12-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 ContentsSectionTitlePage Revision 52-09/29/2016NAPS UFSAR12-iiChapter 12: Radiation ProtectionTable of Contents (continued)SectionTitlePage12.2.2.2Containment 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. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-512A.4.3Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-5 Revision 52-09/29/2016NAPS UFSAR12-iiiChapter 12: Radiation ProtectionTable of Contents (continued)SectionTitlePage12A.4.4Supports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-512A.4.5Missile Effects. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-512A.4.6Effect on Containment Sump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-612A.5REACTOR PRESSURE VESSEL SUPPORT INTEGRITY REVIEWS . . . . . . .12A-612AReferences. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-7 Revision 52-09/29/2016NAPS UFSAR12-ivChapter 12: Radiation ProtectionList of TablesTableTitlePageTable12.1-1Radiation 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/cm3). . . . . . . .12.2-10Table12.2-2Estimate of Annual Inhalation Doses to Plant Personnela. . . . . . . . . . .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, Asymmetric Pressure, Deadweight, and Seismic. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12A-11Table 12A-5Relative Displacement Between Top and Bottom of Nozzle Support a12A-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 52-09/29/2016NAPS UFSAR12-vChapter 12: Radiation ProtectionList of FiguresFigureTitlePageFigure 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 Equivalentvs. 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 52-09/29/2016NAPS UFSAR12-viIntentionally Blank Revision 52-09/29/2016NAPS UFSAR12-1CHAPTER 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 fencesand/or walls of structures. The containment building, turbine building, auxiliary building, servicebuilding, fuel building and other miscellaneous buildings are within the protected area. From aradiological access standpoint, the area within the protected area is the primary restricted area.Other secondary restricted areas exist within the station proper but outside the protected area,such as the Old Steam Generator Storage Facility. Individuals entering restricted areas must havesatisfactorily completed a basic Health Physics training course or possess the equivalent HealthPhysics knowledge, or be escorted by an individual who has those qualifications.Within the restricted areas, Health Physics procedures are implemented as detailed inSections12.1.5 and12.3. It is anticipated that, during normal station operation, areas outside theestablished 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, theywould be detected by periodic radiation surveys 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 toradiation 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 responsibility 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 maximumdose to individuals but also maintaining the collective dose to personnel, i.e., total man-remexposure, as low as is reasonably achievable.To attain the goal of this commitment, system, station, and contractual personnel shallintegrate their efforts as necessary to perform their functions in such a manner that exposure(s) toradiation will be maintained ALARA. As applicable, new procedures shall be formulated whileexisting procedures and practices shall be reviewed and modified, if necessary, to ensure theirconformance to the principle of maintaining exposures ALARA.
Revision 52-09/29/2016NAPS UFSAR12-2Intentionally Blank Revision 52-09/29/2016NAPS UFSAR12.1-112.1SHIELDING12.1.1Design ObjectivesRadiation protection, including radiation shielding, is designed to ensure that the criteriaspecified in 10CFR20 and 10CFR50 are met during 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 retain their same definitions and therefore thedesign basis accident (DBA) analyses do not require recalculation 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 assumedsource terms, occupancy times and acceptance criteria based on zone criteria. Although thesecriteria were used to establish the original shield design, they were never intended to establishrequirements for the radiation protection implementation during plant operation. As time evolves,source terms change. Acceptable doses have typically decreased with time as ambitious ALARAperson-REM goals are established.Current shielding requirements are non-specific and are established through theimplementation of the Radiation Protection Program and ALARA Program. These programsevaluate the need for a combination of exposure saving principals such as reduced source term,decreasing occupancy time, or increased shielding. These programs use shielding as one methodto help ensure compliance with 10CFR20.This section provides the basis for the original plant shielding design. Although currentdose rates may not be consistent with the zone maps in this chapter, these maps are not beingchanged to be current, as that would make them inconsistent with the original design basis criteriafor the shielding. Recent Heath Physics surveys should be consulted for information on currentstation 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 durations of occupancyduring normal operations and during refueling outages. Descriptions of the zone criteria arepresented in Table12.1-1, and the detailed radiation zone criteria for normal and shutdown 'operations are illustrated on Figures12.1-1 through 12.1-5. These figures do not representoperational 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 52-09/29/2016NAPS UFSAR12.1-2Occupancy time is such that individual radiation 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, decontamination 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 contaminated storage pad orspent-fuel-cask-handling area may temporarily exceed 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 the site. All shieldingdose rate calculations are based on 1% failed fuel elements.Maximum accident doses shall not exceed the following:12.1.2Design DescriptionBuilding 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 ShieldingPrimary shielding is provided to limit radiation emanating from the reactor vessel. Suchradiation consists of neutrons diffusing from the core, prompt fission gammas, fission productAccident or CaseControl Roomexclusion area boundary (EAB) & low population zone (LPZ)Design Basis Loss-of-Coolant Accident (LOCA)5 rem TEDE25 rem TEDESteam Generator Tube RuptureFuel Damage or Pre-accident Spike5 rem TEDE25 rem TEDECoincident Iodine Spike5 rem TEDE2.5 rem TEDEMain Steam Line BreakFuel Damage or Pre-accident Spike5 rem TEDE25 rem TEDECoincident Iodine Spike5 rem TEDE2.5 rem TEDELocked Rotor Accident5 rem TEDE2.5 rem TEDERod Ejection Accident5 rem TEDE6.3 rem TEDEFuel Handling Accident5 rem TEDE6.3 rem TEDE Revision 52-09/29/2016NAPS UFSAR12.1-3gammas, and gammas resulting from the slowing down and capture of neutrons. The primaryshielding is designed to:1.Attenuate neutron flux to prevent excessive activation of components and structures.2.Reduce residual radiation from the core to a level that allows access into the normallyinaccessible region between the primary and secondary shields at a reasonable time aftershutdown.3.Reduce the contribution of radiation from the reactor to optimize the thickness of thesecondary shields.The primary shield consists of a water-filled neutron shield tank and a concrete shield. Theneutron 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 dehydration of the primaryshield wall concrete and minimizes the activation of the plant components within the reactorcontainment. A cooling system is provided for the water in the neutron shield tank. (The neutronshield tank cooling water subsystem is discussed in Section9.2.2.)A 15ft. 8in. high x 2inch thick cylindrical lead shield located beneath the neutron shieldtank protects station personnel servicing the neutron detectors during reactor shutdown.Appendix12A contains a detailed description of supplementary neutron shielding. Themanway in the upper part of the primary shield is plugged during reactor operation. Thecontrol-rod drive concrete missile shield located above the reactor vessel is designed to providesome additional neutron shielding. The primary shield arrangement is shown on Figure12.1-6.The shield materials and thicknesses are listed in Table12.1-2. The application of Permalimaterial for supplementary neutron shielding is shown on Figure12.1-7 for Unit1.A 3-1/2inch thick stainless steel radiation shield is provided at the 12-inch diameter IncoreSump Room drain to protect station personnel during normal power operation and duringrefueling outages.12.1.2.2Secondary ShieldingSecondary shielding consists of the shielding for the reactor coolant, the reactorcontainment, fuel handling equipment, auxiliary equipment, the waste storage area, and the yard,as well as accident shielding.Nitrogen-16 is the major source of radioactivity in the reactor coolant during normaloperation, and its shielding requirements control the combined thickness of the crane andcontainment walls. In areas such as the auxiliary building, where N-16 is not the major source ofactivity, activated corrosion and fission products from the reactor coolant system control thesecondary shielding. Activated corrosion and fission products in the reactor coolant system alsoresult in the shutdown radiation levels in the reactor coolant loop areas. Tables11.1-6 and12.1-3 Revision 52-09/29/2016NAPS UFSAR12.1-4list the activities used in designing the containment secondary shielding. Table11.1-6 lists thefission product activities and activated corrosion products in the reactor coolant system with 1%failed fuel. Table12.1-3 lists the activated corrosion product activities and the N-16 activity at thereactor vessel outlet nozzle.12.1.2.3Reactor Coolant Loop ShieldingInterior shield walls separate the reactor coolant loop, pressurizer, incore instrumentation,and containment access sectors. This shielding allows access to the incore instrument sectorduring normal operation and facilitates maintenance in all sectors during shutdown. The cranesupport wall provides limited access protection in the annulus between the crane wall and thereactor containment wall and provides part of the exterior shielding required during poweroperation. Shield walls are provided around each steam generator above the operating floor to aheight required for personnel protection. Shielding beams below the operating floor arestrategically positioned around the steam generators and reactor coolant pumps. The shieldingbeams provide protection for personnel in the wall annulus from gamma streaming up through therelief 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 ShieldingThe containment shielding consists of the steel-lined, steel-reinforced concrete cylinder andhemispherical dome as described in Section3.8.2. This shielding, together with the crane supportwall, attenuates radiation during full-power operation and during the assumed design basisaccident to or below design levels at the outside surface of the containment and at the siteboundaries.
12.1.2.5Fuel-Handling ShieldingFuel-handling shielding is designed to facilitate the removal and transfer of spent-fuelassemblies from the reactor vessel to the spent-fuel pit. It is designed to protect personnel againstthe radiation emitted from the spent-fuel and control-rod assemblies.The refueling cavity above the reactor vessel is flooded to approximately Elevation290 toprovide a temporary water shield above the components being withdrawn from the reactor vessel.The water height is thus approximately 26feet above the reactor vessel flange. This heightensures approximately 7feet of water above the active portion of a withdrawn fuel assembly at itshighest point of travel. Under these conditions, the dose rate is less than 50mRem/hr at the watersurface.After removal of the fuel from the reactor vessel, it is moved to the spent-fuel pit by the fueltransfer mechanism via the fuel transfer canal. The fuel transfer canal is a passageway connectedto the reactor cavity and extending to the inside wall of the containment structure. The canal is Revision 52-09/29/2016NAPS UFSAR12.1-5formed by two shield walls extending upward to the same height as the reactor cavity. Duringrefueling, the canal and the reactor cavity are flooded with water to the same height.The spent-fuel pit in the fuel building is permanently flooded to provide approximately7feet of water above a fuel assembly when it is being withdrawn from the fuel transfer system.Water height above stored fuel assemblies is a minimum of 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 containment wall and the fuel building wall, and inthe area between the reactor containment wall and the fuel transfer canal.12.1.2.6Auxiliary Equipment ShieldingThe auxiliary components exhibit varying degrees of radioactive contamination due to thehandling of various fluids. The auxiliary shielding protects operating and maintenance personnelworking near the various auxiliary system components, such as those in the Chemical and VolumeControl System, the boron recovery system, the waste disposal system, and the sampling system.Controlled access to the auxiliary building is allowed during reactor operation. Major componentsof systems are individually shielded so that compartments may be entered without having to shutdown 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-exchangestructure along the south wall of the auxiliary building. Each ion exchanger or filter is enclosed ina separate, shielded compartment. The concrete thicknesses provided around the shieldedcompartments are sufficient to reduce the dose rate in the surrounding area to less than2.5mRem/hr and the dose rate to any adjacent cubicle to less than 100mRem/hr. The shieldingthicknesses 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 moreshielding than that required for radiation protection.12.1.2.7Waste Storage ShieldingThe waste storage and processing facilities in the auxiliary building, decontaminationbuilding, and clarifier building are shielded to protect operating personnel in accordance with theradiation protection design bases set forth in Section12.1.1.Boron recovery tanks, which are used to store 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 ifnecessary during station operation.
Revision 52-09/29/2016NAPS UFSAR12.1-6The waste gas decay tanks are located in shielded cubicles, which are buried for missileprotection. The resulting dose rate at the ground surface above the tanks is less than0.75mRem/hr.Periodic surveys by Health Physics personnel using portable radiation detectors ensure thatradiation levels outside the shield walls meet design specifications, and they establish accesslimitations within the shielded cubicles. In addition, continuous surveillance is provided in thewaste solidification area of the decontamination building and in the control board area by arearadiation monitors.12.1.2.8Accident ShieldingAccident shielding is provided by the reactor containment, which is a reinforced-concretestructure 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 main control room. This, together with theshielding afforded by its physical separation from the containment structure, ensures that anoperator 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 forpersonnel protection on the site and to ensure that the dose rate at the exclusion boundary fromdirect 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 radiation dose to personnel insidethe control room envelope (from sources both internal and external to the control room envelope)be less than or equal to 5rem TEDE for the 30day duration of the design basis accident. Thecontrol room northern, western, and eastern walls are 2' thick concrete. The southern wall of thecontrol room is 18" thick concrete. The southern wall of the cable vault is 2' thick concrete tobring the total concrete shielding on the side of the control room facing the containment to 42".The ceiling for the control room is 2' thick concrete. The doorways to the control room are on thenorthern wall of the control room facing away from the containment structure and can be coveredwith radiation shielding doors. Based on NUREG-0800, Section6.4 (Reference8), this level ofshielding allows the dose in the control room from containment shine and cloud shine to betreated as negligible.
Revision 52-09/29/2016NAPS UFSAR12.1-7Special consideration has been given to the design of penetrations and 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 roomnormal replenishment air and exhaust systems are isolated automatically by tight closures in theductwork. Breathing-quality air is discharged from high-pressure storage bottles to theMCR/ESGR envelope. The MCR/ESGR envelope is also provided with an emergency ventilationsystem fitted with particulate and impregnated charcoal filters to introduce cleaned outside airinto 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 safeoperating conditions. Portable monitors are available to provide backup to the fixed monitors.As an additional precaution, personnel air packs are available in the control area.12.1.2.11Shielding Review for NUREG-0578In response to the requirements of NUREG-0578, a design review was conducted using theStone & Webster Engineering Corporation GAMTRAN1 computer code with inputs from theACTIVITY-2 and RADIOISOTOPIC computer codes. The NRC-specified source terms wereused. All systems designed to function after an accident were considered as sources, includingsafety injection, recirculation spray, hydrogen recombiner, sampling, auxiliary building sump, anddrain lines. The letdown portion of the chemical and volume main control system was excludedbecause it is isolated and because its use in the post-accident situation would be unacceptable. Allvital areas were identified and evaluated. Areas where continuous occupancy is required are themain control room, the technical support center, the counting room, the operational supportcenter, and the security control center. Limited access is needed to such places as emergencypower 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 TermsThe total quantity of the principle nuclides in process equipment that contains or transportsradioactivity is identified as a function of operating history in Chapter11. Design and expectedvalues of the radioisotopic inventory for both the reactor coolant and main steam systems arelisted in Section11.1. Design and expected values of the radioisotopic 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 52-09/29/2016NAPS UFSAR12.1-8Table11.1-11 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. Saturationactivities 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 radioactive materials is locatedbehind shielding to minimize the radiation exposure of plant personnel. Pipe tunnels, chases, orshafts are provided as required to properly segregate 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 maybe subject to decontamination are typically finished with epoxy, silicone alkyd, or urethaneenamel protective coatings or approved equal. Stainless steel surfaces are not painted. Stainlesssteel is used extensively in the fuel, decontamination, and waste disposal buildings.Tanks such as the high- and low-level waste tanks, evaporator bottoms tanks, fluid wastetreating tank, and contaminated drain collecting tank have been designed to allow for cleaningand to minimize the buildup of radioactive material using the following factors:1.These tanks are vertical cylindrical tanks with flanged and dished heads to allow completedraining.2.The tank outlet lines are at the lowest point of the tank to aid in complete draining.3.The tanks are of stainless steel construction to minimize corrosion and the buildup of activityand to facilitate cleaning.4.The tanks are provided with inspection openings or manholes that can be used duringcleaning.Drip pan bedplates are provided under pumps. Individual equipment cubicles and pipechases containing radioactive fluid system components and equipment have floor drains that arepiped 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 central sample points connect torecirculation 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 52-09/29/2016NAPS UFSAR12.1-912.1.4Area Monitoring12.1.4.1Normal Plant OperationsThe area radiation monitoring system reads out and records the radiation levels in selectedareas throughout the station, and alarms (audibly and visually) if these levels exceed a presetvalue 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 remotely operated from the maincontrol room. Recorders produce a continuous, permanent record of radiation levels while thedetectors are functioning. Area-radiation-monitoring channels for Unit1 are powered from the480V emergency bus1H; channel monitoring systems or areas common to both units are poweredfrom the emergency bus for either Unit1 or Unit2.The area radiation monitors are designed for continuous operation. Continuous, as used todescribe the operation of an area radiation monitor, means that the monitor provides the requiredinformation at all times with the following exceptions: (1)the monitor is not required to be inoperation 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 thisfunction they are generally located near the main entrance pathway for a given building or portionthereof. In some areas they are located at the major work area involved. In all cases they provide arepresentative indication of the radiation level in that vicinity of the plant and not necessarily themaximum that might be measured against one of the nearby shield walls. The audio and visualalarm provides adequate warning to personnel in the event of an abnormally high radiation level.These monitors have remote displays in the main control room indicating the radiation levelsthroughout the plant, and they may be monitored before entry into potentially high radiationfields. When radioactive material is being handled within a given area, such as thedecontamination building, the monitors provide a representative reading based on planned workareas for handling such material.In addition, if the dose rate at the manipulator crane area monitor exceeds a preset value, thealarm 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 monitor is variable, and it is set at a radiation level slightlyabove that of normal background radiation in the respective area. The monitoring equipmentconsists of fixed-position gamma detectors and associated electronic equipment. These channelswarn of any increase in radiation level at locations where personnel may be expected to remainfor extended periods of time. The instruments and their ranges and locations are listed inTable12.1-4.
Revision 52-09/29/2016NAPS UFSAR12.1-10Tests and calibrations of the radiation monitors are performed at intervals specified in theapplicable Technical Procedures. Special restrictions, as specified in the Technical RequirementsManual, are imposed on plant operators or maintenance activities if the area monitors are notfunctional. The manipulator crane monitor is a control function and is part of a redundant alarmsystem with the containment gaseous and particulate monitors. If the manipulator crane monitor isnot functional, the containment gaseous and particulate monitors can still function and can bebacked up by local portable equipment. This portable equipment, together with Health Physicssurveys during maintenance activities, will allow these activities to continue if a normal fixed areamonitor is not functional.The radiation monitors in the Fuel Building also provide a control function. When a Hi-Hiradiation condition is sensed by either of these monitors, during a fuel handling condition, thecontrol room bottled air system will discharge, the control room normal ventilation will isolate,and the control room/emergency switchgear room emergency ventilation system will startautomatically to recirculate and filter control room air.12.1.4.2Post-Accident ConditionsThe containment high-range radiation monitoring system (CHRRMS) provides indicationin the control room of containment 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 control room readout unit and associated interconnecting cable. Thedetectors are located approximately 155degrees apart for Unit1 and 130degrees apart for Unit2on the inside crane wall to provide physical separation. The location also facilitates the periodiccalibration of the detectors since they are close to the operating floor.The CHRRMS components are qualified to IEEE-323-1974, IEEE-344-1975 and meet therequirements of Regulatory Guide1.97, proposed Revision2. The high range monitors arepowered from diverse Class1E vital buses. The indicators in the control room are installed inracks designed per the separation and seismic requirements of Regulatory Guide1.75, Revision1,and IEEE-344-1975 respectively.The addition of the high-range containment radiation monitors is for indication purposesonly and does not affect the logic schemes of any safety-related systems.The Technical Support Center (TSC) and Local Emergency Operations Facility (LEOF)radiation monitoring systems are localized systems and satisfies the guidelines established inNUREG-0696. The radiation monitoring system components consist of a particulate, iodine, andnoble gas monitor and two area monitors.
Revision 52-09/29/2016NAPS UFSAR12.1-11These monitoring systems provide continuous indication of the dose rate and airborneactivity in the TSC and LEOF during an emergency, as well as alerting personnel of adverseconditions. 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 ProceduresA radiation protection program consistent with the requirements of 10CFR20 anddesigned to ensure that doses are kept ALARA is maintained. Applicable HP procedures,(i.e.,RWPs), are used to control access to all radiation and contaminated areas.The station auxiliary systems containing radioactive fluids are designed for remoteoperation by the use of extensive instrumentation for monitoring, remotely operated pneumatic orelectrical control valves, and manually operated valves with extension stems that allow theoperator to operate the valves while behind shield walls.Special tools are used extensively for fuel handling. These tools and processes are describedin Section9.1.4.The operation of the filter transfer shield, which is used for the handling of spent filtercartridges, is described in Section11.5.3. This transfer shield is of lead and steel construction andfunctions only as a transfer and temporary storage device.A lead shield beneath the neutron shield tank in the containment protects personnel duringthe 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 thechanging of the neutron detectors.Persons or groups entering areas of high radiation are equipped with radiation-monitoringdevices. A person entering an area in which the radiation is greater 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 calculations are describedbelow.
12.1.6.1Sample Sink AreaThe receptors for the sampling sink are located just off the surface of the concrete wallbehind the sinks. Two sources of radiation are considered to be significant in this area: the samplepiping, located in a pipe space behind the wall at which the sampling sinks are located; and thevolume control tanks, located in individual cubicles behind the pipe space, as shown inFigure12.1-2 Sh.3.
Revision 52-09/29/2016NAPS UFSAR12.1-12The volume control tanks are separated by a 2-foot-thick concrete wall. Concrete density ofthis and other concrete walls is 146lb/ft3. 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 was approximated as a source by two right circular cylinders84inch in diameter with 0.25-inch steel walls, with liquid volume of 120ft3 and gaseous volumeof 180ft3.The sample piping primarily consists of 3/4-inch or smaller tubing containing processfluids. The piping is located behind an 18-inch concrete wall. For the purpose of this analysis, themaze of pipes was approximated by four disks side-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 steelplate of minimal thickness to represent the pipe wall thickness.A reduction factor was applied to the source intensity to account for the piping density.Although the fluid in the pipes comes from many different process streams, the conservativeassumption was made that all pipes contained primary coolant samples drawn from the hot leg ofthe 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 diskrepresenting the sample pipes and coincident with the disk axis and intersecting the cylindricalaxis of one of the volume control tanks, the dose rate was calculated to be 4.1mRem/hr. Of thetotal, the sample piping contributed approximately 97%.12.1.6.2Valve-Operating Area Outside Demineralizer CubicleIn the valve-operating area outside the demineralizer cubicle on the 244-foot level of theauxiliary 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 the source because it is the most radioactivesource in the area and because the concrete shielding 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 cylinder 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 39ft3 with a height of 7.13feet.
Revision 52-09/29/2016NAPS UFSAR12.1-13A 2-foot-thick concrete wall extends vertically from Elevation244 to the floor below thedemineralizer cubicle. Above the floor, the wall is 4-foot-thick concrete. The floor of thedemineralizer cubicle is 2-foot-thick concrete. Concrete density in all cases is taken as 146lb/ft3.The computer code GAMTRAN, described below, was used to calculate the dose rates atthe 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 developed point kernel code for shield designanalysis. The gamma ray attenuation coefficients used in GAMTRAN are generated using theOGRE (Reference1) pair production and photoelectric cross sections. The Compton scatteringcomponent 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-doseconversion factors were based on curves in the Reactor Shield Design Manual (Reference5).12.1.7Estimates of ExposureRadiation shielding is provided on the basis of maximum concentrations of radioactivematerials within each shielded region (e.g., 1% failed fuel) rather than the annual average values.For batch processes, as an example, the point of the highest radionuclide concentration in thebatching process (e.g., just before draining a tank) is assumed. The shielding designs are thereforeintentionally conservative in that the dose rates reflect maximum rather than average sources to beshielded.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 doserates would be less than 20% of these values.Shielding thicknesses were calculated using the Stone & Webster code GAMTRANdescribed in Section12.1.6.3. Table12.1-5 lists the densities of the materials used for shieldingcalculations. Care was taken to ensure that the material actually used for construction 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 PredictionsIt is general practice to arrive at the radiation zoning by taking liberal estimates of the timeto be spent in each zone and dividing this into 100mrem/week to arrive at a design value in termsof mRem/hr that will not be exceeded in that zone, even under worst-case conditions. The Revision 52-09/29/2016NAPS UFSAR12.1-14shielding is then designed assuming maximum conditions to ensure that these exposure values arenever exceeded under normal operating conditions. (Higher doses may result from specific repairjobs when shielding is not possible.)The radiation zone designations are shown in Figures12.1-1 through12.1-5. Thesedelineate the maximum dose rates at all locations within the major buildings of NorthAnnaUnits1 and2.Because of the conservatism employed in performing the worst-case dose rate calculations,the shielding is conservatively designed, thus ensuring that the average exposures in each zonewill be far less than the maximum.To compute the expected man-rem values per zone and throughout the plant, the followingitems should be considered:1.Time-and-motion study data must be obtained to allocate time spent in each zone in the plantsuch that the sum of these times equals the total time the employee is at the station in anaverage year.2.An "average employee" concept would not apply because some employees never go in somezones, whereas others frequently spend time in these zones.3.Once in a zone, movement within the zone must be considered.4.The innumerable large and small components in each zone that act as object shields wouldhave 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 typeof 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 tobe included in the modeling to compute expected values. Most of them are conservativelyleft out of the worst-case analysis.6.Conservatism in sources (e.g., 1% failed fuel design defect versus 0.2% expected) wouldhave to be eliminated to predict expected dose rates.7.Explicit margins in other source terms would have to be factored out of the analysis.8.In the worst-case model, each source is assumed to be at maximum levels. This assumes allother sources in that system are at minimum levels. Viewed plant wide, however, anactivities balance would have to be used for average expected conditions.9.Much more complicated mathematical models of large components would have to bedeveloped to replace the few region models which are presently used to intentionallyoverestimate the emanation of radiation from these large sources.
Revision 52-09/29/2016NAPS UFSAR12.1-15A man-rem analysis cannot be computed with sufficient accuracy to obtain good data of apredictive nature. However, sufficient operating data on similar plants do exist to provideestimates of man-rem doses for the station as a whole. This operating experience is demonstrativeof the fact that the radiation shielding is conservatively designed. This is a direct result of thedesign of shields for worst-case conditions, conservative dose rate calculations, and implicit andexplicit designer's margins.12.1.7.2Reports From Other PlantsRelative to the estimations of exposure levels during maintenance, refueling, and inserviceinspection activities, such estimates do not lend themselves to prediction analysis based on ananalytical modeling. Reliance should be placed on operating experience at other stations as themost reasonable source of such data. In this connection, VEPCO's engineers participated in theefforts of the Atomic Industrial Forum's Task Force on Occupational Exposures.One survey reported by Charlesworth (Reference6) at the April1971 American PowerConference covered data obtained at seven operating water-cooled reactor plants with a total plantworker dose of 1700man-rem during the previous year for an average of 244man-rem/yr perplant. In this survey, it was found that on an average 75% of these exposures were estimated tohave been received during shutdown operations.Another survey by Goldman (Reference7) summarizes the results of 27plant-years ofoperation from operating reports. This survey indicated a range of 0.5 to 2.3rem/yr with limiteddata on the number distribution of staff in several exposure categories. From these data, Goldmanconcluded that 19plant-years of operating data resulted in an in-plant population average of238man-rem per plant-year. These results are close to the 244man-rem per plant-year reportedby Charlesworth.The average dose rate level in the visitor's center will be less than 0.01mRem/hr abovenatural background based on the worst-case assumption. Assuming that a visitor will spend4hours at the visitor's center four times per year, he would receive a dose of less than0.16mrem/yr.The expected annual doses to onsite personnel are governed by the controls imposed by thestation supervision and/or Health Physics personnel. However, dose estimates for in-stationpersonnel for routine operation are expected to parallel those reported from operating plantexperience as discussed above.Extensive radiation shielding is provided based on the maximum concentration ofradioactive materials within each shielded region rather than on annual average values. Theshielding and occupancy zones for normal operation are intentionally very conservative so thatthe normally received dose rates should be less than 10% of the limits specified in 10CFR20.
Revision 52-09/29/2016NAPS UFSAR12.1-16The 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, andsolid-radwaste-handling components. Since this is the case, the plant shielding and machinerylocations have been designed to provide maximum laydown space, maximum working room, andminimum time required to perform operations consistent with the reasonable operation of theplant. Experience gained in the operation of nuclear plants has been factored into these designswith the objective of minimizing the total man-rem exposure to plant personnel.12.1.7.3Dose From Stored WasteFor the purpose of a conservative analysis, it is assumed that 1Ci of cobalt-60 equivalent isstored in the low-level contaminated storage area (Reference Drawing4). The dose rates at thevarious distances, including the site boundary, per curie of cobalt-60 equivalent, are presented inFigure12.1-10. No credit is taken for the drum shielding and self-shielding of the waste storedoutside 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 access area. The only anticipatedradioactive sources in this area are radioactive samples brought in for analysis and radioisotopesused in analytical equipment such as radiation monitoring equipment. Therefore, any radiationdoses received while in this area will be controlled by adherence to standard health physicspractices for handling radioactive material. Shielding design for the station as a whole ensuresthat contributions from other station areas do not exceed the design levels for their respectiveareas and make no significant contribution to the service building dose rate.
Revision 52-09/29/2016NAPS UFSAR12.1-1712.1REFERENCES1.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 Shielding 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 ofParameters 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 Contamination 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 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 Station Drawings are controlled by station procedure.Drawing NumberDescription1.11715-FK-9BInstrument Piping, Radiation Monitoring, Sheet 2, Units 1 & 22.11715-FK-9AInstrument Piping, Radiation 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, Radiation Monitoring, Sheet 3, Units 1 & 2 Revision 52-09/29/2016NAPS UFSAR12.1-18The following information is HISTORICAL and is not intended or expected to be updated forthe life of the plant.Table12.1-1RADIATION ZONE CRITERIAZoneAccessMaximum DoseRate (mRem/hr)TypicalLocationsFull-Power OperationIContinuous0.75Main control room, outside surface of containment, and all turbine plant and administration areasIIPeriodic2.5Passageways of auxiliary and fuel buildings, in general, and inside reactor containment personnel lockIIILimited15Outside surface of shielded tank cubiclesIVControlled100Annulus between crane wall and containment wallVRestrictedOver 100Inside shielded equipment compartmentsHot Shutdown (after 15-min decay)IIILimited15Reactor containment above operating floor; outside of crane wallVRestrictedOver 100Inside shielded equipment compartmentsCold Shutdown for Maintenance (after 8-hr decay)IIPeriodic2.5Reactor containment above operating floor and outside of crane wallVRestrictedOver 100Inside shielded equipment compartmentsCold Shutdown for RefuelingIIPeriodic2.5Reactor containment above operating floor, outside of crane wall, and adjacent to fuel transfer canal near incore instrumentation devicesVRestrictedOver 100Inside shielded equipment compartmentsSurface of water over raised fuel assembly50Above fuel assembly when over upender or racks Revision 52-09/29/2016NAPS UFSAR12.1-19Table12.1-2CONTAINMENT SHIELDING SUMMARYSymbolFigureShield DescriptionMaterialaThickness (in)A12.1-8Neutron shield tankWaterSteel343B12.1-8Primary shieldConcrete5412.1-7Supplementary neutron shieldPermali6E12.1-8Neutron shield tank supportSteelLead1.52F12.1-6 and12.1-8Cubicle - crane support wallConcrete33F12.1-8Shielding beamsConcrete24G12.1-8Crane support wallConcrete24 H12.1-6 and 12.1-8Containment wallConcrete54I12.1-8Containment domeConcrete30 J12.1-8Floor elevation 243ftConcrete42 - 48 K12.1-8Operating floorConcrete24 L12.1-6 and 12.1-8Refueling cavity wallConcrete42M12.1-8 and 12.1-9Control-rod drive missile shieldConcrete24N12.1-8Refueling cavity waterWater108O12.1-8 and 12.1-9Removable block wallFacing personnel hatch All othersConcrete Concrete18 12P12.1-6Fuel transfer canal wall (containment structure)Concrete54Q12.1-6Fuel transfer canal wall (containment structure)Concrete72R12.1-6Fuel transfer tube shieldingConcrete 36 (min)S12.1-6Fuel transfer canal wall (fuel building)Concrete72T12.1-6Incore instrumentation cubicle wallConcrete42a.All poured concrete is reinforced with steel.
Revision 52-09/29/2016NAPS UFSAR12.1-20U12.1-6Cubicle wallConcrete36V12.1-6Regenerator heat exchanger wallConcrete24W12.1-6Cable vault wallConcrete24X12.1-6Auxiliary feed pump wallConcrete36Y12.1-6Safeguards area wallConcrete12 Unit2 only Z12.1-8Incore sump room drainStainless Steel31/2Table12.1-2(continued)CONTAINMENT SHIELDING SUMMARYSymbolFigureShield DescriptionMaterialaThickness (in)a.All poured concrete is reinforced with steel.
Revision 52-09/29/2016NAPS UFSAR12.1-21Table12.1-3N-16 AND ACTIVATED CORROSION PRODUCT ACTIVITYIsotopeActivity (µCi/cc@577°F)Mn-545.6x10-4Mn-562.1x10-2Fe-597.5x10-4Co-581.8x10-2Co-605.4x10-4N-16 a73.3a.At the reactor vessel outlet nozzle at 2910 MWt.
Revision 52-09/29/2016NAPS UFSAR12.1-22Table12.1-4AREA RADIATION MONITORING LOCATIONS, NUMBER AND RANGESChannelLocation(number)Range (mRem/hr)Reactor containment area - low range (2)(1/2-RM-RMS-163/263)10 104Personnel hatch area (2)
(1/2-RM-RMS-161/261)10 104Manipulator crane (2)(1/2-RM-RMS-162/262)10 104Incore instrumentation transfer area (2)(1/2-RM-RMS-164/264)10 104Decontamination area (1)
(1-RM-RMS-151)10 104New fuel storage area (1)(1-RM-RMS-152)10 104Fuel pit bridge (1)(1-RM-RMS-153)10 104Auxiliary building area (1)
(1-RM-RMS-154)10 104Waste solidification area (1)(1-RM-RMS-155)10 104Sample room (1)(1-RM-RMS-156)10 104Main control room (1)
(1-RM-RMS-157)10 104Laboratory (1)(1-RM-RMS-158)10 104Technical Support Center (2)(1-RM-RMS-184/185/186)10 104Local Emergency Operations Facility (2)
(1-RM-RMS-187/188/189)10 104 Revision 52-09/29/2016NAPS UFSAR12.1-23Table12.1-5MATERIALS USED FOR SOURCE AND DOSE RATE CALCULATIONSMaterialDensity (lb/ft3)Ilmenite concrete240Ordinary concrete146 Steel490.5 Lead707.6 Air, steam, or vapor0.075 WaterPressurized reactor coolant46 All other62.4Core273.4 Revision 52-09/29/2016NAPS UFSAR12.1-24The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 1 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-25The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 2 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-26The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 3 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-27The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 4 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-28The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 5 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-29The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 6 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-30The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 7 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-31The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-1(SHEET 8 OF 8)RADIATION ZONES CONTAINMENT STRUCTURE Revision 52-09/29/2016NAPS UFSAR12.1-32The 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 52-09/29/2016NAPS UFSAR12.1-33The 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 52-09/29/2016NAPS UFSAR12.1-34The 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 52-09/29/2016NAPS UFSAR12.1-35The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-3(SHEET 1 OF 2)RADIATION ZONES FUEL BUILDING Revision 52-09/29/2016NAPS UFSAR12.1-36The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-3(SHEET 2 OF 2)RADIATION ZONES FUEL BUILDING Revision 52-09/29/2016NAPS UFSAR12.1-37The following information is HISTORICAL and is not intended or expected to be updated for thelife of the plant.Figure 12.1-4(SHEET 1 OF 2)RADIATION ZONES DECONTAMINATION BUILDING Revision 52-09/29/2016NAPS UFSAR12.1-38The 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 52-09/29/2016NAPS UFSAR12.1-39The 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 52-09/29/2016NAPS UFSAR12.1-40Figure 12.1-6SHIELD ARRANGEMENT-PLAN Revision 52-09/29/2016NAPS UFSAR12.1-41Figure 12.1-7PERMALI LOCATIONS Revision 52-09/29/2016NAPS UFSAR12.1-42Figure 12.1-8SHIELD ARRANGEMENT ELEVATION Revision 52-09/29/2016NAPS UFSAR12.1-43Figure 12.1-9SHIELD ARRANGEMENT PLAN OPERATING FLOORSee Appendix12A for a discussion of supplementary neutron shielding.
Revision 52-09/29/2016NAPS UFSAR12.1-44Figure 12.1-10DOSE RATE PER CURIE OF CO-60 EQUIVALENTVS. DISTANCE FROM LOW LEVEL CONTAMINATED STORAGE AREA Revision 52-09/29/2016NAPS UFSAR12.2-112.2VENTILATION12.2.1Design ObjectivesOne of the objectives of the ventilation system is to ensure that the airborne radioactivityconcentration in different locations inside the station buildings during normal operation, includinganticipated operational occurrences, are less than those allowed in Table1, Column3, ofAppendixB of 10CFR20, except in the containment structures. Concentrations in areasaccessible to plant administrative personnel and public visitors areas at the site will be less than1% 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 annualinhalation dose rates for plant personnel in each building are listed in Section12.2.6.The calculational methodology used to perform the design and expected airborneradioactivity levels, which are based on the criteria of the old 10CFR20, are valid analyses anddo 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 reducethe radioiodine in the containment to the derived air concentration (DAC) before personnel entry.Personnel entry will be under administrative control only and will be allowed only in accordancewith standard health physics practices, factoring in activity levels, occupancy times, and approvedbreathing equipment, as discussed in Sections12.1.5 and12.2.5.12.2.2Design DescriptionDetailed descriptions of ventilation systems for different buildings are given in thefollowing sections of this report:SectionSection Title9.4.1Main Control Room and Relay Rooms 9.4.2Auxiliary Building 9.4.3Decontamination and Waste 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 52-09/29/2016NAPS UFSAR12.2-212.2.2.1Auxiliary BuildingThe equilibrium airborne activities in the auxiliary building result from the leakage ofprimary coolant from pump seals and valve stems and from small, miscellaneous leaks. Inaddition, a small amount of iodine is released to the auxiliary building atmosphere from thesampling sink drains, but this is negligible compared to the other assumed leaks. All of the iodinesand noble gases associated with these leaks are assumed to be released to the auxiliary buildingair and exhausted through the auxiliary building ventilation, which exhausts a minimum of10building volumes per hour.In the auxiliary building, the primary coolant letdown to the Chemical and Volume ControlSystem passes through a mixed-bed demineralizer with a decontamination factor of 10 for allisotopes except Cs, Mo, Y, and the noble gases, for which the decontamination factor is 1, whichreduces the ionic activity in the coolant.There is a small potential for leakage upstream of the demineralizer. However, in theanalysis, one-third of the leakage is assumed to occur before the demineralizers; the remainingtwo-thirds is assumed to occur after the demineralizers. 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 inthis area.4.Demineralizer and filter vents are contained by a piping system that discharges via a charcoalfilter and radiation monitor.The radioactive demineralizers are all in individual shielding cubicles along the south wallof the auxiliary building. These cubicles are not connected to the ventilation supply or exhaustsystem (Reference Drawings1 &2). The only air normally passing through these cubicles isslight leakage past valve stem extension or pipe penetration sleeves caused by any minordifference in air pressure between floors of the auxiliary building. Therefore, it is not deemednecessary to provide an exhaust system directly from this area.12.2.2.2Containment StructureThe equilibrium airborne activities in the containment structure have as their source theleakage of primary coolant within the containment for up to 18months prior to purging. Nodilution of the containment atmosphere is assumed during the 6-month period before the purge.
Revision 52-09/29/2016NAPS UFSAR12.2-312.2.2.3Turbine BuildingAirborne activity enters the turbine building atmosphere via the main steam leakagespecified in Section11.1. The turbine building ventilation rate is 7x105scfm and the buildingvolume is 4x106ft3.12.2.2.4Fuel 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% failed fuel, is assumed to be inthe spent-fuel pit. For the expected 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 timebefore fuel can be transferred from the core to the spent-fuel pit.Escape rate coefficients for both design and expected cases for the 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 decontamination factor of 200 foriodines, the same decontamination factor used in the analysis of the fuel-handling accident inSection15.4.5.The fuel building has a ventilation exhaust rate of 35,000scfm and a volume of 160,000ft3.12.2.3Source TermsThe activities listed in Table12.2-1 are based on failed fuel and leakage assumptions givenin Section11.1 and the additional assumptions given in Section12.2.2.12.2.4Airborne Radioactivity MonitoringRadioactivity may become airborne through operations such as the welding or grinding of acontaminated component, the decontamination of such components, leakage from a systemcontaining radioactive fluids or gases, or the disturbance of the deposited activity in various areasof the plant. An airborne sampling 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 monitoring any of eight possible ventilation paths but can beprogrammed as to the sequence and duration of monitoring. Seven of these sample points lie inprobable maintenance or fuel-handling areas. The 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 auxiliary building,(6)the containment purge, and (7)the decontamination building. The ventilation vent multi-port Revision 52-09/29/2016NAPS UFSAR12.2-4sampler particulate monitor and the ventilation vent 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 thecontainment recirculation duct.In the event that concurrent operations are being performed in different work areas, themultisample particulate monitor can be placed on manual and alternated at selected intervalsbetween the work areas. Additionally, process radiation monitors continuously monitor selectedventilation 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. Themultisample monitor does not have a local readout and alarm. The above system can besupplemented 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 samplers is used, it is operated for a known amount of time at aknown flow rate. The filters are removed for counting with appropriate instruments. Dependingon the analysis desired, filters can be counted for beta-gamma, alpha, iodines, or gamma isotopic.Theÿ concentrations are then calculated from these data. If required, 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 routine surveys, if personnel duties in the areaare of a routine or fixed nature and other indicators (i.e., related systems level or pressureindicators, the radiation monitoring system, etc.) show no abnormal conditions, the samplers willbe continuously operated and the filters changed and counted routinely at varying intervals.On occasions when it is expected that conditions could change rapidly or vary considerably,the filters will be changed and counted routinely at varying intervals.The air-sampling program is in addition to or supplements any protective equipment that isauthorized or required by 10CFR20.The sensitivity of the particulate monitor is such that the monitor can detect airborneparticulate levels as low as one-third of the permissible 10CFR20 values. Because theparticulates are collected on a moving filter tape, equilibrium is essentially reached in a collectiontime of 5hours.
Revision 52-09/29/2016NAPS UFSAR12.2-5The 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 notsignificant.The total general area ventilation system flow rate is 74,100cfm. The lowest exhaust flowrate from any building area that exhausts to the general area ventilation system and that isnormally occupied by operating personnel is 12,400cfm. Airborne concentrations in this area aretherefore diluted by a factor of approximately six between the point of intake and the samplingpoint. The sensitivity of the monitors is such that as low as six-tenths of the permissible10CFR20 level for Kr-85 and I-131 is detectable by the ventilation vent sample gas andparticulate monitors. The central air ventilation system flow rate is 60,600cfm. This systemexhausts air from cubicles not normally occupied by operating personnel. The lowest rate ofexhaust 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 monitoredas part of the routine health physics surveys as described in Section12.3.1. The portablemonitoring equipment used in these surveys is described above.The primary function of the central area ventilation vent sample is to warn of abnormalreleases indicative of gross equipment malfunction. In addition, the possible radiation sourceswithin the cubicle areas are limited by design, as discussed in 12.2.2.1. Therefore, the ventilationvent sample monitor, in conjunction with the routine health physics airborne sampling program,provides adequate protection for operating personnel.Background radiation levels and other factors that affect the sensitivity were difficult toquantify until after the station was in operation. To minimize the background contribution, themonitors were located on the upper level of the auxiliary building where the radiation levels wereexpected to be the lowest. Lead shielding reduces the background radiation to a level that does notinterfere with the detector sensitivity. Stainless steel sample lines minimize deposition andplateout losses.The post-accident air monitoring may be performed with portable air samplers, and incompliance with the TMI-2 Lessons Learned requirements. Cartridges are removed and countedin 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-channel analyzers are available for use in air monitoring. The requiredprocedures are in effect. Thus, the capability exists for accurately monitoring iodine in thepresence of noble gases.To comply with the NRC's directive to provide the ability to monitor the post-accidentrelease of potentially high levels of radioactivity via the ventilation system, as expressed in Revision 52-09/29/2016NAPS UFSAR12.2-6NUREG-0578 and clarified in NUREG-0737, high-range effluent monitors have been installed invarious release paths of the plant. They are described in Section11.4.3.12.2.5Operating ProceduresAir sampling and bioassays are used to identify hazards, to evaluate individual exposures,and to assess protection afforded. When the use of respirators is considered necessary, their use isin accordance with written procedures for personnel training and for the selection, fitting, testing,and maintenance of the equipment.Respiratory equipment approved by the National Institute for Occupational Safety andHealth/Mine Safety and Health Administration (NIOSH/MSHA) is used. Equipment not testedand certified by NIOSH/MSHA requires an authorization and exemption be approved by theUSNRC 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 with silicone face-pieces. Regulator use is not to be initiated attemperatures greater than 135°F. Units may be used in areas where temperatures exceed 135°F ifregulator use is initiated prior to entry into the areas. Authorization has been received to use MSAModel Firehawk M7 SCBA charged with 35% oxygen and 65% nitrogen. All units are to beequipped with rubber face-pieces. Breathing gas quality and composition, including hydrocarbonexclusion, are ensured by strict controls and maintained in accordance with the latest revision ofCompressed Gas Association (CGA) specification4.3, GradeE for Oxygen and CGAspecification10.1, GradeB for Nitrogen.12.2.5.1Filter ChangesBefore a filter change, all filter casings are isolated to prevent the flow of air through thecontaminated filters. Filters are removed from their frames and placed directly into a plastic bag.All filter assemblies are provided with adequate working 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 ofportable ventilation units, are provided at the lower level in the ventilation purge exhaust duct toallow removal of radioactive gases from the steam generators or other areas of maintenance.These connections are capped during normal containment operation and the caps are removedwhen necessary to connect flexible duct.In the decontamination building spent-fuel cask area, a flexible hose connection ispermanently installed on the exhaust duct to permit the removal of airborne radioactivity during Revision 52-09/29/2016NAPS UFSAR12.2-7maintenance 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 DosesThe 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 based 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 containment structure, turbinebuilding, auxiliary building, and fuel building are listed in Table12.2-2. Airborne concentrationsused for inhalation dose estimates are 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-breathing equipment normally would beused. After plant shutdown and containment purge, as done in preparation for refuelingoperations, there would be no significant 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 containment 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.1.2x107lb/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 52-09/29/2016NAPS UFSAR12.2-8f.0.1 partition factor for iodines from liquid to steam in the steam generator.g.4.0x106ft3 per unit free volume of the turbine building.h.No credit taken for plateout or decontamination 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) total primary system to auxiliary building leakage, dividedas follows:1)50% from sampling purges, with a partition factor of 103 for iodines released to thebuilding atmosphere.2)16.7% upstream from the mixed-bed demineralizers, with a partition factor of 10 foriodines released to the building atmosphere.3)33.3% downstream from the mixed-bed demineralizers, with a decontamination factorof 10 and a partition factor of 103 for iodines released to the building atmosphere.c.8.1x105ft3 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 5.45x1013/cm2-sec.e.157 fuel assemblies per core.f.One-third of a core from each unit in the spent-fuel pit in the fuel building (105 fuelassemblies).g.A decontamination factor of 100 for iodine in the spent-fuel pit.h.Escape rate coefficients for the spent-fuel pit of 6.5x10-13sec-1 for noble gases and1.3x10-13sec-1 for iodines.i.1.85x105ft3 free volume of the fuel building.j.3.5x104scfm ventilation rate.
Revision 52-09/29/2016NAPS UFSAR12.2-9k.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 12.2REFERENCES1.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 Authorization 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, UnitNos. 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 Station Drawings are controlled by station procedure.Drawing NumberDescription1.11715-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)xMPCiµCi/cc()-----------------------------------------------------------------------------------------------------------------------------------------------=30Remyr-----------1 yr2000 hr------------------x Revision 52-09/29/2016NAPS UFSAR12.2-10Table12.2-1EQUILIBRIUM ACTIVITIES IN DIFFERENT PLANT BUILDINGS (CI/CM3)Auxiliary BuildingTurbine BuildingContainment StructureFuel BuildingIsotopeDesignExpectedDesignExpectedDesignExpectedDesignExpectedKr-85m1.3x10-081.3x10-09----1.4x10-061.5x10-071.2x10-152.3x10-16Kr-853.1x10-083.1x10-09----2.5x10-032.0x10-042.9x10-104.9x10-11Kr-877.1x10-097.1x10-10----2.5x10-072.5x10-08----Kr-882.2x10-082.2x10-09----1.6x10-061.6x10-074.0x10-197.9x10-20Xe-131m1.5x10-121.5x10-13----7.4x10-057.4x10-063.5x10-097.0x10-10Xe-133m1.9x10-081.9x10-09----2.7x10-052.7x10-064.9x10-109.7x10-11Xe-1331.7x10-061.7x10-07----5.7x10-035.7x10-043.0x10-086.1x10-10Xe-135m9.1x10-109.1x10-11----6.8x10-076.8x10-083.3x10-136.5x10-14Xe-1353.7x10-083.7x10-09----1.1x10-051.1x10-065.3x10-111.1x10-11Xe-1383.3x10-093.3x10-10----3.0x10-083.0x10-09----I-1313.0x10-093.0x10-102.1x10-111.4x10-122.2x10-062.0x10-072.7x10-115.4x10-12I-1321.1x10-091.1x10-103.0x10-121.7x10-133.9x10-073.7x10-082.3x10-114.7x10-12I-1334.9x10-094.9x10-102.3x10-111.3x10-122.9x10-062.7x10-073.1x10-126.2x10-13I-1346.3x10-106.3x10-113.4x10-131.4x10-146.8x10-086.7x10-09----I-1352.6x10-092.6x10-107.1x10-123.3x10-131.1x10-069.9x10-082.5x10-155.1x10-16µ Revision 52-09/29/2016NAPS UFSAR12.2-11Table12.2-2ESTIMATE OF ANNUAL INHALATION DOSES TO PLANT PERSONNELaLocationEstimated Annual Dose (rem)Containment structure, Unit10.78Containment structure, Unit2 0.78 Turbine building0.0023Auxiliary building0.060Fuel building0.0024ba.Personnel whose work areas are normally in the locations designated above. Other plant personnel, such as administrative personnel, are expected 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 52-09/29/2016NAPS UFSAR12.2-12Intentionally Blank Revision 52-09/29/2016NAPS UFSAR12.3-112.3HEALTH PHYSICS PROGRAM12.3.1Program Objectives and ProceduresThe Radiological Protection program provides the guidance and technical support requiredwith the handling and evaluation of radiological hazards associated with the operation andmaintenance of the station. The administration of the program is the responsibility of the ManagerRadiological Protection.The Radiological Protection program consist of administrative and technical proceduresand other associated Health Physics documents. This program and its revisions are approved bythe Facility Safety Review Committee and is available for onsite review by the NRC. Each stationemployee receives training in basic radiation protection as described in Section13.2. A RadiationWork Permit system is included in the Radiation Protection program and is described in theapplicable Health Physics procedures. Protective clothing and other requirements are listed on orreferenced by the permit.Operating guidelines and rules to ensure that Total Effective Dose Equivalent (TEDE) willbe ALARA during operation and maintenance are provided in the Radiological Protectionprogram. Each station employee will be oriented as to its contents and usually quizzed to ensurehis/her competence. Individuals deliberately violating procedures set forth in the program will besubject to administrative action.Periodic radiation and contamination surveys by health physics personnel ensure thatcurrent radiological conditions are known. Results of these surveys are posted at the entrance tothe radiological control area, the station's main health physics control point. Station personneltherefore have access to information regarding current radiological conditions in the area theyintend to visit.Station personnel will be issued dosimetry equipment, including indicating dosimeters, foractivities within the radiological controlled areas. A system has been devised whereby theindividual's accumulated exposure, after performing a job within the radiological control areas, islogged, 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, HealthPhysics 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 ofindividuals will minimize exposure.Personnel doses will be limited to 10CFR20.1201 limits. Administrative controls will beimplemented 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 both general area and contact dose rate readings.
Revision 52-09/29/2016NAPS UFSAR12.3-2The In-Plant Radiation Monitoring Program ensures the capability to accurately determinethe airborne iodine concentration in vital areas under accident conditions. This program includes(1)training of personnel, (2)procedures for monitoring, and (3)provisions for maintenance ofsampling and analysis equipment.Health physics personnel perform regular in-plant surveys in all areas where personnelaccess is required. The frequency depends on the area in question and on current plant conditions,and is defined in the Radiological 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 EquipmentThe health physics facility is located in the service building corridor leading to the auxiliarybuilding and thus is convenient to all personnel entering and exiting the RCA. The facilitiesinclude office space, briefing room, labs, a count room, change rooms, dosimetry issue area,instrument issue, laundry area and a personnel decontamination area. These facilities are shownon Reference Drawing1.Locker rooms are provided for personnel entering the RCA. A change out area is located inthe RCA for the donning and storage of protective clothing. An ample supply of coveralls, labcoats, hoods, shoe covers, rubber gloves, plastic 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 asrequired. Showers and sinks are provided to aid in any personnel decontamination effort.Fixed and portable instrumentation is available for counting and/or detecting and indicatingradiation levels from all radiation sources at the station. A sufficient number are on hand to ensurecontinued availability. Calibration/recalibration is performed in accordance with applicabletechnical procedures.Respiratory protection devices are available to protect personnel from airborne radioactivityand are issued in accordance with the applicable RWP.Radiation areas are clearly posted and warning signs, barricades and locked doors are usedin accordance with the Radiation Protection program to protect personnel from inadvertent accessto high radiation areas.Additional shielding material is available as needed and can be used on either a permanentor temporary basis. The material consist of lead blankets, steel sheets and concrete blocks. Aspecial transfer cask is available for handling highly radioactive filters. Remote-handling tools areavailable for handling small lightweight objects or remotely operating valves or other Revision 52-09/29/2016NAPS UFSAR12.3-3components, while cranes and monorails can afford the distance required for handling heavierobjects.Personnel exiting any RCA are monitored for radioactive contamination in accordance withthe Radiation Protection program. Additional monitoring is performed for personnel exiting theprimary restricted area.12.3.3Personnel DosimetryExternal dosimetry is provided for all personnel who enter any radiological controlled areaor radioactive material storage area at the station. Thermoluminescent dosimetry (TLD) badgesare used to determine lens dose equivalent, shallow dose equivalent, effective dose equivalent anddeep dose equivalent as required by 10CFR20. Indicating dosimeters are used to estimate dosesin the periods between badge readings. Extremity dosimetry is worn in accordance with theapplicable RWP.TLD dosimeters will be calibrated 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 Bioassayprogram quantifies the amount of radioactive material present in workers and converts the resultsto 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 minimizeairborne radioactive material. Frequencies, procedures and types of analyses are defined in theRadiation 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-bodycounted as required by radiation protection.Standard lab equipment is available to prepare samples as required for counting. Distillingapparatus and ion-exchange columns are available 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 Station Drawings are controlled by station procedure.Drawing NumberDescription1.11715-FM-5AArrangement: Service Building, Sheet 1 Revision 52-09/29/2016NAPS UFSAR12.3-4Intentionally Blank Revision 52-09/29/2016NAPS UFSAR12.4-112.4RADIOACTIVE MATERIALS SAFETY12.4.1Materials Safety ProgramsEstablished health physics procedures require the notification of the Radiation ProtectionDepartment of the arrival of radioactive materials at the station. Appropriate surveys andinventory are then taken and the material is taken to a designated area for storage and/or use.High-activity sources, such as reactor start-up sources, are normally stored in their shippingcontainers, in other appropriate containers, or under water until their use is required, at which timeHealth Physics coverage will be provided. Sources such as those required for calibratinghigh-range gamma survey meters are obtained from manufacturers in shielded devices designedso that the sources cannot be readily removed and so that doses to those using the sources can bekept ALARA. Other calibration sources will be stored in locked areas and/or shielded containers,and their removal will be by authorized personnel only.The use of unsealed by-product material received at the site is essentially limited to that ofhealth physics or chemistry personnel in the preparation of low-level calibration sources for countroom equipment. It is not expected that any unsealed, special nuclear material will be received atthe site.The Radiological Protection Plan requires that no radioactive material or suspectedradioactive material be carried or removed from a restricted area without Health Physics'notification and approval. Within the restricted area, all unattended tools, loose components, orequipment containing or contaminated with radioactive 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 tools back and forth between clean and radiological controlled areas.These tools are periodically checked and decontaminated as required. When special tools arerequired and used, they must be surveyed by Health Physics before leaving the radiologicalcontrolled areas for storage or use in other areas of the station.Hot storage areas are provided to contain and control radioactive material. These areas areequipped with locks to preclude unauthorized entrance and will provide storage for contaminateditems and highly radioactive items such as incore detectors until they are used elsewhere orshipped off the site. The Old Steam Generator Storage 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 radioactivetools and equipment.
Revision 52-09/29/2016NAPS UFSAR12.4-212.4.2Facilities and EquipmentThe facilities available for handling radioactive material that is considered waste aredescribed in Chapter11. A decontamination facility is described in Section9.5.9. A tool andequipment storage facility, is mentioned in Section12.4.1. The exhausts for the hot-lab hoods andlaundry 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 ProceduresThe Manager Radiological Protection is responsible for the station Radiation Protectionprogram. His duties, experience and qualifications are described in Dominion Nuclear FacilityQuality Assurance Program Description, Topical Report DOM-QA-1. Reporting to the ManagerRadiological Protection are supervisors, health physicists and technicians. There are at least fivepersons assigned to the Health Physics Department at the station, 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 MaterialsThe following by-product, source, and special nuclear materials exceed the amounts inTable1, Regulatory Guide1.70.3, Additional Information, Radioactive Materials Safety forNuclear Power Plants, dated February1974:*Cs-137 - sealed source for instrument calibration.*Am-Be - sealed neutron source for instrument calibration.
Revision 52-09/29/2016NAPS UFSAR12A-iAppendix12A1Description of Neutron Supplementary Shield1.Appendix12A was submitted as AppendixQ in the original FSAR.
Revision 52-09/29/2016NAPS UFSAR12A-iiIntentionally Blank Revision 52-09/29/2016NAPS UFSAR12A-1APPENDIX12ADESCRIPTION OF NEUTRON SUPPLEMENTARY SHIELDIn compliance with 10CFR50.55(e), NRC RegionII was notified on April28,1978, thatthe 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 wasinformed that VEPCO was investigating several methods of reducing the radiation levels.A final report was submitted on January31,1979, describing the shielding design thatreduces the dose rates to within the Chapter12 limits. As part of this shielding design effort, acomprehensive re-evaluation of the reactor pressure vessel (RPV) support 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 neutron shielding in place, the fuel assembly impact loads haveincreased by approximately 10%. This change alone would reduce the margins previouslyreported; however, the loads are still less than the allowable values. Recent testing on fuel gridimpact strength has resulted in Westinghouse's increasing the allowable loads by approximately25% above those in the report. These new allowables have been previously reported to the NRCon the Diablo Canyon docket (Docket Nos.50-275 and50-323). When using the new allowableloads along with the revised impact loads, the revised margin is higher than in the report. The"better estimate" factor of safety 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 thatreported. This is a result of the time phasing of the component forces.The original supplementary neutron shield restored expected dose rates inside containmentto the original UFSAR Chapter12 limits, and it did not change the conclusions previouslyestablished at the time. Section 12.3, Health Physics Program, now controls personal exposurethrough ALARA for dose rate concerns, not the original UFSAR Chapter12 limits Table12.1-1which is considered historical.In October 2010, the supplementary neutron shielding saddle assemblies were observed tobe installed over microtherm insulation. The saddle assemblies had to be removed, except for theencased metal piece screwed to the supplementary neutron shield collar, to remove themicrotherm from the reactor pressure vessel nozzles to meet the analysis of GSI-191. The saddleassemblies were in such degraded condition they could not be reinstalled.
Revision 52-09/29/2016NAPS UFSAR12A-212A.1INTRODUCTIONThe radiation levels inside the reactor containment, determined by radiation surveys(Reference1) on Unit1, were greater than the design levels presented in Chapter12 at twolocations:1.The annulus area between the crane wall and 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 floor in the annulus area atopenings in the crane wall on the order of 2500mRem/hr neutron and 200mRem/hr gamma. Thegamma radiation levels were primarily attributable to neutron capture reactions in thecontainment concrete and steel structures. This conclusion was consistent with thermal neutronflux measurements on the order of 3x104n/cm2-sec using thermoluminescent dosimetry. Thesurvey results indicated dose rates in the personnel airlock on the order of 40mR/hr neutron and2mR/hr gamma.Based on the higher-than-anticipated radiation levels inside the containment, additionalneutron shielding was designed and installed in both units.The neutron attenuation effectiveness of the shield was conservatively calculated, and thesafety analysis demonstrated that the installation of the proposed shielding had no effect on thesafety of the plant or the integrity of the reactor vessel support system, and that it substantiallyreduced the combined neutron and gamma dose rates in the personnel airlock and in areasrequired for general containment access.In October 2010, the supplementary neutron shielding saddle assemblies were observed tobe installed over microtherm insulation. The saddle assemblies had to be removed, except for theencased metal piece screwed to the supplementary neutron shield collar, to remove themicrotherm from the reactor pressure vessel nozzles to meet the analysis of GSI-191. The saddleassemblies were in such degraded condition they could not be reinstalled. Following themodification, Health Physics surveys of the Unit 1 and 2 containments while at power verifiedthat the remaining supplementary neutron shield was still able to meet the design criteria toreduce gamma and neutron radiation in the outer crane wall annulus area.12A.2NEUTRON SHIELD DESIGN CRITERIAThe neutron shield is designed to:1.Reduce radiation levels both in the portion of the annulus area between the crane wall and thecontainment wall on the operating floor that is required for general containment access and inthe personnel airlock to the levels presented in Chapter12.
Revision 52-09/29/2016NAPS UFSAR12A-32.Be a structure that does not require removal during refueling and concurrent personnelradiation exposure.3.Have negligible effect on the safety of the plant or the integrity of the reactor vessel supportsystem and reactor coolant system. The effects of the shield on reactor pressure vesselinternals response and cavity pressure will not impair the safety of the plant or the integrityof the RPV supports.4.Be a structure incapable of becoming a potential missile that could adversely affect anysafety-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 thereactor 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. Thesecond method involved the use of the MORSE Monte 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 were calculated using both Monte Carlo programswithout 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 then calculated for the same detector locations with thecollar/saddle shield in place, using both Monte Carlo computer programs. Table12A-2 shows theneutron dose rates for the two calculational methods.The assessment of the effectiveness of the collar/saddle shield was concentrated at theopenings in the crane wall above the operating floor. The effect of the crane wall is such that thedose rates in the annular region between the crane wall and containment 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 criterion at that location as a result of theeffectiveness of the collar/saddle shield.It is also expected, as noted previously, that the actual neutron dose rates will fall within therange predicted by the two analyses. For the highest neutron radiation 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 rates on the operating floor areprimarily attributable to (neutron-gamma) reactions with the containment concrete and liner, we Revision 52-09/29/2016NAPS UFSAR12A-4expect the combined neutron-gamma dose rates in the annular region between 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. The opening opposite the personnel 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 theimpact of removing the saddle shields on the supplementary neutron shield effectiveness, HealthPhysics performed surveys of the 291ft. elevation 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' personnel airlocks have dose rates within theoriginal 2.5mRem/hr criterion.12A.4SHIELD DESIGN12A.4.1DescriptionThe supplementary neutron shield is composed of these main components:1.Collar Assembly: As shown in Figure12A-2, the cylindrical collar assembly is composed ofsix segments, each with an extended base and centering tabs. The segments rest on the top ofthe neutron shield tank and are fastened together by a metal strap to form the collar. Thecollar fits around the reactor pressure vessel over the insulation and extends to the spacesbetween the nozzles. Each collar segment consists of an outer steel casing, and is filled witha silicon-based neutron-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 thecollar assembly as it is screwed to the supplementary neutron shield collar.3.Dust Cover Blocks: The dust cover blocks are silicone-based neutron-attenuating materialblocks encased in stainless steel sheet metal. The blocks are shaped to cover the dust coverson the RPV nozzle support structure and to partially fill the space between the dust cover andthe 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 cranewall openings extending from directly opposite the personnel hatch to the elevator entranceand over the portion of the fuel transfer canal behind the crane wall, as shown inFigure12A-5.
Revision 52-09/29/2016NAPS UFSAR12A-512A.4.2LocationThe neutron-shielding components, with the exception of the shielding in the crane wallopenings, are all located inside the upper reactor cavity. The bases of the six collar segments reston the top of the neutron shield tank. The collar segments are strapped together in contact with theRPV insulation. In this position, the collar segments are placed directly in the path of escapingneutrons emanating from the annulus between the reactor pressure vessel and the neutron shieldtank.The dust cover blocks, shown in Figures12A-2 and12A-4, are positioned on top of theneutron 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.4.3MaterialsThe neutron-attenuating material used in the collar and dust cover blocks is a silicon-basedelastomer with a hydrogen density of approximately 0.06gm/cm3 (4.3% by weight). The shieldmaterial will be impregnated with boron carbide (B4C) to 2.0% by weight, with the resultanteffective boron density of 0.02gm/cm3 (1.5% by weight).The material used for attenuating neutrons in the crane wall openings is Permali, TypeJN, adensified beechwood laminate that incorporates 6% hydrogen and 3% boron.The outer wall of the collar segments is constructed of 3/8-inch carbon steel, and the innerwall is 10-gauge stainless steel. The dust cover blocks are encapsulated with stainless steel.12A.4.4SupportsThe entire extended base of the collar rests on top of the neutron shield tank. The innercylindrical surface rests against the RPV insulation. Additionally, collar segments are heldtogether 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 structure dust coversand are laterally restrained by the collar base.Shielding sections are supported 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 52-09/29/2016NAPS UFSAR12A-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 breakwould 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 arenot 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 thesupplementary neutron shield that was analyzed for effect on the containment sump. With theremoval of the saddle strips, the other pieces of the supplementary neutron shield do not requirean analysis for effects on the containment sump due to their composition, size, and shape.12A.5REACTOR PRESSURE VESSEL SUPPORT INTEGRITY REVIEWSA 27-node model was used to calculate the pressure-time history in the reactor cavityfollowing a postulated 150-in2, cold-leg, limited-displacement rupture. The computer codeRELAP4/MOD58 (with air) was used to calculate the pressure-time transients.The pressure transients were then transformed 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 and the primary shield wall(PSW) were higher than previously determined. Peak horizontal RPV force increased from 1540to 1660kips and peak moment increased from 26x103 to 49.5x103in-kips.A recalculated RPV support stiffness, using additional flexibility in the sliding block, wasused in the development of RPV and PSW motion in response to forces on the reactor pressurevessel.The most important changes involved the so-called Case1 (maximum horizontal RPVdisplacement). The maximum horizontal displacement in fact was relatively unchanged (from0.072 to 0.071inch), 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 responsedata. New loads for the RPV support and the neutron shield tank were developed and are Revision 52-09/29/2016NAPS UFSAR12A-7presented in Tables12A-3 and12A-4. The RPV nozzle support loads are shown to be higher thanpreviously reported. It is concluded, however, that none exceed the integrity definition inherent inFigure12A-7. This figure shows that the new load data remain within the structural integrity limitenvelope.Revised relative displacement data are presented in Table12A-5. While these data againshow differences, these values are shown to have little effect when compared with the allowabledisplacement envelope.It is therefore concluded that fundamental conclusions relating to the integrity ofRPVsupports and the extent of permissible local plasticity are unchanged.The re-evaluation of the system included the assessment of changes in load effects in thesteam generator and reactor coolant pump supports. No design-basis loads were affected and nochanges to data reported in Section5.5.9 are required.The analysis of the neutron shield tank and primary shield wall showed that the appliedloads 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 ShieldingComputer 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 52-09/29/2016NAPS UFSAR12A-8Table 12A-1COMPARISON OF CALCULATED NEUTRON DOSE RATES WITH MEASUREMENTS MADE AT NORTHANNA UNIT1, ADJUSTED TO 100% POWERNeutron Dose Rate (mRem/hr)Type of DataAnalytical ApproachFlux-to-Dose Response FunctionDetector Locationa3456Calculated doseCOHORT IIANSI/ANS-6.1.1-19771920257029302410Equivalent rateMORSESnyder-Neufeld2260330024202300Measurement (uncorrected for instrument overres-ponse)2090264028601430a.Refer to Figure12A-1.
Revision 52-09/29/2016NAPS UFSAR12A-9Table 12A-2CALCULATED NEUTRON DOSE RATES WITH SUPPLEMENTARY NEUTRON SHIELDINGExpected Neutron Dose Rate as Measured with PNR-4 Detector (mRem/hr)AnalyticalApproachDetector Location a12 b3456COHORT 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 52-09/29/2016NAPS UFSAR12A-10Table 12A-3REACTOR PRESSURE VESSEL SUPPORT AND NEUTRON SHIELD TANK LOADS PHASELoadTypeFH kipsFVkipsVSWkipsMSWin-kipsPkipsVBkipsMBin-kipsTin-kipsPipe rupture a125312493509370,268 106726831,8977296Seismic+/-121 +/-81 +/-259+/-32,467 +/-316+/-278 +/-84,658+/-3883Total13741330 3768402,7351383546116,55511,179Design capability of NST/RPV support844100025,748 b617,993 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 52-09/29/2016NAPS UFSAR12A-11Table 12A-4REACTOR PRESSURE VESSEL NOZZLE SUPPORT LOADS PHASE, INCLUDING REACTOR PRESSURE VESSEL INTERNALS MOVEMENT, ASYMMETRIC PRESSURE, DEADWEIGHT, AND SEISMICLoads at Nozzle Supports (kips)123456CommentTime (sec)FHFVFHFVFHFVFHFVFHFVFHFVMaximum horizontal0.073731253291-1224910-3211249-123892598658239-1647Maximum vertical - up0.1650517551488380-171302-509403-403610-158666Maximum vertical -
down0.1400 974-1275 925 -597-252-76-962-549-749-1508264-2120Maximum relative hori-zontal0.13501139-9731090-886-284-663-1126-811-880-1004232-1382Maximum rotation0.08001233-3181197702-3121151-1216841-962746291-2643 Revision 52-09/29/2016NAPS UFSAR12A-12Table 12A-5RELATIVE DISPLACEMENT BETWEEN TOP AND BOTTOM OF NOZZLE SUPPORT aNozzle Support bMaximum 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)1DHDV 0.0401000.0098220.018163 0.0258770.030514
-0.0089190.036592
-0.0068010.039348
-0.0019302DHDV 0.0400000.040737 0.014012 0.016620 0.029768
-0.0044220.035851
-0.006620 0.038988 0.0277343DHDV -0.0080660.057320 -0.002929 0.011255-0.005692
-0.000949-0.006815
-0.005331-0.007745 0.0478344DHDV -0.039262 0.041790 -0.013612 0.017877-0.029809
-0.004280-0.035804
-0.006277-0.038403 0.0343815DHDV -0.031263-0.000081 -0.010673 0.030184-0.023157
-0.010780-0.028006
-0.007147-0.030362
-0.0051646DHDV 0.003300-0.010485 0.000854 0.0350610.003795
-0.0141380.003067
-0.008331 0.004534
-0.017848a.Key: RPV = reactor pressure vessel; PSW = primary shield wall.b.Negative value for Dv means nozzle support in compression, and positive value means nozzle support in tension.
Revision 52-09/29/2016NAPS UFSAR12A-13Table 12A-6SURVEY RESULTS OF UNIT1 REACTOR CONTAINMENT AT THE 291FT. ELEVATION ON 11/10/10Survey PointaGamma 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 52-09/29/2016NAPS UFSAR12A-14Table 12A-7SURVEY RESULTS OF UNIT2 REACTOR CONTAINMENT AT THE 291FT. ELEVATION ON 10/20/10Survey PointaGamma 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 52-09/29/2016NAPS UFSAR12A-15Figure 12A-1PLAN VIEW OF OPERATING FLOOR SHOWING DETECTOR LOCATIONS Revision 52-09/29/2016NAPS UFSAR12A-16Figure 12A-2COLLAR DETAILS Revision 52-09/29/2016NAPS UFSAR12A-17Figure 12A-3PLAN VIEW OF UNIT2 CONTAINMENT FOR SURVEY POINTS Revision 52-09/29/2016NAPS UFSAR12A-18Figure 12A-4SHIELD DUST COVER BLOCKS Revision 52-09/29/2016NAPS UFSAR12A-19Figure 12A-5CRANE WALL OPENINGS WITH PERMALI ELEVATION 291FT. 10 IN.
Revision 52-09/29/2016NAPS UFSAR12A-20Figure 12A-6LOCATION OF SUPPLEMENTARY NEUTRON SHIELDS Revision 52-09/29/2016NAPS UFSAR12A-21Figure 12A-7RPV NOZZLE SUPPORT LOADS Revision 52-09/29/2016NAPS UFSAR12A-22Figure 12A-8PLAN VIEW OF UNIT1 CONTAINMENT FOR SURVEY POINTS