ML110040881

Bellefonte Units 3 & 4 Cola (Final Safety Analysis Report), Rev. 3 - FSAR Chapter 12 Radiation Protection - Sections 12.01 - 12.05, Appendices 12AA

ML110040881
Person / Time
Site:	Bellefonte
Issue date:	12/22/2010
From:	Arent G Tennessee Valley Authority
To:	Document Control Desk, Office of New Reactors
Spink T
References
BELLEFONTE.P02.NP, BELLEFONTE.P02.NP.3, TENNVALLEY, TENNVALLEY.SUBMISSION.6, +reviewedmmc1
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CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS Section Title Page 12.1 ASSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS-LOW-AS-REASONABLY ACHIEVABLE (ALARA).............. 12.1-1 12.1.2.4.3 Equipment Layout.................................................................. 12.1-1 12.1.3 COMBINED LICENSE INFORMATION....................................... 12.1-1 12.2 RADIATION SOURCES .................................................................... 12.2-1 12.2.1.1.10 Miscellaneous Sources ......................................................... 12.2-1 12.2.3 COMBINED LICENSE INFORMATION....................................... 12.2-2 12.3 RADIATION PROTECTION DESIGN FEATURES ........................... 12.3-1 12.3.1.2 Radiation Zoning and Access Control ................................... 12.3-1 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION.......................................... 12.3-1 12.3.5.1 Administrative Controls for Radiological Protection............... 12.3-5 12.3.5.2 Criteria and Methods for Radiological Protection .................. 12.3-5 12.3.5.3 Groundwater Monitoring Program ......................................... 12.3-5 12.3.5.4 Record of Operational Events of Interest for Decommissioning .................................................................. 12.3-5 12.4 DOSE ASSESSMENT....................................................................... 12.4-1 12.4.1.9 Dose to Construction Workers............................................... 12.4-1 12.4.1.9.1 Site Layout............................................................................. 12.4-1 12.4.1.9.2 Radiation Sources ................................................................. 12.4-1 12.4.1.9.3 Construction Worker Dose Estimates.................................... 12.4-2 12.4.1.9.4 Compliance with Dose Regulations ....................................... 12.4-3 12.4.1.9.5 Collective Doses to BLN Unit 4 Workers ............................... 12.4-3 12.4.1.9.6 Operating Unit Radiological Surveys ..................................... 12.4-3 12.5 HEALTH PHYSICS FACILITIES DESIGN......................................... 12.5-1 12.5.2.2 Facilities................................................................................. 12.5-1 12.5.4 CONTROLLING ACCESS AND STAY TIME .............................. 12.5-1 12.5.5 COMBINED LICENSE INFORMATION....................................... 12.5-1 12-i Revision 3

TABLE OF CONTENTS (Continued)

Section Title Page APP. 12AA RADIATION PROTECTION PROGRAM DESCRIPTION ... 12AA-1 12AA.5.4.14 Groundwater Monitoring Program ....................................... 12AA-3 12AA.5.4.15 Record of Operational Events of Interest for Decommissioning ................................................................ 12AA-4 12-ii Revision 3

LIST OF TABLES Number Title 12.4-201 Construction Worker Dose Comparison to 10 CFR 20.1301 Criteria 12AA-201 Very High Radiation Areas (VHRA) 12-iii Revision 3

LIST OF FIGURES Number Title 12.3-201 Radiation Zones, Normal Operations/Shutdown Annex Building, Elevation 100-0 & 107-2 12.3-202 Radiation Zones, Post-Accident Annex Building, Elevation 100-0 & 107-2 12.3-203 Radiological Access Controls, Normal Operations/

Shutdown Annex Building, Elevation 100-0 & 107-2 12-iv Revision 3

CHAPTER 12 RADIATION PROTECTION 12.1 ASSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS-LOW-AS-REASONABLY ACHIEVABLE (ALARA)

This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

COL 12.1-1 This section incorporates by reference NEI 07-08A, Generic FSAR Template Guidance for Ensuring That Occupational Radiation Exposures Are as Low as Is Reasonably Achievable (ALARA), Revision 0. See Table 1.6-201. ALARA practices are developed in a phased milestone approach as part of the procedures necessary to support the Radiation Protection Program.

Table 13.4-201 describes the major milestones for ALARA procedures development and implementation.

Revise the last sentence of NEI 07-08A Subsection 12.1.2 to read:

ALARA procedures are established, implemented, maintained and reviewed consistent with 10 CFR 20.1101 and the quality assurance criteria described in Part III of the Quality Assurance Program Description, which is discussed in Section 17.5.

Add the following information at the end of DCD Subsection 12.1.2.4:

12.1.2.4.3 Equipment Layout SUP 12.1-1 A video record of the equipment layout in areas where radiation fields are expected to be high following operations may be used to assist in ALARA planning and to facilitate decommissioning.

12.1.3 COMBINED LICENSE INFORMATION COL 12.1-1 This COL item is addressed in NEI 07-08A and Appendix 12AA.

12.1-1 Revision 3

12.2 RADIATION SOURCES This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

12.2.1.1.10 Miscellaneous Sources Add the following information at the end of DCD Subsection 12.2.1.1.10:

COL 12.2-1 Licensed sources containing byproduct, source, and special nuclear material that warrant shielding design consideration meet the applicable requirements of 10 CFR Parts 20, 30, 31, 32, 33, 34, 40, 50, and 70.

There are byproduct and source materials with known isotopes and activity manufactured for the purpose of measuring, checking, calibrating, or controlling processes quantitatively or qualitatively.

These sources include but are not limited to:

• Sources in field monitoring equipment.

• Sources in radiation monitors to maintain a threshold sensitivity.

• Sources used for radiographic operations.

• Depleted uranium slabs used to determine beta response and correction factors for portable monitoring instrumentation.

• Sources used to calibrate and response check field monitoring equipment (portable and fixed).

• Liquid standards and liquids or gases used to calibrate and verify calibration of laboratory counting and analyzing equipment.

• Radioactive waste generated by the use of radioactive sources.

Specific details of these sources are maintained in a database on-site following procurement. This database, at a minimum, contains the following information:

• Isotopic composition

• Location in the plant 12.2-1 Revision 3

• Source strength

• Source geometry Written procedures are established and implemented that address procurement, receipt, inventory, labeling, leak testing, surveillance, control, transfer, disposal, storage, issuance and use of these radioactive sources. These procedures are developed in accordance with the radiation protection program to comply with 10 CFR Parts 19 and 20. A supplementary warning symbol is used in the presence of large sources of ionizing radiation consistent with the guidance in Regulatory Issue Summary (RIS) 2007-03.

Sources maintained on-site for instrument calibration purposes are shielded while in storage to keep personnel exposure ALARA. Sources used to service or calibrate plant instrumentation are also routinely brought on-site by contractors.

Radiography is performed by the licensed utility group or licensed contractors.

These sources are maintained and used in accordance with the provisions of the utility group's or contractors license. Additional requirements and restrictions may apply depending on the type of source, use, and intended location of use. If the utility group or contractor source must be stored on-site, designated plant personnel must approve the storage location, and identify appropriate measures for maintaining security and personnel protection.

12.2.3 COMBINED LICENSE INFORMATION COL 12.2-1 This COL item is addressed in Subsection 12.2.1.1.10.

12.2-2 Revision 3

12.3 RADIATION PROTECTION DESIGN FEATURES This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

DEP 18.8-1 12.3.1.2 Radiation Zoning and Access Control Add the following information at the end of the second paragraph of DCD Subsection 12.3.1.2.

Figure 12.3-201, Figure 12.3-202, and Figure 12.3-203 replace DCD Figure 12.3-1 (Sheet 11), DCD Figure 12.3-2 (Sheet 11), and DCD Figure 12.3-3 (Sheet 11), respectively, to reflect the relocation of the Operations Support Center.

12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION Add the following text to the end of DCD Subsection 12.3.4.

COL 12.3-2 Procedures detail the criteria and methods for obtaining representative measurement of radiological conditions, including in-plant airborne radioactivity concentrations in accordance with applicable portions of 10 CFR Part 20 and consistent with the guidance in Regulatory Guides 1.21-Appendix A, 8.2, 8.8, and 8.10. Additional discussion of radiological surveillance practices is included in the radiation protection program description provided in Appendix 12AA.

Surveillance requirements are determined by the functional manager in charge of radiation protection based on actual or potential radiological conditions encountered by personnel and the need to identify and control radiation, contamination, and airborne radioactivity. These requirements are consistent with the operational philosophy in Regulatory Guide 8.10. Frequency of scheduled surveillance may be altered by permission of the functional manager in charge of radiation protection or their designee. Radiation Protection periodically provides cognizant personnel with survey data that identifies radiation exposure gradients in area resulting from identified components. This data includes recent reports, with survey data, location and component information.

12.3-1 Revision 3

The following are typical criteria for frequencies and types of surveys:

Job Coverage Surveys

• Radiation, contamination, and/or airborne surveys are performed and documented to support job coverage.

• Radiation surveys are sufficient in detail for Radiation Protection to assess the radiological hazards associated with the work area and the intended/

specified work scope.

• Surveys are performed commensurate with radiological hazard, nature and location of work being conducted.

• Job coverage activities may require surveys to be conducted on a daily basis where conditions are likely to change.

Radiation Surveys

• Radiation surveys are performed at least monthly in any radiological controlled area (RCA) where personnel may frequently work or enter.

Survey frequencies may be modified by the functional manager in charge of radiation protection as previously noted.

• Radiation surveys are performed prior to or during entry into known or suspected high radiation areas for which up to date survey data does not exist.

• Radiation surveys are performed prior to work involving highly contaminated or activated materials or equipment.

• Radiation surveys are performed at least semiannually in areas outside the RCA. Areas to be considered include shops, offices, and storage areas.

• Radiation surveys are performed to support movement of highly radioactive material.

• Neutron radiation surveys are performed when personnel may be exposed to neutron emitting sources.

Contamination Surveys

• Contamination surveys are performed at least monthly in any RCA where personnel may frequently work or enter. Survey frequencies may be modified by the functional manager in charge of radiation protection as previously noted.

12.3-2 Revision 3

• Contamination surveys are performed during initial entry into known or suspected contamination area(s) for which up to date survey data does not exist.

• Contamination surveys are performed at least daily at access points, change areas, and high traffic walkways in RCAs that contain contaminated areas. Area access points to a High Radiation Area or Very High Radiation Area are surveyed prior to or upon access by plant personnel or if access has occurred.

• Contamination surveys are performed at least semiannually in areas outside the RCA. Areas to be considered include shops, offices, and storage areas.

• A routine surveillance is conducted in areas designated by the functional manager in charge of radiation protection or their designee likely to indicate alpha radioactivity. If alpha contamination is identified, frequency and scope of the routine surveillance is increased.

Airborne Radioactivity Surveys

• Airborne radioactivity surveys are performed during any work or operation in the RCA known or suspected to cause airborne radioactivity (e.g.,

grinding, welding, burning, cutting, hydrolazing, vacuuming, sweeping, use of compressed air, using volatiles on contaminated material, waste processing, or insulation).

• Airborne radioactivity surveys are performed during a breach of a radioactive system, which contains or is suspected of containing significant levels of contamination.

• Airborne radioactivity surveys are performed during initial entry (and periodically thereafter) into any known or suspected airborne radioactivity area.

• Airborne radioactivity surveys are performed immediately following the discovery of a significant radioactive spill or spread of radioactive contamination, as determined by the functional manager in charge of radiation protection.

• Airborne radioactivity surveys are performed daily in occupied radiological controlled areas where the potential for airborne radioactivity exists, including containment.

• Airborne radioactivity surveys are performed any time respiratory protection devices, alternative tracking methods such as derived air concentration-hour (DAC-hr), and/or engineering controls are used to control internal exposure.

12.3-3 Revision 3

• Airborne radioactivity surveys are performed using continuous air monitors (CAMs) for situations in which airborne radioactivity levels can fluctuate and early detection of airborne radioactivity could prevent or minimize inhalations of radioactivity by workers. Determination of air flow patterns are considered for locating air samplers.

• Airborne radioactivity surveys are performed prior to use and monthly during use on plant service air systems used to supply air for respiratory protection to verify the air is free of radioactivity.

• Tritium sampling is performed near the spent fuel pit when irradiated fuel is in the pit and other areas of the plant where primary system leaks occur and tritium is suspected.

Appropriate counting equipment is used based on the sample type and the suspected identity of the radionuclides for which the sample is being done. Survey results are documented, retrievable, and processed per site document control and records requirements consistent with Regulatory Guide 8.2. Completion of survey documentation includes the update of room/area posting maps and revising area or room postings and barricades as needed.

Air samples indicating activity levels greater than a procedure specified percentage of DAC are forwarded to the radiochemistry laboratory for isotopic analysis. Samples which cannot be analyzed on-site are forwarded to an off-site laboratory or a contractor for analysis; or, the DAC percentage may be hand calculated using appropriate values from 10 CFR Part 20, Appendix B.

The responsible radiation protection personnel review survey documentation to evaluate if surveys are appropriate and obtained when required, records are complete and accurate, and adverse trends are identified and addressed.

An in-plant radiation monitoring program maintains the capability to accurately determine the airborne iodine concentration in areas within the facility where personnel may be present under accident conditions. This program includes the training of personnel, procedures for monitoring, and provisions for maintenance of sampling and analysis equipment consistent with Regulatory Guides 1.21 (Appendix A) and 8.8. Training and personnel qualifications are discussed in Appendix 12AA.

A portable monitor system meeting the requirements of NUREG-0737, Item III.D.3.3, is available. The system uses a silver zeolite or charcoal iodine sample cartridge and a single-channel analyzer. The use of this portable monitor is incorporated in the emergency plan implementing procedures. The portable monitor is part of the in-plant radiation monitoring program. It is used to determine the airborne iodine concentration in areas where plant personnel may be present during an accident. Accident monitoring instrumentation complies with applicable parts of 10 CFR Part 50, Appendix A.

12.3-4 Revision 3

Sampling cartridges can be removed to a low background area for further analysis. These cartridge samples can be purged of any entrapped noble gases, when necessary, prior to being analyzed.

12.3.5.1 Administrative Controls for Radiological Protection COL 12.3-1 This COL Item is addressed in Subsection 12.5.4 and Appendix 12AA.

12.3.5.2 Criteria and Methods for Radiological Protection COL 12.3-2 This COL Item is addressed in Subsection 12.3.4.

12.3.5.3 Groundwater Monitoring Program COL 12.3-3 This COL Item is addressed in Appendix 12AA.

12.3.5.4 Record of Operational Events of Interest for Decommissioning COL 12.3-4 This COL Item is addressed in Appendix 12AA.

12.3-5 Revision 3

12.4 DOSE ASSESSMENT This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

Add the following new subsections after DCD Subsection 12.4.1.8:

SUP 12.4-1 12.4.1.9 Dose to Construction Workers This section evaluates the potential radiological dose impacts to construction workers at the Bellefonte Nuclear Station, Units 3 and 4 (BLN) resulting from the operation of the BLN Unit 3. Since a portion of the Unit 4 construction period overlaps operation of Unit 3, construction workers at Unit 4 would be exposed to direct radiation and gaseous radioactive effluents from Unit 3. Doses to construction workers during construction of Unit 3 are not evaluated since the only radiation sources prior to the start-up of Unit 3 are background sources.

12.4.1.9.1 Site Layout The BLN power block areas are shown on FSAR Figure 2.1-201. Construction activity for Unit 4 is outside the protected area for Unit 3 but inside the owner controlled area.

12.4.1.9.2 Radiation Sources Construction workers at a new facility on the BLN site are not be exposed to any radiation sources until Unit 3 becomes operational. Workers constructing Unit 4 may be exposed to direct radiation and to gaseous radioactive effluents emanating from the routine operation of Unit 3. Radiation dose to construction workers is from direct radiation and from airborne effluents from BLN Unit 3, and from background radiation.

The radiation exposure at the site boundary is considered in DCD Section 12.4.2.

As stated in that section, direct radiation from the containment and other plant buildings is negligible. Additionally, there is no contribution from refueling water since the refueling water is stored inside the containment instead of in an outside storage tank.

Small quantities of monitored airborne effluents are normally released through the plant vent or the turbine building vent. The plant vent provides the release path for containment venting releases, auxiliary building ventilation releases, annex building releases, radwaste building releases, and gaseous radwaste system discharge. The turbine building vents provide the release path for the condenser air removal system, gland seal condenser exhaust and the turbine building ventilation releases. The ventilation system is described in DCD Section 9.4. The 12.4-1 Revision 3

expected radiation sources (nuclides and activities) in the gaseous effluents are listed in DCD Table 11.3-3.

Exposure of Unit 4 construction workers to radioactive liquid effluents is not evaluated because the discharge structure and blowdown piping is completed during Unit 3 construction. The only exposure of Unit 4 construction workers to liquid effluents is due to the tie-in of Unit 4 pipeline. The exposure from this activity is minimal.

12.4.1.9.3 Construction Worker Dose Estimates The determination of construction worker dose due to Unit 3 operation depends on the airborne effluent release and the atmospheric transport to the worker location. The atmospheric dispersion calculation used the guidance provided in Regulatory Guide 1.111, meteorological data for the year beginning April 1, 2006 and ending March 31, 2007, and downwind distances to the construction worker locations. The XOQDOQ computer code (NUREG/CR-2919) was used to determine the /Q and D/Q values for the nearest location along the Unit 3 protected area fence in each direction as well as the nearest point of the Unit 4 shield building construction area.

Construction worker doses are conservatively estimated using the following information:

• The estimated maximum dose rate for each pathway.

- External exposure to contaminated ground.

- External exposure to noble gas radionuclides in the airborne plume.

- Inhalation of air.

• A construction worker exposure time of 2080 hours0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br /> per year.

• A peak loading of 2100 construction workers per year for Unit 4 construction.

The use of 2080 hours0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br /> assumes the worker works 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week for 52 weeks per year.

The methodology used to calculate the doses to construction workers from normal effluent releases complies with the guidance provided in Regulatory Guide 1.109.

Construction worker doses were estimated by use of GASPAR computer code (NUREG/CR-4653). The Total Effective Dose Equivalent (TEDE), which is the sum of the Deep Dose Equivalent (DDE) and the Committed Effective Dose Equivalent (CEDE), was determined based on the GASPAR results. The annual 12.4-2 Revision 3

TEDE dose was corrected for the actual time the construction workers are onsite by multiplying by a ratio of hours worked per year to hours in a year.

12.4.1.9.4 Compliance with Dose Regulations BLN Unit 4 construction workers are, for the purposes of radiation protection, members of the general public. This means that the dose to the individual does not exceed 100 mrem per year, the limit for a member of the public. The construction workers do not deal with radiation sources.

Dose limits to the public are provided in 10 CFR 20.1301 and 10 CFR 20.1302.

Because the construction workers are considered members of the public, the requirements of 10 CFR 20.1201 through 20.1204 do not apply.

The 10 CFR 20.1301 limits annual doses from licensed operations to individual members of the public to 100 mrem TEDE. In addition, the dose from external sources to unrestricted areas must be less than 2 mrem in any one hour. This applies to the public both outside and inside access controlled areas. The maximum dose rates are given in Table 12.4-201. For an occupational year, dose at the Unit 4 construction area is 0.54 mrem TEDE. The maximum dose anywhere onsite that is accessible to a construction worker is 7.1 mrem per year in the NNE sector at the Unit 3 fence line. This assumes the worker stands at this point on the fence line for all working hours for the entire year. This value is less than the limits specified for members of the public. Therefore, construction workers can be considered to be members of the general public and do not require radiation monitoring.

12.4.1.9.5 Collective Doses to BLN Unit 4 Workers The collective dose is the sum of all doses received by all workers. It is a measure of population risk. The total worker collective dose is 1.13 person-rem. This estimate is based upon the construction workforce of 2100 and assumes 2080 hours0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br /> per year occupancy for each worker.

SUP 12.4-1 12.4.1.9.6 Operating Unit Radiological Surveys The operating unit conducts radiological surveys in the unrestricted and controlled area and radiological surveys for radioactive materials in effluents discharged to unrestricted and controlled areas in implementing 10 CFR 20.1302. These surveys demonstrate compliance with the dose limits of 10 CFR 20.1301 for construction workers.

12.4-3 Revision 3

TABLE 12.4-201 CONSTRUCTION WORKER DOSE COMPARISON TO 10 CFR 20.1301 CRITERIA Dose Limits(1)

Type of Dose (TEDE) Estimated Dose(2)

Annual total effective dose equivalent 100 mrem 0.54 mrem Maximum dose in any hour 2 mrem 2.6E-04 mrem Notes:

1. 10 CFR 20.1301 criteria.

2. Estimated dose is at Unit 4 shield building construction area. Total body dose calculated using the methodology in Regulatory Guide 1.109.

12.4-4 Revision 3

12.5 HEALTH PHYSICS FACILITIES DESIGN This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

12.5.2.2 Facilities Revise the first sentence of DCD Subsection 12.5.2.2 to read:

DEP 18.8-1 The ALARA briefing room is located off the main corridor immediately beyond the main entry to the annex building.

12.5.4 CONTROLLING ACCESS AND STAY TIME Add the following text to the end of DCD Subsection 12.5.4.

COL 12.3-1 A closed circuit television system may be installed in high radiation areas to allow remote monitoring of individuals entering high radiation areas by personnel qualified in radiation protection procedures.

12.5.5 COMBINED LICENSE INFORMATION COL 12.5-1 This COL Item is addressed in Appendix 12AA.

12.5-1 Revision 3

Add the following Appendix after Section 12.5 of the DCD.

APPENDIX 12AA RADIATION PROTECTION PROGRAM DESCRIPTION COL 12.1-1 This appendix incorporates by reference NEI 07-03A, Generic FSAR Template COL 12.3-1 Guidance for Radiation Protection Program Description. See Table 1.6-201. The COL 12.5-1 numbering of NEI 07-03A is revised from 12.5# to 12AA.5# through the document, with the following revisions and additions as indicated by strikethroughs and underlines. Table 13.4-201 provides milestones for radiation protection program implementation.

Revise bullet number 3 of NEI 07-03A Section 12.5 as follows:

3. Prior to initial loading of fuel in the reactor, all of the radiation program functional areas described in Appendix 12AA Section 12.5 will be fully implemented, with the exception of the organization, facilities, equipment, instrumentation, and procedures necessary for transferring, transporting or disposing of radioactive materials in accordance with 10 CFR Part 20, Subpart K, and applicable requirements in 10 CFR Part 71. In addition, the position of radiation protection manager (as described in sSection 13.112.5.2.3) will be filled and at least one (1) radiation protection technician for each operating shift, selected, trained, and qualified consistent with the guidance in Regulatory Guide 1.8, will be onsite and on duty when fuel is initially loaded in the reactor, and thereafter, whenever fuel is in the reactor.

Revise the first paragraph of NEI 07-03A Subsection 12.5.2 as follows:

Qualification and training criteria for site personnel are consistent with the guidance in Regulatory Guide 1.8 and are described in FSAR Chapter 13. Specific radiation protection responsibilities for key positions within the plant organization are described in Section 13.1 below.

Subsections 12.5.2.1 through 12.5.2.5 of NEI 07-03A are not incorporated into Appendix 12AA.

Subsection 12.5.3.1 of NEI 07-03A is not incorporated into Appendix 12AA.

Facilities are described in DCD Subsection 12.5.2.2.

Add the following text after the first paragraph of NEI 07-03A Subsection 12.5.3.3.

If circumstances arise in which NIOSH tested and certified respiratory equipment is not used, compliance with 10 CFR 20.1703(b) and 20.1705 is maintained.

12AA-1 Revision 3

The following headings (and associated material) in Subsection 12.5.4.2 of NEI 07-03A are described in DCD Subsection 12.5.3, and are therefore not incorporated into Appendix 12AA:

• Radwaste Handling

• Spent Fuel Handling

• Normal Operation

• Sampling Add the following text after the second paragraph of NEI 07-03A Subsection 12.5.4.4.

COL 12.3-1 Table 12AA-201 identifies plant areas designated as Very High Radiation Areas (VHRAs), lists corresponding plant layout drawings showing the VHRA in DCD Section 12.3, specifies the condition under which the area is designated VHRA, identifies the primary source of the VHRA, and summarizes the frequency of access and reason for access. VHRAs are listed as Radiation Zone IX, which corresponds to a dose rate greater than 500 rad/hr.

In each of the VHRAs, with the exception of the Reactor Vessel Cavity and Delay-Bed / Guard-Bed Compartment, the primary radioactive source is transient (such as fuel passing through the transfer tube), removable (such as resin in the demineralizers), or can be relocated. When the primary source is removed, the dose rate in each of these areas will be less than Zone IX and, in effect, the area will no longer be a VHRA. With planning, the need for human entrance to a VHRA when the primary source is present can be largely or entirely avoided.

In addition to the access control requirements for high radiation areas, the following control measures are implemented to control access to very high radiation areas in which radiation levels could be encountered at 500 rads or more in one hour at one meter from a radiation source or any surface through which the radiation penetrates:

- Sign(s) conspicuously posted stating GRAVE DANGER, VERY HIGH RADIATION AREA.

- Area is locked. Each lock shall have a unique core. The keys shall be administratively controlled by the functional manager in charge of radiation protection as described in Section 13.1.

- Plant Managers (or designee) approval required for entry.

12AA-2 Revision 3

- Radiation Protection personnel shall accompany person(s) making the entry. Radiation Protection personnel shall assess the radiation exposure conditions at the time of the entry.

A verification walk down will be performed with the purpose of verifying barriers to the Very High Radiation Areas in the final design of the facility are consistent with Regulatory Guide 8.38 guidance as part of the implementation of the Radiation Protection and ALARA programs on the schedule identified in Table 13.4-201.

Revise the third paragraph of NEI 07-03A Subsection 12.5.4.7 as follows.

COL 12.1-1 As described in Sections 12.1, 12.5.1Appendix 12AA and 12.5.2 13.1, COL 12.3-1 management policy is established, and organizational responsibilities and COL 12.5-1 authorities are assigned to implement an effective program for maintaining occupational radiation exposures ALARA. Procedures are established and implemented that are in accordance with 10 CFR 20.1101 and consistent with the guidance in Regulatory Guides 8.8 and 8.10. Examples of such procedures include the following:

Add the following text after the last bullet of NEI 07-03A Subsection 12.5.4.8.

COL 12.5-1 This subsection adopts NEI 08-08A (Reference 201), for a description of the operational and programmatic elements and controls that minimize contamination of the facility, site, and the environment, to meet the requirements of 10 CFR 20.1406.

Revise the first paragraph of Subsection 12.5.4.12 of NEI 07-03A to read:

COL 12.5-1 The radiation protection program and procedures are established, implemented, maintained, and reviewed consistent with the 10 CFR 20.1101 and the quality assurance criteria described in Part III of the Quality Assurance Program Description described in Section 17.5.

Add the following Subsection to the information incorporated from NEI 07-03A.

COL 12.3-3 12AA.5.4.14 Groundwater Monitoring Program A groundwater monitoring program beyond the normal radioactive effluent monitoring program is developed. If necessary to support this groundwater 12AA-3 Revision 3

monitoring program, design features will be installed during the plant construction process. Areas of the site to be specifically considered in this groundwater monitoring program are (all directions based on plant standard):

• West of the auxiliary building in the area of the fuel transfer canal.

• West and south of the radwaste building.

• East of the auxiliary building rail bay and the radwaste building truck doors This subsection adopts NEI 08-08A (Reference 201) for the Groundwater Monitoring Program description.

Add the following Subsection to the information incorporated from NEI 07-03A.

COL 12.3-4 12AA.5.4.15 Record of Operational Events of Interest for Decommissioning This subsection adopts NEI 08-08A (Reference 201) for discussion of recordkeeping practices important to decommissioning.

Revise the REFERENCES section of NEI 07-03A, Reference 8, as follows:

8. Regulatory Guide 1.97, Revision 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident. 4, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants.

Add the following reference to the NEI 07-03A REFERENCES.

201. NEI 08-08A, Generic FSAR Template Guidance for Life Cycle Minimization of Contamination, Revision 0, October 2009.

12AA-4 Revision 3

Part 2, FSAR TABLE 12AA-201 (Sheet 1 of 3) 12.3-1 VERY HIGH RADIATION AREAS (VHRA)

Room VHRA Location DCD Primary Source(s) VHRA Frequency of Number Figure 12.3-1, Conditional Access to VHRA Sheet No. Notes Areas While VHRA Conditions Exist 11105 Reactor Vessel Cavity 3, 4, 5 Neutron activation of the material Note 1 None Required in and around the cavity during reactor operations, such as the concrete shield walls and the reactor insulation 12151 Spent Fuel Pool Cooling 3 Resin in vessels Notes 6, 8 None Required System / Liquid Radwaste System Demineralizer/Filter room (Inside Wall) 12153 Delay-Bed/ Guard-Bed 3 Activated carbon holding Note 10 None Required Compartment radioactive gases 12371 Filter-Storage Area 6, 7 Spent filter cartridges Notes 4, 6, 7 None Required 12372 Resin Transfer Pump/ 6 Spent resin in lines Note 6 None Required Valve Room 12373 Spent-Resin Tank Room 6 Spent resin in tanks Note 6 None Required 12AA-5 Revision 3

Part 2, FSAR TABLE 12AA-201 (Sheet 2 of 3) 12.3-1 VERY HIGH RADIATION AREAS (VHRA)

Room VHRA Location DCD Primary Source(s) VHRA Frequency of Number Figure 12.3-1, Conditional Access to VHRA Sheet No. Notes Areas While VHRA Conditions Exist 12374 Waste Disposal 6 Spent resin in vault Note 6 None Required Container Area 12463 Cask Loading Pit 6 Spent fuel Notes 2, 6 None Required 12563 Spent Fuel Pit 5, 6 Spent fuel Note 6 None Required Fuel Transfer Areas 12564 Fuel Transfer Tube 6 Fuel in transit Notes 2, 5, 9 None Required 11205 Reactor Vessel Nozzle 5 Fuel in transit Notes 2, 3, 9 None Required Area 11504 Refueling Cavity 6 Fuel in transit Notes 2, 3, 9 None Required 12AA-6 Revision 3

Part 2, FSAR TABLE 12AA-201 (Sheet 3 of 3) 12.3-1 VERY HIGH RADIATION AREAS (VHRA)

Room VHRA Location DCD Primary Source(s) VHRA Frequency of Number Figure 12.3-1, Conditional Access to VHRA Sheet No. Notes Areas While VHRA Conditions Exist Notes

1. VHRA during full power operation; less than 10 Rem/hr 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after plant shutdown.

2. During underwater spent fuel transfer operations, this area can be as high as VHRA.

3. During underwater reactor internals transfers/ storage, this area can be as high as VHRA.

4. During spent resin waste disposal container transfer or loading, this area can be as high as VHRA. The contact dose rate of spent resin containers can be greater than 1000 Rem/hr.

5. Discussion about the Spent Fuel Transfer Canal and Tube Shielding is provided in DCD Subsection 12.3.2.2.9.

6. Source is transient, removable, or can be relocated.

7. VHRA when hatch is removed during spent resin container handling operation.

8. In the event that the room does need to be accessed for maintenance or other reasons, temporary shielding is put in place and the resin is removed from the vessels. These measures reduce exposure rates in the room, such that this room is no longer a VHRA. Remote handling is used for any tasks that require the opening of the access hatch in the ceiling of this room when media is present.

9. These areas have no planned reasons for entry and are only classified as VHRAs during periods of fuel movement.

In the event that these rooms do need to be accessed to repair the Fuel-Transfer System, Fuel Transfer Tube Gate Valve, or other components, it is done during a non-fuel movement time. This keeps the dose received by the worker as low as reasonably achievable.

10. Inspection of the equipment in this room, when required, is done using remote viewing equipment. Two plugs between Room 12153 and 12155 contain instruments and the plugs are expected to be removed every 12 to 18 months for performance of maintenance. Administrative procedures are implemented to protect workers pursuant to Regulatory Guide 8.38.

12AA-7 Revision 3