ML17061A157: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 30: Line 30:
CALLAWAY - SPLIST OF TABLES (Continued)NumberTitle12.0-vRev. OL-1412/0412.4-3Average Occupational Radiation Exposure (Man-Rem Dose) Per PWR Unit for the Period 1969-1977 (Ref. 1, 2, 3)12.4-4Average Occupational Radiation Exposure (Man-Rem Dose) Based on PWR Plant Age (Ref 1, 2, 3)12.4-5Distribution of the Number of Personnel (>100 Milirem
CALLAWAY - SPLIST OF TABLES (Continued)NumberTitle12.0-vRev. OL-1412/0412.4-3Average Occupational Radiation Exposure (Man-Rem Dose) Per PWR Unit for the Period 1969-1977 (Ref. 1, 2, 3)12.4-4Average Occupational Radiation Exposure (Man-Rem Dose) Based on PWR Plant Age (Ref 1, 2, 3)12.4-5Distribution of the Number of Personnel (>100 Milirem
ÚYr) According to Work Function12.4-6Distribution of Personnel (>100 Millirem/Yr) According to Employee Category12.4-7Percentages of Personnel Dose by Work Function (Ref3)12.4-8Annual Occupational Exposures for Various PWR Vendor's Units (REF. 1,2,3)12.4-9Average Annual Occupational Exposure and Power for Individual PWRS (REF 1, 2, 3)12.4-10Average Individual Exposure Based on PWR Plant Age (REF. 1,2,3) 12.4-11Cumulative Average of Annual Exposure by YearS of Operation - PWRS (REF. 1, 2, 3)12.4-12Estimates of Occupancy Times in Plant Radiation Areas and Gamma Doses to Plant Personnel12.4-13Distribution of Direct Radiation Man-rem Doses According to Work Functions12.4-14Annual Occupancy in Plant Areas Containing Airborne Radioactivity12.4-15Doses to plant Personnel Caused by Airborne Radioactivity CALLAWAY - SP12.0-viRev. OL-1412/04LIST OF FIGURESNumberTitle12.2-1Old Steam Generator Storage Facility General Layout12.3-1Typical Valve Compartment Arrangement12.3-2Radiation Zones for Normal Operation 12.3-3Control Room Isometric12.3-4Decontamination System12.4-1Cumulative Average of Annual Exposure by Years of Operation - PWRs (Table 12.4-11)
ÚYr) According to Work Function12.4-6Distribution of Personnel (>100 Millirem/Yr) According to Employee Category12.4-7Percentages of Personnel Dose by Work Function (Ref3)12.4-8Annual Occupational Exposures for Various PWR Vendor's Units (REF. 1,2,3)12.4-9Average Annual Occupational Exposure and Power for Individual PWRS (REF 1, 2, 3)12.4-10Average Individual Exposure Based on PWR Plant Age (REF. 1,2,3) 12.4-11Cumulative Average of Annual Exposure by YearS of Operation - PWRS (REF. 1, 2, 3)12.4-12Estimates of Occupancy Times in Plant Radiation Areas and Gamma Doses to Plant Personnel12.4-13Distribution of Direct Radiation Man-rem Doses According to Work Functions12.4-14Annual Occupancy in Plant Areas Containing Airborne Radioactivity12.4-15Doses to plant Personnel Caused by Airborne Radioactivity CALLAWAY - SP12.0-viRev. OL-1412/04LIST OF FIGURESNumberTitle12.2-1Old Steam Generator Storage Facility General Layout12.3-1Typical Valve Compartment Arrangement12.3-2Radiation Zones for Normal Operation 12.3-3Control Room Isometric12.3-4Decontamination System12.4-1Cumulative Average of Annual Exposure by Years of Operation - PWRs (Table 12.4-11)
CALLAWAY - SP12.1-1Rev. OL-155/0612.1ENSURINGTHATOCCUPATIONALRADIATIONEXPOSURESARE  ASLOWASREASONABLYACHIEVABLE12.1.1POLICY CONSIDERATIONSThe policy of Union Electric, through guidance from Regulatory Guides 1.8, 8.8, and 8.10 and 10 CFR Part 20, includes a commitment to ensure that occupational radiation exposures (ORE) are as low as reasonably achievable (ALARA). Union Electric will establish radiation protection practices for all applicable plant activities and will provide a qualified radiation protection organization to accomplish this goal. Utility management recognizes and will emphasize the importance of each individual's responsibilities to maintain ORE-ALARA. Section 12.1.1 of the Site Addendum identifies who has the ultimate responsibility for establishing an ALARA philosophy for operating, maintaining, refueling, and testing. The operating organizational structure and responsibilities are discussed in Section 13.1.2
CALLAWAY - SP12.1-1Rev. OL-155/0612.1ENSURINGTHATOCCUPATIONALRADIATIONEXPOSURESARE  ASLOWASREASONABLYACHIEVABLE12.1.1POLICY CONSIDERATIONSThe policy of Union Electric, through guidance from Regulatory Guides 1.8, 8.8, and 8.10 and 10 CFR Part 20, includes a commitment to ensure that occupational radiation exposures (ORE) are as low as reasonably achievable (ALARA). Union Electric will establish radiation protection practices for all applicable plant activities and will provide a qualified radiation protection organization to accomplish this goal. Utility management recognizes and will emphasize the importance of each individual's responsibilities to maintain ORE-ALARA. Section 12.1.1 of the Site Addendum identifies who has the ultimate responsibility for establishing an ALARA philosophy for operating, maintaining, refueling, and testing. The operating organizational structure and responsibilities are discussed in Section 13.1.2. Section 12.1.1 of the Site Addendum also indicates the responsibility of key personnel in implementation of ORE-ALARA policies. 12.1.2DESIGN CONSIDERATIONSA major objective in the design of the Callaway power block has been to limit the potential radiation exposures of operating personnel. Union Electric has emphasized this objective to its lead architect engineer (Bechtel) from the outset of the project, through its reviews of project design criteria and frequent design reviews as the detailed design of the power block has evolved. Specific design considerations and the guidelines employed in developing the Callaway design are presented in Sections 12.3.1.1 and 12.3.3.3. Estimated occupational radiation doses are given in Section 12.4. To ensure that ORE are ALARA, significant elements of the design program have been implemented as described in the following sections. 12.1.2.1PlantDesignThe design engineers and first level supervisors assigned by Bechtel to the Callaway project have, in most cases, performed similar design work on other nuclear power plants. Through this experience, they have developed sensitivity to and knowledge of radiation protection aspects of design, which have been applied to Callaway. Bechtel design engineers are also made aware of other operating experience through NRC Licensee Event Reports, NRC Operating Experience Bulletin Information Reports, and Bechtel-generated problem alert reports. The Bechtel designers have, where practical, followed the design guidelines of NRC Regulatory Guide 8.8. The designers have also followed recommendations of the NSSS supplier (Westinghouse). Many of these recommendations are available in documented form, in the Westinghouse information packages and in Reference1. Other CALLAWAY - SP12.1-2Rev. OL-155/06recommendations developed from discussions with Westinghouse. Westinghouse representatives met with Bechtel designers, as appropriate. Design of NSSS equipment, within Westinghouse's scope of supply, to ALARA objectives is described in Reference1. In the early stages of the plant design, calculations were performed to quantify potential concentrations of radioactivity throughout process systems and buildings of the power block. These calculations made use of measured data from operating plants, and also followed the methodology of NRC Regulatory Guide 1.112. Systems designs and equipment specifications have been influenced by these assessments (see Section12.1.2.5). Additionally, radiation dose rates have been estimated throughout the power block. Specifications for equipment and/or systems to be purchased reflect the objective to keep ORE-ALARA. For example, the specifications for the solid radwaste system state that "special consideration shall be given to eliminate points where radioactive materials may tend to accumulate."  The specifications also require provisions for remote flushing and rinsing of portions of equipment that contain radioactive material and state that "special design consideration shall be given to minimizing operator exposure to radiation during the maintenance of equipment."12.1.2.2Scale-ModelProgramThe use of scale models is an innovative aspect of the Callaway project that has proven effective for design and design review. Three classes of scale models have been used with respect to radiation protection:  (1) preliminary design model, (2) study models, and (3) final design model. The preliminary design model (3/8 inch = 1 foot) was, in effect, a three-dimensional layout tool. Model builders constructed the main elements of the buildings (walls, floors, columns, etc.) and set in place major equipment (tanks, pumps, motors, etc.). Plant design engineers then routed piping, electrical cable trays, and HVAC ducts, and located valves and valve actuators on the model. The three-dimensional aspects of the model, compared to conventional layout drawings, placed radiation protection, among other considerations, into sharp focus for both design engineers and reviewers and facilitated evaluation of design alternatives, such as choosing the best valve placement. Prior to completion of the preliminary design models, several reviews by utility and SNUPPS staff were held, as discussed further in Section 12.1.2.4. Upon completion of the preliminary design model, the approved layouts were committed to paper as design drawings. Study models were less formal in concept than the preliminary design and final design models and were constructed to evaluate specific design features or alternatives. For example, a study model was constructed of portions of the auxiliary building in order to evaluate the arrangement of radioactive demineralizers. The results of the review of this model are discussed in Section 12.1.2.5.
. Section 12.1.1 of the Site Addendum also indicates the responsibility of key personnel in implementation of ORE-ALARA policies. 12.1.2DESIGN CONSIDERATIONSA major objective in the design of the Callaway power block has been to limit the potential radiation exposures of operating personnel. Union Electric has emphasized this objective to its lead architect engineer (Bechtel) from the outset of the project, through its reviews of project design criteria and frequent design reviews as the detailed design of the power block has evolved. Specific design considerations and the guidelines employed in developing the Callaway design are presented in Sections 12.3.1.1 and 12.3.3.3. Estimated occupational radiation doses are given in Section 12.4
. To ensure that ORE are ALARA, significant elements of the design program have been implemented as described in the following sections. 12.1.2.1PlantDesignThe design engineers and first level supervisors assigned by Bechtel to the Callaway project have, in most cases, performed similar design work on other nuclear power plants. Through this experience, they have developed sensitivity to and knowledge of radiation protection aspects of design, which have been applied to Callaway. Bechtel design engineers are also made aware of other operating experience through NRC Licensee Event Reports, NRC Operating Experience Bulletin Information Reports, and Bechtel-generated problem alert reports. The Bechtel designers have, where practical, followed the design guidelines of NRC Regulatory Guide 8.8. The designers have also followed recommendations of the NSSS supplier (Westinghouse). Many of these recommendations are available in documented form, in the Westinghouse information packages and in Reference1. Other CALLAWAY - SP12.1-2Rev. OL-155/06recommendations developed from discussions with Westinghouse. Westinghouse representatives met with Bechtel designers, as appropriate. Design of NSSS equipment, within Westinghouse's scope of supply, to ALARA objectives is described in Reference1. In the early stages of the plant design, calculations were performed to quantify potential concentrations of radioactivity throughout process systems and buildings of the power block. These calculations made use of measured data from operating plants, and also followed the methodology of NRC Regulatory Guide 1.112. Systems designs and equipment specifications have been influenced by these assessments (see Section12.1.2.5). Additionally, radiation dose rates have been estimated throughout the power block. Specifications for equipment and/or systems to be purchased reflect the objective to keep ORE-ALARA. For example, the specifications for the solid radwaste system state that "special consideration shall be given to eliminate points where radioactive materials may tend to accumulate."  The specifications also require provisions for remote flushing and rinsing of portions of equipment that contain radioactive material and state that "special design consideration shall be given to minimizing operator exposure to radiation during the maintenance of equipment."12.1.2.2Scale-ModelProgramThe use of scale models is an innovative aspect of the Callaway project that has proven effective for design and design review. Three classes of scale models have been used with respect to radiation protection:  (1) preliminary design model, (2) study models, and (3) final design model. The preliminary design model (3/8 inch = 1 foot) was, in effect, a three-dimensional layout tool. Model builders constructed the main elements of the buildings (walls, floors, columns, etc.) and set in place major equipment (tanks, pumps, motors, etc.). Plant design engineers then routed piping, electrical cable trays, and HVAC ducts, and located valves and valve actuators on the model. The three-dimensional aspects of the model, compared to conventional layout drawings, placed radiation protection, among other considerations, into sharp focus for both design engineers and reviewers and facilitated evaluation of design alternatives, such as choosing the best valve placement. Prior to completion of the preliminary design models, several reviews by utility and SNUPPS staff were held, as discussed further in Section 12.1.2.4. Upon completion of the preliminary design model, the approved layouts were committed to paper as design drawings. Study models were less formal in concept than the preliminary design and final design models and were constructed to evaluate specific design features or alternatives. For example, a study model was constructed of portions of the auxiliary building in order to evaluate the arrangement of radioactive demineralizers. The results of the review of this model are discussed in Section 12.1.2.5
.
CALLAWAY - SP12.1-3Rev. OL-155/06The final design model (3/4 inch = 1 foot) was built by model makers from design drawings, but was constructed sufficiently early to permit design changes, e.g., to facilitate maintenance. The final design model comprises the reactor building, auxiliary building, control building, radwaste building, and turbine building. Additional design reviews by Bechtel and by SNUPPS were performed, using the final design model (see Sections 12.1.2.3 and 12.1.2.4). These reviews again focused on such factors as operability, maintainability, and radiation protection. 12.1.2.3Second-LevelDesignReviewsby the Lead Architect-EngineerThe term second-level review pertains to reviews beyond that of first-level supervision. On-project reviews are conducted by engineering and supervisory personnel from groups, other than the group that originated the design, and by higher level supervisory personnel. The reviews are generally interdisciplinary. For example, layout of shielding and valve operators in the radwaste building is performed by plant design personnel and is reviewed by mechanical/nuclear engineering personnel, as well as by the project engineers for plant design and systems. These reviews bring a broader base of experience to bear on all aspects of design, including radiation protection. Participants are professional engineers with from 5 to 20 years of nuclear power plant design experience. Off-project reviews are performed, when requested by the project organization, by members of the Bechtel technical staff. There have been no off-project reviews specifically addressed to radiation protection. However, shielding calculations have been the responsibility of the chief nuclear engineer and his staff and, through this involvement in the design, the staff personnel have contributed to the ALARA review. 12.1.2.4DesignReviewsbySNUPPSUtility reviews of all safety-related systems, structures, and equipment have been coordinated through the SNUPPS Technical Committee, which is composed of senior-level utility engineers, one from each SNUPPS utility. It is the responsibility of the Technical Committee to obtain comments from appropriate personnel within their companies and to bring those comments to meetings of the Committee, where decisions are made on the basis of discussion and eventually a vote. Throughout the design and construction phase of the SNUPPS project, the Technical Committee has met or had a telephone conference call on the average of once every week and a half, and total meeting days per year of the Technical Committee have averaged about 50. The SNUPPS technical director participates in Technical Committee meetings. Assisting the Technical Committee are various plant review groups, which are ad hoc groups of utility personnel selected to review and recommend action on specific aspects of design. The total meeting days per year have averaged about 15. A SNUPPS staff member participates in each meeting.
CALLAWAY - SP12.1-3Rev. OL-155/06The final design model (3/4 inch = 1 foot) was built by model makers from design drawings, but was constructed sufficiently early to permit design changes, e.g., to facilitate maintenance. The final design model comprises the reactor building, auxiliary building, control building, radwaste building, and turbine building. Additional design reviews by Bechtel and by SNUPPS were performed, using the final design model (see Sections 12.1.2.3 and 12.1.2.4). These reviews again focused on such factors as operability, maintainability, and radiation protection. 12.1.2.3Second-LevelDesignReviewsby the Lead Architect-EngineerThe term second-level review pertains to reviews beyond that of first-level supervision. On-project reviews are conducted by engineering and supervisory personnel from groups, other than the group that originated the design, and by higher level supervisory personnel. The reviews are generally interdisciplinary. For example, layout of shielding and valve operators in the radwaste building is performed by plant design personnel and is reviewed by mechanical/nuclear engineering personnel, as well as by the project engineers for plant design and systems. These reviews bring a broader base of experience to bear on all aspects of design, including radiation protection. Participants are professional engineers with from 5 to 20 years of nuclear power plant design experience. Off-project reviews are performed, when requested by the project organization, by members of the Bechtel technical staff. There have been no off-project reviews specifically addressed to radiation protection. However, shielding calculations have been the responsibility of the chief nuclear engineer and his staff and, through this involvement in the design, the staff personnel have contributed to the ALARA review. 12.1.2.4DesignReviewsbySNUPPSUtility reviews of all safety-related systems, structures, and equipment have been coordinated through the SNUPPS Technical Committee, which is composed of senior-level utility engineers, one from each SNUPPS utility. It is the responsibility of the Technical Committee to obtain comments from appropriate personnel within their companies and to bring those comments to meetings of the Committee, where decisions are made on the basis of discussion and eventually a vote. Throughout the design and construction phase of the SNUPPS project, the Technical Committee has met or had a telephone conference call on the average of once every week and a half, and total meeting days per year of the Technical Committee have averaged about 50. The SNUPPS technical director participates in Technical Committee meetings. Assisting the Technical Committee are various plant review groups, which are ad hoc groups of utility personnel selected to review and recommend action on specific aspects of design. The total meeting days per year have averaged about 15. A SNUPPS staff member participates in each meeting.
CALLAWAY - SP12.1-4Rev. OL-155/06Radiation protection has been an important aspect of the reviews by the Technical Committee and the plant review groups and has been the subject of numerous specific reviews, as discussed further in Section 12.1.2.5. At least six distinct reviews of the scale models have specifically included consideration of radiation protection during plant operation and maintenance. At these reviews, representatives of SNUPPS have included, in addition to Technical Committee members and SNUPPS staff, the health physics superintendant from Ginna Station, several licensed senior reactor operators, operations and maintenance supervisors from other nuclear power plants, and utility engineering personnel experienced in nuclear plant operation. Several design changes resulted from these reviews. Examples are described in Section12.1.2.5
CALLAWAY - SP12.1-4Rev. OL-155/06Radiation protection has been an important aspect of the reviews by the Technical Committee and the plant review groups and has been the subject of numerous specific reviews, as discussed further in Section 12.1.2.5. At least six distinct reviews of the scale models have specifically included consideration of radiation protection during plant operation and maintenance. At these reviews, representatives of SNUPPS have included, in addition to Technical Committee members and SNUPPS staff, the health physics superintendant from Ginna Station, several licensed senior reactor operators, operations and maintenance supervisors from other nuclear power plants, and utility engineering personnel experienced in nuclear plant operation. Several design changes resulted from these reviews. Examples are described in Section12.1.2.5. During design and construction of the first SNUPPS unit, SNUPPS staff and qualified personnel from other SNUPPS utilities participated in a review of the effectiveness of design to the ALARA objectives. Periodically during construction, preoperational testing, and start-up of the SNUPPS units, qualified personnel from each SNUPPS utility will participate in ALARA reviews of their respective plants. 12.1.2.5ExamplesofRadiationProtectionDesignReviewsa.Radiation Zone DrawingsEvery location within the power block has been assigned a radiation zone classification. The method of establishing radiation zone classifications has been as follows. Shortly after initiation of the SNUPPS project, Bechtel prepared shielding design criteria which were reviewed by a SNUPPS plant review group. Participants in that review were utility engineering personnel, including one person with an SRO license from Ginna Station.
. During design and construction of the first SNUPPS unit, SNUPPS staff and qualified personnel from other SNUPPS utilities participated in a review of the effectiveness of design to the ALARA objectives. Periodically during construction, preoperational testing, and start-up of the SNUPPS units, qualified personnel from each SNUPPS utility will participate in ALARA reviews of their respective plants. 12.1.2.5ExamplesofRadiationProtectionDesignReviewsa.Radiation Zone DrawingsEvery location within the power block has been assigned a radiation zone classification. The method of establishing radiation zone classifications has been as follows. Shortly after initiation of the SNUPPS project, Bechtel prepared shielding design criteria which were reviewed by a SNUPPS plant review group. Participants in that review were utility engineering personnel, including one person with an SRO license from Ginna Station.
Subsequently, Bechtel prepared radiation zone drawings which define the zone classification of each location in the power block. These drawings were reviewed by the utilities and SNUPPS staff. Participants in those reviews included the Technical Committee, utility engineering personnel, utility operating and health physics personnel, and SNUPPS staff (technical director and civil/structural engineer). b.Reactor CavitySNUPPS learned in the Spring of 1975 that, unless neutron shielding was provided for the reactor cavity, neutron dose rates in the containment would be 10 to 100 times too high to permit operator access to the containment for reasonable periods of time during full-power operation. This conclusion was based on measured dose rates in the Calvert Cliffs plant, which is designed (as is SNUPPS) for access to the outside of the reactor vessel for performance of inservice inspection. SNUPPS had Bechtel undertake a study of possible neutron shield configurations, the effect of the shield on subcompartment pressure and pressure loadings on the reactor vessel, and obtainable dose reduction factors. SNUPPS has had numerous CALLAWAY - SP12.1-5Rev. OL-155/06design review meetings with Bechtel. Participants for SNUPPS have included:  SNUPPS staff (executive director, technical director, and licensing manager), the Technical Committee, and utility engineering personnel. SNUPPS also contracted for an outside review by the NUS Corporation, which included independent estimates of neutron streaming and the effectiveness of neutron shielding materials. The design of the neutron shield in the containment is described in Sections 3.8.3.1.4 and 12.3.2.2.1. Subsequent to NRC approval of Callaway's revised GDC-4 submittal, a permanent cavity seal/neutron shield was installed to reduce neutron dose rates in containment.c.Radioactive DemineralizersAs a result of a review of a study model of the auxiliary building, the following design changes were made:1.A single wall between the corridor and the valve and piping compartments was designed to replace space-consuming overlapping (staggered for radiation protection) access walls. This allows more room in the corridor and more accessibility for maintenance within the compartments. 2.Vertical valve controls were designed to replace horizontal controls. This eliminates the need for 90-degree turns in the valve control fixtures and eliminates the access difficulty, which horizontal valve control rods pose as obstacles to maintenance. 3.A concrete shielding floor was provided above the valve and piping compartments to minimize exposure to the valve control operators. d.Airborne Radioactivity in ContainmentAirborne radioactivity (predominantly noble gases) has limited containment access in operating PWRs. Leakage of noble gases from the reactor coolant through the packing of the pressurizer spray valve has been determined to be a significant contributor to the gaseous activity in containment. To alleviate this situation, the following design provisions have been incorporated in the SNUPPS plants:1.A packless, low-leakage, ball-type pressurizer spray valve. 2.Provisions for continuous stripping of noble gases from the reactor coolant (see Section 9.3.4). 3.Addition of a mini-purge system to permit purging of the containment during power operation, prior to operator access (see Section 9.4.6).
Subsequently, Bechtel prepared radiation zone drawings which define the zone classification of each location in the power block. These drawings were reviewed by the utilities and SNUPPS staff. Participants in those reviews included the Technical Committee, utility engineering personnel, utility operating and health physics personnel, and SNUPPS staff (technical director and civil/structural engineer). b.Reactor CavitySNUPPS learned in the Spring of 1975 that, unless neutron shielding was provided for the reactor cavity, neutron dose rates in the containment would be 10 to 100 times too high to permit operator access to the containment for reasonable periods of time during full-power operation. This conclusion was based on measured dose rates in the Calvert Cliffs plant, which is designed (as is SNUPPS) for access to the outside of the reactor vessel for performance of inservice inspection. SNUPPS had Bechtel undertake a study of possible neutron shield configurations, the effect of the shield on subcompartment pressure and pressure loadings on the reactor vessel, and obtainable dose reduction factors. SNUPPS has had numerous CALLAWAY - SP12.1-5Rev. OL-155/06design review meetings with Bechtel. Participants for SNUPPS have included:  SNUPPS staff (executive director, technical director, and licensing manager), the Technical Committee, and utility engineering personnel. SNUPPS also contracted for an outside review by the NUS Corporation, which included independent estimates of neutron streaming and the effectiveness of neutron shielding materials. The design of the neutron shield in the containment is described in Sections 3.8.3.1.4 and 12.3.2.2.1. Subsequent to NRC approval of Callaway's revised GDC-4 submittal, a permanent cavity seal/neutron shield was installed to reduce neutron dose rates in containment.c.Radioactive DemineralizersAs a result of a review of a study model of the auxiliary building, the following design changes were made:1.A single wall between the corridor and the valve and piping compartments was designed to replace space-consuming overlapping (staggered for radiation protection) access walls. This allows more room in the corridor and more accessibility for maintenance within the compartments. 2.Vertical valve controls were designed to replace horizontal controls. This eliminates the need for 90-degree turns in the valve control fixtures and eliminates the access difficulty, which horizontal valve control rods pose as obstacles to maintenance. 3.A concrete shielding floor was provided above the valve and piping compartments to minimize exposure to the valve control operators. d.Airborne Radioactivity in ContainmentAirborne radioactivity (predominantly noble gases) has limited containment access in operating PWRs. Leakage of noble gases from the reactor coolant through the packing of the pressurizer spray valve has been determined to be a significant contributor to the gaseous activity in containment. To alleviate this situation, the following design provisions have been incorporated in the SNUPPS plants:1.A packless, low-leakage, ball-type pressurizer spray valve. 2.Provisions for continuous stripping of noble gases from the reactor coolant (see Section 9.3.4). 3.Addition of a mini-purge system to permit purging of the containment during power operation, prior to operator access (see Section 9.4.6).
CALLAWAY - SP12.1-6Rev. OL-155/06e.Steam Generator MaintenanceA final example of the results of scale-model reviews by SNUPPS is provision of permanent maintenance platforms below the steam generators to facilitate access and thereby reduce associated personnel radiation doses for maintenance operations, such as eddy-current inspection of tubes and sludge-lancing. 12.1.2.6DecommissioningThe following features of the plant design will assist decommissioning crews to maintain ORE-ALARA during the eventual decommissioning of Callaway. a.The building arrangements, compartmentation, corridors, doorways, and hatches provide the ability to remove most items of equipment intact or, alternatively, to isolate and entomb specific areas. b.The design features to maintain ORE-ALARA throughout the plant operating life are also applicable to the eventual decommissioning of the plant. These features include equipment design for ease of accessibility and maintenance, floor and wall coatings, provisions for remote flushing of equipment, ability to use remote handling equipment, decontamination techniques, and component design features to minimize crud buildup. c.Specifications and limitations on cobalt content in equipment components will serve to limit radiation doses from crud buildup during both operation and subsequent decommissioning. d.The amount of potentially radioactive buried pipe is limited. 12.1.3OPERATIONAL CONSIDERATIONSRefer to Section 12.1.1.3 of the Site Addendum for operational considerations. 12.
CALLAWAY - SP12.1-6Rev. OL-155/06e.Steam Generator MaintenanceA final example of the results of scale-model reviews by SNUPPS is provision of permanent maintenance platforms below the steam generators to facilitate access and thereby reduce associated personnel radiation doses for maintenance operations, such as eddy-current inspection of tubes and sludge-lancing. 12.1.2.6DecommissioningThe following features of the plant design will assist decommissioning crews to maintain ORE-ALARA during the eventual decommissioning of Callaway. a.The building arrangements, compartmentation, corridors, doorways, and hatches provide the ability to remove most items of equipment intact or, alternatively, to isolate and entomb specific areas. b.The design features to maintain ORE-ALARA throughout the plant operating life are also applicable to the eventual decommissioning of the plant. These features include equipment design for ease of accessibility and maintenance, floor and wall coatings, provisions for remote flushing of equipment, ability to use remote handling equipment, decontamination techniques, and component design features to minimize crud buildup. c.Specifications and limitations on cobalt content in equipment components will serve to limit radiation doses from crud buildup during both operation and subsequent decommissioning. d.The amount of potentially radioactive buried pipe is limited. 12.1.3OPERATIONAL CONSIDERATIONSRefer to Section 12.1.1.3 of the Site Addendum for operational considerations. 12.
Line 42: Line 38:
==1.4REFERENCES==
==1.4REFERENCES==
1."Design, Inspection, Operation, and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures As Low As Reasonably Achievable," WCAP - 8872, April 1977 CALLAWAY - SP12.2-1Rev. OL-2211/1612.2RADIATIONSOURCESThe sources of radiation that form the basis for shield design calculations and the sources of airborne radioactivity required for the design of personnel protective measures and for dose assessment are discussed and identified in this section.12.2.1CONTAINED SOURCESThe shielding design is based on full-power operation with 0.25 percent fuel cladding defects (Ref. 1, 2, 3, 4). The sources were obtained by multiplying the ANSI N237 fission product sources by two (Ref. 5). Sources in the primary coolant include fission products released from fuel clad defects and activation and corrosion products. The sources in the primary coolant are discussed in Section 11.1 and listed in Table 11.1-4. Throughout most of the primary coolant system, activation products, principally nitrogen-16 during reactor operation, are the primary radiation sources for shielding design. For all systems transporting radioactive materials, conservative allowance is made for transit decay, while, at the same time, providing for daughter product formation.In this section, the design sources are presented by building location and system. General location of the equipment discussed in this section is shown in the general arrangement drawings provided in Section1.2
1."Design, Inspection, Operation, and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures As Low As Reasonably Achievable," WCAP - 8872, April 1977 CALLAWAY - SP12.2-1Rev. OL-2211/1612.2RADIATIONSOURCESThe sources of radiation that form the basis for shield design calculations and the sources of airborne radioactivity required for the design of personnel protective measures and for dose assessment are discussed and identified in this section.12.2.1CONTAINED SOURCESThe shielding design is based on full-power operation with 0.25 percent fuel cladding defects (Ref. 1, 2, 3, 4). The sources were obtained by multiplying the ANSI N237 fission product sources by two (Ref. 5). Sources in the primary coolant include fission products released from fuel clad defects and activation and corrosion products. The sources in the primary coolant are discussed in Section 11.1 and listed in Table 11.1-4. Throughout most of the primary coolant system, activation products, principally nitrogen-16 during reactor operation, are the primary radiation sources for shielding design. For all systems transporting radioactive materials, conservative allowance is made for transit decay, while, at the same time, providing for daughter product formation.In this section, the design sources are presented by building location and system. General location of the equipment discussed in this section is shown in the general arrangement drawings provided in Section1.2
.12.2.1.1Containment12.2.1.1.1Reactor Core The primary radiations within the containment during normal operation are neutrons and gamma rays emanating from the reactor core. Tables 12.2-1 and 12.2-2 list neutron and gamma multigroup fluxes at the core centerline location outside the reactor vessel. The tables are based on nuclear parameter values discussed in Chapter 4.0
.12.2.1.1Containment12.2.1.1.1Reactor Core The primary radiations within the containment during normal operation are neutrons and gamma rays emanating from the reactor core. Tables 12.2-1 and 12.2-2 list neutron and gamma multigroup fluxes at the core centerline location outside the reactor vessel. The tables are based on nuclear parameter values discussed in Chapter 4.0. Table 12.2-4 lists core gamma fluxes at the core centerline location outside the reactor vessel after shutdown, for shielding requirements during shutdown and inservice inspection. 12.2.1.1.2Reactor Coolant SystemSources of radiation in the reactor coolant system are fission products released from fuel and activation and corrosion products which are circulated in the reactor coolant. These sources are listed in Table 11.1-4, and their bases are discussed in Section11.1
. Table 12.2-4 lists core gamma fluxes at the core centerline location outside the reactor vessel after shutdown, for shielding requirements during shutdown and inservice inspection. 12.2.1.1.2Reactor Coolant SystemSources of radiation in the reactor coolant system are fission products released from fuel and activation and corrosion products which are circulated in the reactor coolant. These sources are listed in Table 11.1-4, and their bases are discussed in Section11.1
.During operation, the activation product nitrogen-16 is the predominant activity in the reactor coolant pumps, steam generators, and reactor coolant piping. The contained source of radiation within the pressurizer is comprised of a liquid volume activity, a vapor volume activity, and a deposited activity. These activities are identified in Table12.2-3
.During operation, the activation product nitrogen-16 is the predominant activity in the reactor coolant pumps, steam generators, and reactor coolant piping. The contained source of radiation within the pressurizer is comprised of a liquid volume activity, a vapor volume activity, and a deposited activity. These activities are identified in Table12.2-3
.
.
CALLAWAY - SP12.2-2Rev. OL-2211/1612.2.1.1.3Secondary Coolant CycleUnder normal operating conditions, there is insignificant radioactive contamination present within the steam and power conversion system. It is possible to spread contamination to this cycle via steam generator tube leakage. Based on the primary-to-secondary leak rate given in Table 11.1A-1, the equilibrium secondary system activities are developed in Section ll.l and provided in Table 11.1-4. The condensate demineralizers and steam generator blowdown system further reduce the radioactivity level in the secondary cycle, as described in Section11.1
CALLAWAY - SP12.2-2Rev. OL-2211/1612.2.1.1.3Secondary Coolant CycleUnder normal operating conditions, there is insignificant radioactive contamination present within the steam and power conversion system. It is possible to spread contamination to this cycle via steam generator tube leakage. Based on the primary-to-secondary leak rate given in Table 11.1A-1, the equilibrium secondary system activities are developed in Section ll.l and provided in Table 11.1-4. The condensate demineralizers and steam generator blowdown system further reduce the radioactivity level in the secondary cycle, as described in Section11.1. An evaluation of the secondary coolant activity in Tables 11.1-6 (Sheet 4) and 11.1-4 verifies that shielding is not required for the steam and power conversion system, with the exception of the components that could potentially concentrate the radioactivity. The condensate demineralizers and the steam generator blowdown demineralizers are the only components which could potentially concentrate the radioactivity. These are discussed in Sections 12.2.1.4 and 12.2.1.5, respectively. 12.2.1.1.4Auxiliary Systems Residual heat removal system-see Section 12.2.1.2.1
. An evaluation of the secondary coolant activity in Tables 11.1-6 (Sheet 4) and 11.1-4 verifies that shielding is not required for the steam and power conversion system, with the exception of the components that could potentially concentrate the radioactivity. The condensate demineralizers and the steam generator blowdown demineralizers are the only components which could potentially concentrate the radioactivity. These are discussed in Sections 12.2.1.4 and 12.2.1.5, respectively. 12.2.1.1.4Auxiliary Systems Residual heat removal system-see Section 12.2.1.2.1
.Chemical and volume control system-see Section12.2.1.2.2. Steam generator blowdown system-see Section 12.2.1.5. 12.2.1.1.5Spent Fuel Handling and TransferThe spent fuel assemblies are the predominant source of radiation in the containment after plant shutdown for refueling. A reactor operating time necessary to establish fission product buildup near equilibrium for the reactor at rated power is used in determining the source strength. Shielding requirements for spent fuel transfer are based on the fission product activity present 72 hours after shutdown to conservatively take credit for the time elapsed prior to the initiation of refueling operations. Source terms for spent fuel are listed in Table12.2-4. 12.2.1.2AuxiliaryBuilding12.2.1.2.1Residual Heat Removal SystemThe pumps, heat exchangers, and associated piping of the residual heat removal (RHR) system contain radioactive materials. For plant shutdown, the RHR pumps and heat exchanger sources result from the radioactive isotopes carried in the reactor coolant, discussed in Section 12.2.1.1.2, considering 4 hours of decay following shutdown. The radiation source terms for the RHR system are listed in Table12.2-5.
.Chemical and volume control system-see Section12.2.1.2.2
CALLAWAY - SP12.2-3Rev. OL-2211/1612.2.1.2.2Chemical and Volume Control SystemThe CVCS source activity is the reactor coolant inventory which is provided in Table 11.1-4. More than 1minute of N-16 coolant activity decay is provided before the letdown line exits the containment, and, therefore, is not significant in determining shielding requirements for the CVCS equipment outside the containment.Major equipment items include the letdown and excess letdown heat exchangers, mixed bed and cation bed demineralizers, reactor coolant filter, volume control tank, and charging pumps. The seal water subsystem for the reactor coolant pumps includes the injection and return filters and the seal water heat exchanger. The design activities of the CVCS components are listed in Table 12.2-6. Heat exchanger and piping activities are derived from primary coolant activities. Radiation sources in the various pumps are assumed to be identical to the liquid sources in the tank from which the pump takes suction. 12.2.1.2.3Nuclear Sampling SystemThe major radiation sources in the nuclear sampling system originate from the RCS, RHR, and CVCS systems. The greatest radiation exposure would be to personnel taking the samples. To minimize this exposure, an integral shield has been incorporated into the sampling station design. Plant procedures provide contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere during the recovery phase following plant transient or accident.The sampling systems, which provide the capability to make required analyses are listed on Tables9.3-3 , 9.3-4 , 9.3-5, and 9.3-6. Figures 9.3-2 , 9.3-3, and 9.3-4 show the piping and instrumentation diagrams for the sampling systems. Process radioactivity monitors for the sampling systems are indicated in the tables and figures mentioned above. The area and airborne radioactivity monitors for worker protection are given in Section12.3.4 and are shown in Figure12.3-2
. Steam generator blowdown system-see Section 12.2.1.5
. 12.2.1.1.5Spent Fuel Handling and TransferThe spent fuel assemblies are the predominant source of radiation in the containment after plant shutdown for refueling. A reactor operating time necessary to establish fission product buildup near equilibrium for the reactor at rated power is used in determining the source strength. Shielding requirements for spent fuel transfer are based on the fission product activity present 72 hours after shutdown to conservatively take credit for the time elapsed prior to the initiation of refueling operations. Source terms for spent fuel are listed in Table12.2-4
. 12.2.1.2AuxiliaryBuilding12.2.1.2.1Residual Heat Removal SystemThe pumps, heat exchangers, and associated piping of the residual heat removal (RHR) system contain radioactive materials. For plant shutdown, the RHR pumps and heat exchanger sources result from the radioactive isotopes carried in the reactor coolant, discussed in Section 12.2.1.1.2, considering 4 hours of decay following shutdown. The radiation source terms for the RHR system are listed in Table12.2-5
.
CALLAWAY - SP12.2-3Rev. OL-2211/1612.2.1.2.2Chemical and Volume Control SystemThe CVCS source activity is the reactor coolant inventory which is provided in Table 11.1-4. More than 1minute of N-16 coolant activity decay is provided before the letdown line exits the containment, and, therefore, is not significant in determining shielding requirements for the CVCS equipment outside the containment.Major equipment items include the letdown and excess letdown heat exchangers, mixed bed and cation bed demineralizers, reactor coolant filter, volume control tank, and charging pumps. The seal water subsystem for the reactor coolant pumps includes the injection and return filters and the seal water heat exchanger. The design activities of the CVCS components are listed in Table 12.2-6. Heat exchanger and piping activities are derived from primary coolant activities. Radiation sources in the various pumps are assumed to be identical to the liquid sources in the tank from which the pump takes suction. 12.2.1.2.3Nuclear Sampling SystemThe major radiation sources in the nuclear sampling system originate from the RCS, RHR, and CVCS systems. The greatest radiation exposure would be to personnel taking the samples. To minimize this exposure, an integral shield has been incorporated into the sampling station design. Plant procedures provide contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere during the recovery phase following plant transient or accident.The sampling systems, which provide the capability to make required analyses are listed on Tables9.3-3
, 9.3-4, 9.3-5, and 9.3-6. Figures 9.3-2
, 9.3-3, and 9.3-4 show the piping and instrumentation diagrams for the sampling systems. Process radioactivity monitors for the sampling systems are indicated in the tables and figures mentioned above. The area and airborne radioactivity monitors for worker protection are given in Section12.3.4 and are shown in Figure12.3-2
.12.2.1.3FuelBuilding12.2.1.3.1Spent Fuel Storage and Transfer The predominant radioactivity sources in the spent fuel storage and transfer areas in the fuel building are the spent fuel assemblies. Spent fuel assembly sources are discussed in Section12.2.1.1.5. For shielding design, the fuel storage pool assumptions are given in Section9.1.2. The major radionuclide concentrations in the water are provided in Table12.2-7
.12.2.1.3FuelBuilding12.2.1.3.1Spent Fuel Storage and Transfer The predominant radioactivity sources in the spent fuel storage and transfer areas in the fuel building are the spent fuel assemblies. Spent fuel assembly sources are discussed in Section12.2.1.1.5. For shielding design, the fuel storage pool assumptions are given in Section9.1.2. The major radionuclide concentrations in the water are provided in Table12.2-7
.The fuel transfer tube is completely shielded with permanent shielding to within radiation zone limits. No special access control, radiation monitoring, or posting is required. The CALLAWAY - SP12.2-4Rev. OL-2211/16expansion bellows for the fuel transfer tube are under water in the fuel transfer canal in the fuel building (see Figure3.8-48). There is no bellows inspection room or opening. The fuel transfer tube is completely surrounded by concrete or water, with the exception of the seismic gaps, so that no personnel access is possible. The fuel transfer tube in the seismic gap between the containment wall and the internal containment structure and in the seismic gap between the containment wall and the fuel building is shielded, using permanently installed lead loaded silicone foam rubber, to meet the radiation zone limits (see Figure12.3-2). Therefore, there is no unshielded portion of the fuel transfer tube.Shielding during dry fuel storage, transfer, and transport operations is provided by the HI-STORM UMAX Multi-Purpose Canister (MPC), the HI-TRAC VW transfer cask, the Vertical Ventilated Module (VVM), and the VVM closure lid. In addition to this shielding, the HI-STORM UMAX Certificate of Compliance (CoC) requires a minimum decay time of 3 years for fuel assemblies selected for dry fuel storage. 12.2.1.3.2Fuel Pool Cooling and Cleanup SystemSources in the fuel pool cooling and cleanup system (FPCCS) are the result of the transfer of radioactive isotopes from the reactor coolant into the fuel storage pool during refueling operations. The reactor coolant activities for fission, corrosion, and activation products (Table 11.1-4) are decayed for the amount of time required to remove the reactor vessel head following shutdown, are reduced by operation of the letdown system filters and demineralizers following shutdown, and are diluted by the total volumes of the water in the reactor vessel, refueling pool, and fuel storage pool. This activity then undergoes subsequent decay and accumulation on the FPCC filters and demineralizers.12.2.1.4TurbineBuilding12.2.1.4.1Main Steam Supply and Power Conversion SystemsPotential radioactivity in the main steam supply and power conversion systems is a result of steam generator tube leaks and fuel cladding defects, as discussed in Section12.2.1.1.3
.The fuel transfer tube is completely shielded with permanent shielding to within radiation zone limits. No special access control, radiation monitoring, or posting is required. The CALLAWAY - SP12.2-4Rev. OL-2211/16expansion bellows for the fuel transfer tube are under water in the fuel transfer canal in the fuel building (see Figure3.8-48). There is no bellows inspection room or opening. The fuel transfer tube is completely surrounded by concrete or water, with the exception of the seismic gaps, so that no personnel access is possible. The fuel transfer tube in the seismic gap between the containment wall and the internal containment structure and in the seismic gap between the containment wall and the fuel building is shielded, using permanently installed lead loaded silicone foam rubber, to meet the radiation zone limits (see Figure12.3-2). Therefore, there is no unshielded portion of the fuel transfer tube.Shielding during dry fuel storage, transfer, and transport operations is provided by the HI-STORM UMAX Multi-Purpose Canister (MPC), the HI-TRAC VW transfer cask, the Vertical Ventilated Module (VVM), and the VVM closure lid. In addition to this shielding, the HI-STORM UMAX Certificate of Compliance (CoC) requires a minimum decay time of 3 years for fuel assemblies selected for dry fuel storage. 12.2.1.3.2Fuel Pool Cooling and Cleanup SystemSources in the fuel pool cooling and cleanup system (FPCCS) are the result of the transfer of radioactive isotopes from the reactor coolant into the fuel storage pool during refueling operations. The reactor coolant activities for fission, corrosion, and activation products (Table 11.1-4) are decayed for the amount of time required to remove the reactor vessel head following shutdown, are reduced by operation of the letdown system filters and demineralizers following shutdown, and are diluted by the total volumes of the water in the reactor vessel, refueling pool, and fuel storage pool. This activity then undergoes subsequent decay and accumulation on the FPCC filters and demineralizers.12.2.1.4TurbineBuilding12.2.1.4.1Main Steam Supply and Power Conversion SystemsPotential radioactivity in the main steam supply and power conversion systems is a result of steam generator tube leaks and fuel cladding defects, as discussed in Section12.2.1.1.3. This radioactivity is sufficiently low that no radiation shielding for equipment in the turbine building is required in order to meet the radiation zone requirements. The isotopic concentrations for a condensate demineralizer bed and other secondary system sources are listed in Table12.2-8.
. This radioactivity is sufficiently low that no radiation shielding for equipment in the turbine building is required in order to meet the radiation zone requirements. The isotopic concentrations for a condensate demineralizer bed and other secondary system sources are listed in Table12.2-8
CALLAWAY - SP12.2-5Rev. OL-2211/1612.2.1.5RadwasteBuilding12.2.1.5.1Boron Recycle and Liquid Radwaste SystemsThe system sources are radioisotopes, including fission and activation products, present in the reactor coolant. The components of the systems contain varying degrees of activity, depending on the detailed system and equipment design. The concentrations of radionuclides in the process fluids at various locations in the radwaste systems, such as pipes, tanks, filters and demineralizers are discussed in Section 11.1 and are listed in Tables 11.1-4 and 11.1-6. These nuclide concentrations for 0.25 percent failed fuel have been used in the final shielding design. Shielding for each component of the radwaste systems is based on maximum activity conditions, as given in Sections11.1 and 11.2. 12.2.1.5.2Gaseous Radwaste SystemRadiation sources for each component of the waste gas system are based on operation under the conditions given in Sections11.1 and 11.3. Tabulation of the activities is shown in Table12.2-9. 12.2.1.5.3Solid Radwaste SystemRadiation sources for each component providing influent to the solid radwaste system are based on operation under the conditions given in Sections 11.1 and 11.4. Tabulation of the activities is shown in Table11.4-3. 12.2.1.6SourcesResultingfromDesignBasisAccidentsThe radiation sources from design basis accidents include the design basis inventory of radioactive isotopes in the reactor coolant, plus postulated fission product releases from the fuel. Accident parameters and sources are discussed and evaluated in Chapters11.0 and 15.0. 12.2.1.7StoredRadioactivityThe principal sources of activity not stored inside the plant structures are the original steam generators, the original reactor vessel closure head, spent fuel assemblies stored within the Independent Spent Fuel Storage Installation (ISFSI), the Forced Helium Dehydration (FHD) skid stored within the ISFSI Support Building, the reactor makeup water storage tank (RMWST), the refueling water storage tank (RWST), the condensate storage tank (CST), and two radwaste discharge monitor tanks. The original steam generators and original reactor vessel closure head are stored in the Old Steam Generator Storage Facility (OSGSF). As shown in Figure 12.2-1, the OSGF contains the following 3 rooms or areas: the Old Steam Generator (OSG) bay, the Old CALLAWAY - SP12.2-6Rev. OL-2211/16Reactor Vessel Head (ORVH) room, and miscellaneous radioactive materials (RAM) storage room. Contaminated materials in the RAM storage room are limited to those not exceeding the threshold of Category 2 radioactive materials as defined in 10 CFR 37.The walls and ceiling of the OSGSF are constructed of reinforced concrete with a minimum thickness of 2 feet. Shielding calculations that conservatively assume 100% of the activity in the OSGSF is from high energy Co-60 have been performed to confirm the external dose rate will not exceed 2 mrem/hr for the OSGSF. The OSGSF is expected to yield an external surface dose rate of 0.6 mrem/hr or less. Personnel access to the OSGSF is controlled by two labyrinth entryways with locked steel doors. One entryway leads into the OSG bay, and one entryway leads into the RAM storage room. A locked access door is also provided between the OSG bay and ORVH room. Additionally, a lockable steel rollup door is also provided for exterior access to the RAM storage room. The OSG bay and ORVH room are treated as confined spaces due to emergency egress requirements associated with Building Officials and Code Administrators (BOCA) National Building Code, 1999. Personnel access to the facility is infrequent. The annual dose to an individual at the site boundary due to normal operation of the ISFSI is limited to the values specified in 10 CFR 72.104. The RMWST and CST are expected to contain concentrations of radionuclides which yield a surface dose rate of 0.5 mrem/hr or less. The RWST is expected to have a maximum contact dose of less than 10mrem/hr when the water is returned from the refueling pool. This will be rapidly reduced by processing through the fuel storage pool purification filter and demineralizer. The radwaste discharge monitor tanks will also have a dose rate of less than 10mrem/hr at contact. Tabulations of the activities within these tanks are provided in Table11.1-6 (Sheets1, 2, and 3). All spent fuel is stored in the fuel storage pool until it is placed in either a spent fuel storage canister for storage in the ISFSI, or a spent fuel shipping cask for transport offsite. Storage space is allocated in the radwaste building for the storage of spent filter cartridges and dewatered spent resins, and chemical/hazardous  wastes. Radioactive wastes stored inside the plant structures are shielded so that there is Zone A access outside the structure (see Figure 12.3-2). If it becomes necessary to temporarily store radioactive wastes outside the plant structures, adequate radiation protection measures will be taken by the radiation protection staff. 12.2.2AIRBORNE RADIOACTIVE MATERIAL SOURCESThis section identifies the models, parameters, and sources required to evaluate airborne radionuclide concentrations during plant operations in various plant radiation areas where personnel occupancies are expected. An analysis of operating plant measurements of average airborne radionuclide concentrations (Ref. 6) and their respective MPCS for various stations in the auxiliary building (these stations include the waste handling area and the waste gas decay tank rooms, which are located in the radwaste building) indicates that the concentration to CALLAWAY - SP12.2-7Rev. OL-2211/16MPC ratios in all stations is well within the limits of Note5 of AppendixB to 10 CFR 20.1-20.601 to consider these airborne activities as insignificant. It is expected that the SNUPPS units would not have airborne radioactivity concentrations significantly greater than the operating plant data for corresponding locations. However, it is possible that, within these stations, there may be rooms where maximum airborne concentrations can occur due to localized leakage of radioactive fluids, but these rooms house equipment and components that handle highly radioactive fluids and consequently are D or E zones and, therefore, will be normally inaccessible. Also the ventilation systems in the auxiliary and radwaste buildings are designed in such a manner that airborne contamination from high radiation zones will not generally spread into low radiation zones, since the airflow is from regions of lower potential for contamination to those with higher potential for contamination. Consequently, negligible airborne radioactivity concentrations are expected in those areas of the auxiliary and radwaste buildings which are accessible. Airborne radioactivity concentrations in the turbine building are also expected to be negligible, since possible leaks into the turbine building are only from the secondary side, and, also, the turbine building ventilation exhaust is high (at least 90,000 cfm). For example, airborne concentrations are calculated to be 2.3 x 10
.
CALLAWAY - SP12.2-5Rev. OL-2211/1612.2.1.5RadwasteBuilding12.2.1.5.1Boron Recycle and Liquid Radwaste SystemsThe system sources are radioisotopes, including fission and activation products, present in the reactor coolant. The components of the systems contain varying degrees of activity, depending on the detailed system and equipment design. The concentrations of radionuclides in the process fluids at various locations in the radwaste systems, such as pipes, tanks, filters and demineralizers are discussed in Section 11.1 and are listed in Tables 11.1-4 and 11.1-6. These nuclide concentrations for 0.25 percent failed fuel have been used in the final shielding design. Shielding for each component of the radwaste systems is based on maximum activity conditions, as given in Sections11.1 and 11.2. 12.2.1.5.2Gaseous Radwaste SystemRadiation sources for each component of the waste gas system are based on operation under the conditions given in Sections11.1 and 11.3. Tabulation of the activities is shown in Table12.2-9
. 12.2.1.5.3Solid Radwaste SystemRadiation sources for each component providing influent to the solid radwaste system are based on operation under the conditions given in Sections 11.1 and 11.4. Tabulation of the activities is shown in Table11.4-3
. 12.2.1.6SourcesResultingfromDesignBasisAccidentsThe radiation sources from design basis accidents include the design basis inventory of radioactive isotopes in the reactor coolant, plus postulated fission product releases from the fuel. Accident parameters and sources are discussed and evaluated in Chapters11.0 and 15.0. 12.2.1.7StoredRadioactivityThe principal sources of activity not stored inside the plant structures are the original steam generators, the original reactor vessel closure head, spent fuel assemblies stored within the Independent Spent Fuel Storage Installation (ISFSI), the Forced Helium Dehydration (FHD) skid stored within the ISFSI Support Building, the reactor makeup water storage tank (RMWST), the refueling water storage tank (RWST), the condensate storage tank (CST), and two radwaste discharge monitor tanks. The original steam generators and original reactor vessel closure head are stored in the Old Steam Generator Storage Facility (OSGSF). As shown in Figure 12.2-1, the OSGF contains the following 3 rooms or areas: the Old Steam Generator (OSG) bay, the Old CALLAWAY - SP12.2-6Rev. OL-2211/16Reactor Vessel Head (ORVH) room, and miscellaneous radioactive materials (RAM) storage room. Contaminated materials in the RAM storage room are limited to those not exceeding the threshold of Category 2 radioactive materials as defined in 10 CFR 37.The walls and ceiling of the OSGSF are constructed of reinforced concrete with a minimum thickness of 2 feet. Shielding calculations that conservatively assume 100% of the activity in the OSGSF is from high energy Co-60 have been performed to confirm the external dose rate will not exceed 2 mrem/hr for the OSGSF. The OSGSF is expected to yield an external surface dose rate of 0.6 mrem/hr or less. Personnel access to the OSGSF is controlled by two labyrinth entryways with locked steel doors. One entryway leads into the OSG bay, and one entryway leads into the RAM storage room. A locked access door is also provided between the OSG bay and ORVH room. Additionally, a lockable steel rollup door is also provided for exterior access to the RAM storage room. The OSG bay and ORVH room are treated as confined spaces due to emergency egress requirements associated with Building Officials and Code Administrators (BOCA) National Building Code, 1999. Personnel access to the facility is infrequent. The annual dose to an individual at the site boundary due to normal operation of the ISFSI is limited to the values specified in 10 CFR 72.104. The RMWST and CST are expected to contain concentrations of radionuclides which yield a surface dose rate of 0.5 mrem/hr or less. The RWST is expected to have a maximum contact dose of less than 10mrem/hr when the water is returned from the refueling pool. This will be rapidly reduced by processing through the fuel storage pool purification filter and demineralizer. The radwaste discharge monitor tanks will also have a dose rate of less than 10mrem/hr at contact. Tabulations of the activities within these tanks are provided in Table11.1-6 (Sheets1, 2, and 3). All spent fuel is stored in the fuel storage pool until it is placed in either a spent fuel storage canister for storage in the ISFSI, or a spent fuel shipping cask for transport offsite. Storage space is allocated in the radwaste building for the storage of spent filter cartridges and dewatered spent resins, and chemical/hazardous  wastes. Radioactive wastes stored inside the plant structures are shielded so that there is Zone A access outside the structure (see Figure 12.3-2). If it becomes necessary to temporarily store radioactive wastes outside the plant structures, adequate radiation protection measures will be taken by the radiation protection staff. 12.2.2AIRBORNE RADIOACTIVE MATERIAL SOURCESThis section identifies the models, parameters, and sources required to evaluate airborne radionuclide concentrations during plant operations in various plant radiation areas where personnel occupancies are expected. An analysis of operating plant measurements of average airborne radionuclide concentrations (Ref. 6) and their respective MPCS for various stations in the auxiliary building (these stations include the waste handling area and the waste gas decay tank rooms, which are located in the radwaste building) indicates that the concentration to CALLAWAY - SP12.2-7Rev. OL-2211/16MPC ratios in all stations is well within the limits of Note5 of AppendixB to 10 CFR 20.1-20.601 to consider these airborne activities as insignificant. It is expected that the SNUPPS units would not have airborne radioactivity concentrations significantly greater than the operating plant data for corresponding locations. However, it is possible that, within these stations, there may be rooms where maximum airborne concentrations can occur due to localized leakage of radioactive fluids, but these rooms house equipment and components that handle highly radioactive fluids and consequently are D or E zones and, therefore, will be normally inaccessible. Also the ventilation systems in the auxiliary and radwaste buildings are designed in such a manner that airborne contamination from high radiation zones will not generally spread into low radiation zones, since the airflow is from regions of lower potential for contamination to those with higher potential for contamination. Consequently, negligible airborne radioactivity concentrations are expected in those areas of the auxiliary and radwaste buildings which are accessible. Airborne radioactivity concentrations in the turbine building are also expected to be negligible, since possible leaks into the turbine building are only from the secondary side, and, also, the turbine building ventilation exhaust is high (at least 90,000 cfm). For example, airborne concentrations are calculated to be 2.3 x 10
-3 µCi/cc and 6.9x10
-3 µCi/cc and 6.9x10
-12 µCi/cc for I-131 and Xe-133, respectively in the turbine building. Higher airborne concentrations can, however, occur in the containment, both during power operation and refueling--the former due to coolant leakage and the latter primarily due to the evaporation of the refueling pool. Likewise, airborne concentrations can also occur in the fuel building both during power operation and refueling due to the evaporation of the spent fuel pool. During power operation, the airborne radioactivity in the fuel building will be almost all due to tritium, since the continuous operation of the fuel pool cleanup system will effectively remove other isotopes from the pool. The assumptions and parameters required to evaluate the airborne radionuclide concentrations in the containment and fuel building both during power operation and refueling are listed in Table12.2-11. The concentrations in these buildings are listed in Table12.2-12. Even though some of these airborne concentrations may be high, limited occupancies in these areas ensures that the doses from airborne radioactivity to an individual will be a small fraction of the 10 CFR 20 limits for occupation exposures. Airborne radioactivity is monitored inside the plant, as described in Section12.3.4, and in process equipment and effluents, as described in Section11.5
-12 µCi/cc for I-131 and Xe-133, respectively in the turbine building. Higher airborne concentrations can, however, occur in the containment, both during power operation and refueling--the former due to coolant leakage and the latter primarily due to the evaporation of the refueling pool. Likewise, airborne concentrations can also occur in the fuel building both during power operation and refueling due to the evaporation of the spent fuel pool. During power operation, the airborne radioactivity in the fuel building will be almost all due to tritium, since the continuous operation of the fuel pool cleanup system will effectively remove other isotopes from the pool. The assumptions and parameters required to evaluate the airborne radionuclide concentrations in the containment and fuel building both during power operation and refueling are listed in Table12.2-11. The concentrations in these buildings are listed in Table12.2-12. Even though some of these airborne concentrations may be high, limited occupancies in these areas ensures that the doses from airborne radioactivity to an individual will be a small fraction of the 10 CFR 20 limits for occupation exposures. Airborne radioactivity is monitored inside the plant, as described in Section12.3.4, and in process equipment and effluents, as described in Section11.5. 12.2.2.1ModelForCalculatingAirborneConcentrationsFor those regions which are characterized by a constant leakrate of the radioactive source at constant source strength and a constant exhaust rate of the contaminant, the CALLAWAY - SP12.2-8Rev. OL-2211/16peak or equilibrium airborne concentration of the radioisotope in the regions can be calculated, using the following equation:whereandFrom the above equation, it is evident that the peak or equilibrium concentration, CEqi , of the i th radioisotope in the applicable region will be given by the following expression:With high exhaust rates, this peak concentration will be reached within a few hours.  (1) (LR)i=Leak or evaporation rate of the  i th radioisotope in gm/sec, in the applicable region A i=activity concentration of the i th leaking or evaporating radioisotope in  
. 12.2.2.1ModelForCalculatingAirborneConcentrationsFor those regions which are characterized by a constant leakrate of the radioactive source at constant source strength and a constant exhaust rate of the contaminant, the CALLAWAY - SP12.2-8Rev. OL-2211/16peak or equilibrium airborne concentration of the radioisotope in the regions can be calculated, using the following equation:whereandFrom the above equation, it is evident that the peak or equilibrium concentration, CEqi , of the ith radioisotope in the applicable region will be given by the following expression:With high exhaust rates, this peak concentration will be reached within a few hours.  (1) (LR)i=Leak or evaporation rate of the  i th radioisotope in gm/sec, in the applicable region Ai=activity concentration of the i th leaking or evaporating radioisotope in  
µCI/gm(PF)i=partition factor or the fraction of the leaking activity that is airborne for the i th radioisotope Ti=Total removal rate constant for the i th radioisotope in sec
µCI/gm(PF)i=partition factor or the fraction of the leaking activity that is airborne for the i th radioisotope Ti=Total removal rate constant for the i th radioisotope in sec
-1 from the applicable region
-1 from the applicable region
=(di + e)di + e are the removal rate constants in sec
=(di + e)di + e are the removal rate constants in sec
-1 due to radioactive decay for the i th radioisotope and the exhaust from the applicable regiont=time interval between the start of the leak and the time at which the concentration is evaluated in secondsV=free volume of the region in which the leak occurs in cc Ci(t)=airborne concentration of the i th radioisotope at time t in mCi/cm3 in the applicable region CEqi = (LR)i Ai (PF)i /VTi(2)Cit()LR()iAiPF()i1eTit--()VTi--------------------------------------------------------------
-1 due to radioactive decay for the i th radioisotope and the exhaust from the applicable regiont=time interval between the start of the leak and the time at which the concentration is evaluated in secondsV=free volume of the region in which the leak occurs in cc C i(t)=airborne concentration of the i th radioisotope at time t in mCi/cm3 in the applicable region CEqi = (LR)i A i (PF)i /VTi(2)Cit ()LR ()iAiPF ()i 1eTi t--()VTi--------------------------------------------------------------
=
=
CALLAWAY - SP12.2-9Rev. OL-2211/1612.
CALLAWAY - SP12.2-9Rev. OL-2211/1612.


==2.3REFERENCES==
==2.3REFERENCES==
1.C. M. Lederer, et. al., TableofIsotopes, Lawrence Radiation Laboratory, University of California (March, 1968). 2.ReactorPhysicsConstants, Argonne National Laboratory, ANL-5800 (July, 1963).3.H. Soodak, ReactorHandbook, Vol. III, Part A, Physics, second edition (1962). 4.D. A. Klopp, NAP-Multigroup Time-dependent Neutron Activation Prediction Code, IITRI-A6088-21 (January, 1966), conditions as given in Sections 11.1
1.C. M. Lederer, et. al., TableofIsotopes, Lawrence Radiation Laboratory, University of California (March, 1968). 2.ReactorPhysicsConstants, Argonne National Laboratory, ANL-5800 (July, 1963).3.H. Soodak, ReactorHandbook, Vol. III, Part A, Physics, second edition (1962). 4.D. A. Klopp, NAP-Multigroup Time-dependent Neutron Activation Prediction Code, IITRI-A6088-21 (January, 1966), conditions as given in Sections 11.1 , 11.2 , and 11.4. 5.ANSI N 237, "Source Term Specification," Final Draft, 1977. 6.NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station - Unit 1, Prepared for USNRC by EG & Idaho, Inc., July 1978.7.NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous & Liquid Effluents from Pressurized Water Reactors, USNRC, April 1976.
, 11.2, and 11.4. 5.ANSI N 237, "Source Term Specification," Final Draft, 1977. 6.NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station - Unit 1, Prepared for USNRC by EG & Idaho, Inc., July 1978.7.NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous & Liquid Effluents from Pressurized Water Reactors, USNRC, April 1976.
CALLAWAY - SPRev. OL-135/03TABLE 12.2-1  NEUTRON FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)Energy GroupNeutron Flux(neutrons/cm 2-sec)1(E > 1.0 Mev)7.6 x 10 81(5.53 Kev < E  1.0 Mev)1.2 x 10 103(0.625 ev  E  5.53 Kev)7.1 x 10 94(E < 0.625 ev)1.8 x 10 9 CALLAWAY - SPRev. OL-135/03TABLE 12.2-2  GAMMA FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)GroupFlux (Mev/cm 2-sec)GroupEnergy(Mev/)1  3.7 x 10 9 7.52  3.3 x 10 9 4.03  1.7 x 10 9 2.54  1.0 x 10 9 0.8 CALLAWAY - SPRev. OL-135/03TABLE 12.2-3  PRESSURIZER SHIELDING SOURCE TERMSLiquidVolumeIsotopeActivity
CALLAWAY - SPRev. OL-135/03TABLE 12.2-1  NEUTRON FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)Energy GroupNeutron Flux(neutrons/cm 2-sec)1(E > 1.0 Mev)7.6 x 1081(5.53 Kev < E  1.0 Mev)1.2 x 10103(0.625 ev  E  5.53 Kev)7.1 x 1094(E < 0.625 ev)1.8 x 109 CALLAWAY - SPRev. OL-135/03TABLE 12.2-2  GAMMA FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)GroupFlux (Mev/cm 2-sec)GroupEnergy(Mev/)1  3.7 x 10 97.52  3.3 x 10 94.03  1.7 x 10 92.54  1.0 x 10 90.8 CALLAWAY - SPRev. OL-135/03TABLE 12.2-3  PRESSURIZER SHIELDING SOURCE TERMSLiquidVolumeIsotopeActivity
(&#xb5;Ci/gm)N-161.8 (max)Cr-5l1.68E-03Mn-543.10E-04Fe-551.61E-03Fe-599.30E-04Co-581.53E-02Co-602.02E-03Br-832.53E-04Br-843.08E-05Br-853.39E-07Rb-861.48E-04Rb-881.32E-03Sr-896.84E-04Sr-902.11E-05Sr-911.29E-04Y-89M6.16E-08Y-901.35E-05 Y-91M8.44E-05Y-911.33E-04Y-937.05E-06Zr-955.72E-05Nb-954.84E-05Nb-95M2.92E-05Mu-997.34E-02Tc-993.68E-09Ru-1034.14E-05Ru-1061.00E-05Te-125M2.75E-05Te-127M2.73E-04Te-1273.26E-04Te-129M1.27E-03Te-1298.23E-04Te-131M6.13E-04Te-1311.15E-04Te-1321.24E-02I-1298.55E-13I-1305.17E-04I-1313.83E-01I-1321.71E-02I-1331.46E-01I-1349.09E-04I-1352.68E-02Cs-1345.24E-02Cs-1351.04E-10Cs-1362.10E-02Cs-1373.79E-02Ba-137M3.58E-02Ba-1403.55E-04La-1403.43E-04Ce-1411.32E-04Ce-1436.86E-05Ce-1446.86E-05Pr-1438.67E-05 Pr-1446.86E-05SteamVolumeIsotopeActivity
(&#xb5;Ci/gm)N-161.8 (max)Cr-5l1.68E-03Mn-543.10E-04Fe-551.61E-03Fe-599.30E-04Co-581.53E-02Co-602.02E-03Br-832.53E-04Br-843.08E-05Br-853.39E-07Rb-861.48E-04Rb-881.32E-03Sr-896.84E-04Sr-902.11E-05Sr-911.29E-04Y-89M6.16E-08Y-901.35E-05 Y-91M8.44E-05Y-911.33E-04Y-937.05E-06Zr-955.72E-05Nb-954.84E-05Nb-95M2.92E-05Mu-997.34E-02Tc-993.68E-09Ru-1034.14E-05Ru-1061.00E-05Te-125M2.75E-05Te-127M2.73E-04Te-1273.26E-04Te-129M1.27E-03Te-1298.23E-04Te-131M6.13E-04Te-1311.15E-04Te-1321.24E-02I-1298.55E-13I-1305.17E-04I-1313.83E-01I-1321.71E-02I-1331.46E-01I-1349.09E-04I-1352.68E-02Cs-1345.24E-02Cs-1351.04E-10Cs-1362.10E-02Cs-1373.79E-02Ba-137M3.58E-02Ba-1403.55E-04La-1403.43E-04Ce-1411.32E-04Ce-1436.86E-05Ce-1446.86E-05Pr-1438.67E-05 Pr-1446.86E-05SteamVolumeIsotopeActivity
(&#xb5;Ci/gm)Kr-83M9.74E-04Kr-85M1.14E-02Kr-851.13E+00Kr-871.92E-03Kr-881.36E-02Kr-896.89E-06Xe-131M1.31E-01Xe-133M1.35E-01Xe-1331.59E+01Xe-135M9.03E-05Xe-1356.79E-02Xe-1371.49E-05Xe-1382.69E-04LiquidVolumeIsotopeActivity
(&#xb5;Ci/gm)Kr-83M9.74E-04Kr-85M1.14E-02Kr-851.13E+00Kr-871.92E-03Kr-881.36E-02Kr-896.89E-06Xe-131M1.31E-01Xe-133M1.35E-01Xe-1331.59E+01Xe-135M9.03E-05Xe-1356.79E-02Xe-1371.49E-05Xe-1382.69E-04LiquidVolumeIsotopeActivity
(&#xb5;Ci/gm)
(&#xb5;Ci/gm)
CALLAWAY - SPRev. OL-135/03TABLE 12.2-4  SPENT FUEL SHUTDOWN SOURCES (FULL CORE)Photon EnergyTime After Shutdown(Mev)4 Hours12 Hours1 Day1 Week1 Month3 Months0.43.1 x 10 112.3 x 10111.9 x 10119.2 x 10103.8 x 10101.3 x 10100.81.3 x 10 129.8 x 10118.0 x 10114.0 x 10112.3 x 10111.2 x 10111.33.9 x 10 112.9 x 10112.5 x 10111.6 x 10111.2 x 10115.8 x 10101.75.1 x 10 113.8 x 10113.3 x 10112.3 x 10116.2 x 10102.9 x 1092.27.2 x 10 102.6 x 10101.5 x 10108.5 x 1096.7 x 1095.0 x 1092.58.9 x 10 104.7 x 10103.7 x 10102.5 x 10107.9 x 1093.5 x 1083.58.2 x 10 92.0 x 1091.3 x 1099.6 x 1082.0 x 1081.5 x 107Values given in Mev/cc-sec.
CALLAWAY - SPRev. OL-135/03TABLE 12.2-4  SPENT FUEL SHUTDOWN SOURCES (FULL CORE)Photon EnergyTime After Shutdown(Mev)4 Hours12 Hours1 Day1 Week1 Month3 Months0.43.1 x 10 112.3 x 10 111.9 x 10 119.2 x 10 103.8 x 10 101.3 x 10 100.81.3 x 10 129.8 x 10 118.0 x 10 114.0 x 10 112.3 x 10 111.2 x 10 111.33.9 x 10 112.9 x 10 112.5 x 10 111.6 x 10 111.2 x 10 115.8 x 10 101.75.1 x 10 113.8 x 10 113.3 x 10 112.3 x 10 116.2 x 10 102.9 x 10 92.27.2 x 10 102.6 x 10 101.5 x 10 108.5 x 10 96.7 x 10 95.0 x 10 92.58.9 x 10 104.7 x 10 103.7 x 10 102.5 x 10 107.9 x 10 93.5 x 10 83.58.2 x 10 92.0 x 10 91.3 x 10 99.6 x 10 82.0 x 10 81.5 x 10 7Values given in Mev/cc-sec.
CALLAWAY - SPRev. OL-135/03TABLE 12.2-5  RADIATION SOURCES RESIDUAL HEAT REMOVAL SYSTEMIsotopeActivities(&#xb5;Ci/gm)Cr-511.51E-03Mn-542.48E-04Fe-551.28E-03 Fe-597.98E-04Co-581.28E-02Co-601.60E-03 Br-832.49E-03Br-842.28E-05Br-85negligible Kr-83M1.34E-02Kr-85M1.20E-01Kr-851.68E-02 Kr-871.48E-02Kr-881.57E-01Kr-89negligible Rb-861.39E-04 Rb-881.75E-01Rb-895.28E-08Sr-895.73E-04 Sr-901.64E-05Sr-918.01E-04Y-89M5.15E-08 Y-902.58E-06Y-91M5.30E-04Y-911.06E-04 Y-934.25E-05Zr-954.79E-05Nb-954.00E-05 Nb-95M1.45E-06Mo-991.32E-01Tc-992.00E-10 Ru-1033.59E-05Ru-1068.00E-06Te-125M2.32E-05Te-127M2.24E-04Te-1275.63E-04Te-129M1.12E-03 Te-1297.66E-04Te-131M1.82E-03Te-1313.34E-04 Te-1322.08E-02I-1292.63E-14I-1302.75E-03 I-1314.37E-01I-1326.36E-02I-1335.46E-01 I-1343.17E-03I-1352.06E-01Xe-131M3.94E-02 Xe-133M2.07E-01 Xe-1331.06E+01Xe-135M3.56E-02Xe-1355.42E-01 Xe-137negligibleXe-1388.11E-07Cs-1344.10E-02 Cs-1354.48E-13Cs-1362.11E-02Cs-1372.95E-02 Cs-1384.47E-04Ba-137M2.79E-02Ba-1403.58E-04 La-1402.54E-04Ce-1411.14E-04Ce-1436.03E-05 Ce-1445.41E-05Pr-1438.18E-05Pr-1445.41E-05IsotopeActivities(&#xb5;Ci/gm)
CALLAWAY - SPRev. OL-135/03TABLE 12.2-5  RADIATION SOURCES RESIDUAL HEAT REMOVAL SYSTEMIsotopeActivities(&#xb5;Ci/gm)Cr-511.51E-03Mn-542.48E-04Fe-551.28E-03 Fe-597.98E-04Co-581.28E-02Co-601.60E-03 Br-832.49E-03Br-842.28E-05Br-85negligible Kr-83M1.34E-02Kr-85M1.20E-01Kr-851.68E-02 Kr-871.48E-02Kr-881.57E-01Kr-89negligible Rb-861.39E-04 Rb-881.75E-01Rb-895.28E-08Sr-895.73E-04 Sr-901.64E-05Sr-918.01E-04Y-89M5.15E-08 Y-902.58E-06Y-91M5.30E-04Y-911.06E-04 Y-934.25E-05Zr-954.79E-05Nb-954.00E-05 Nb-95M1.45E-06Mo-991.32E-01Tc-992.00E-10 Ru-1033.59E-05Ru-1068.00E-06Te-125M2.32E-05Te-127M2.24E-04Te-1275.63E-04Te-129M1.12E-03 Te-1297.66E-04Te-131M1.82E-03Te-1313.34E-04 Te-1322.08E-02I-1292.63E-14I-1302.75E-03 I-1314.37E-01I-1326.36E-02I-1335.46E-01 I-1343.17E-03I-1352.06E-01Xe-131M3.94E-02 Xe-133M2.07E-01 Xe-1331.06E+01Xe-135M3.56E-02Xe-1355.42E-01 Xe-137negligibleXe-1388.11E-07Cs-1344.10E-02 Cs-1354.48E-13Cs-1362.11E-02Cs-1372.95E-02 Cs-1384.47E-04Ba-137M2.79E-02Ba-1403.58E-04 La-1402.54E-04Ce-1411.14E-04Ce-1436.03E-05 Ce-1445.41E-05Pr-1438.18E-05Pr-1445.41E-05IsotopeActivities(&#xb5;Ci/gm)
CALLAWAY - SPTABLE 12.2-5 (Sheet 2)Rev. OL-135/03I-1314.37E-01I-1326.36E-02 I-1335.46E-01I-1343.17E-03I-1352.06E-01 XE-131M3.94E-02XE-133M2.07E-01XE-1331.06E+01 XE-135M3.56E-02XE-1355.42E-01XE-1372.78E-21 XE-1388.11E-07CS-1344.10E-02CS-1354.48E-13 CS-1362.11E-02CS-1372.95E-02CS-1384.47E-04 BA-137M2.79E-02BA-1403.58E-04LA-1402.54E-04 CE-1411.14E-04CE-1436.03E-05CE-1445.41E-05 PR-1438.18E-05PR-1445.41E-05IsotopeActivities(&#xb5;Ci/gm)
CALLAWAY - SPTABLE 12.2-5 (Sheet 2)Rev. OL-135/03I-1314.37E-01I-1326.36E-02 I-1335.46E-01I-1343.17E-03I-1352.06E-01 XE-131M3.94E-02XE-133M2.07E-01XE-1331.06E+01 XE-135M3.56E-02XE-1355.42E-01XE-1372.78E-21 XE-1388.11E-07CS-1344.10E-02CS-1354.48E-13 CS-1362.11E-02CS-1372.95E-02CS-1384.47E-04 BA-137M2.79E-02BA-1403.58E-04LA-1402.54E-04 CE-1411.14E-04CE-1436.03E-05CE-1445.41E-05 PR-1438.18E-05PR-1445.41E-05IsotopeActivities(&#xb5;Ci/gm)
CALLAWAY - SPRev. OL-135/03TABLE 12.2-6  CHEMICAL AND VOLUME CONTROL SYSTEM SOURCESLETDOWN MIXED BED DEMINERALIZERIsotopeActivity(&#xb5;Ci/cc)Cr-513.30E+01Mn-543.36E+01Fe-552.23E+02Fe-592.80E+01Co-586.94E+02Co-602.96E+02Br-836.26E-01Br-847.47E-02Br-858.17E-04Rb-861.15E+00Rb-881.78E+00Sr-892.30E+01Sr-903.25E+00Sr-913.42E-01Y-89M2.07E-03Y-903.21E+00 Y-91M2.25E-01Y-915.13E+00Y-931.88E-02Zr-952.40E+00Nb-953.44E+00Nb-95M2.40E+00Mo-993.02E+02Tc-999.73E-04Ru-1031.12E+00Ru-1061.14E+00Te-125M1.04E+00Te-127M1.72E+01Te-1271.74E+01Te-129M2.99E+01Te-1291.92E+01Te-131M1.95E+00Te-1313.64E-01Te-1325.49E+01I-1291.12E-06 CALLAWAY - SPTABLE 12.2-6 (Sheet 2)Rev. OL-135/03Reactor Coolant FilterNOTE:All other demineralizers and filters throughout the plant are shielded with these same source terms, since these are the most radioactive.I-1301.41E+00I-1312.83E+03I-1326.73E+01I-1334.30E+02I-1342.21E+00I-1356.92E+01Cs-1343.88E+03Cs-1351.73E-05Cs-1361.22E+02Cs-1373.26E+03Ba-137M3.08E+03Ba-1403.66E+00La-1403.98E+00Ce-1412.96E+00Ce-1437.16E-02Ce-1447.20E+00Pr-1439.56E-01P4-1447.20E+00Specific SourceGamma Energy  Strength(Mev/) (Mev/cc-sec)  0.8 5.7 x 10 7  1.3 1.5 x 10 7IsotopeActivity(&#xb5;Ci/cc)
CALLAWAY - SPRev. OL-135/03TABLE 12.2-6  CHEMICAL AND VOLUME CONTROL SYSTEM SOURCESLETDOWN MIXED BED DEMINERALIZERIsotopeActivity (&#xb5;Ci/cc)Cr-513.30E+01Mn-543.36E+01Fe-552.23E+02Fe-592.80E+01Co-586.94E+02Co-602.96E+02Br-836.26E-01Br-847.47E-02Br-858.17E-04Rb-861.15E+00Rb-881.78E+00Sr-892.30E+01Sr-903.25E+00Sr-913.42E-01Y-89M2.07E-03Y-903.21E+00 Y-91M2.25E-01Y-915.13E+00Y-931.88E-02Zr-952.40E+00Nb-953.44E+00Nb-95M2.40E+00Mo-993.02E+02Tc-999.73E-04Ru-1031.12E+00Ru-1061.14E+00Te-125M1.04E+00Te-127M1.72E+01Te-1271.74E+01Te-129M2.99E+01Te-1291.92E+01Te-131M1.95E+00Te-1313.64E-01Te-1325.49E+01I-1291.12E-06 CALLAWAY - SPTABLE 12.2-6 (Sheet 2)Rev. OL-135/03Reactor Coolant FilterNOTE:All other demineralizers and filters throughout the plant are shielded with these same source terms, since these are the most radioactive.I-1301.41E+00I-1312.83E+03I-1326.73E+01I-1334.30E+02I-1342.21E+00I-1356.92E+01Cs-1343.88E+03Cs-1351.73E-05Cs-1361.22E+02Cs-1373.26E+03Ba-137M3.08E+03Ba-1403.66E+00La-1403.98E+00Ce-1412.96E+00Ce-1437.16E-02Ce-1447.20E+00Pr-1439.56E-01P4-1447.20E+00Specific SourceGamma Energy  Strength(Mev/) (Mev/cc-sec)  0.8 5.7 x 10 7  1.3 1.5 x 10 7IsotopeActivity (&#xb5;Ci/cc)
CALLAWAY - SPTABLE 12.2-6 (Sheet 3)Rev. OL-135/03RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities
CALLAWAY - SPTABLE 12.2-6 (Sheet 3)Rev. OL-135/03RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities
(&#xb5;Ci/gm)IsotopeActivities
(&#xb5;Ci/gm)IsotopeActivities
(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)Cr-511.90E-03Cr-511.73E-04Kr-83M2.69E-01Mn-543.10E-04Mn-542.82E-05Kr-85M2.01E+00Fe-551.60E-03Fe-551.46E-04Kr-852.40E-01 Fe-591.00E-03Fe-599.10E-05Kr-876.11E-01Co-581.60E-02Co-581.46E-03Kr-883.11E+00Co-602.00E-03Co-601.82E-04Kr-893.19E-03Br-831.00E-02Br-831.00E-03Xe-131M5.65E-01Br-845.42E-03Br-845.42E-04Xe-133M2.96E+00Br-856.25E-04Br-856.25E-05Xe-1331.52E+02Kr-83M4.54E-02Rb-868.08E-05Xe-135M3.88E-02Kr-85M2.25E-01Rb-881.90E-01Xe-1357.16E+00Kr-851.67E-02Sr-896.65E-05Xe-1376.88E-03Kr-871.32E-01Sr-901.90E-06Xe-1381.17E-01Kr-884.23E-01Sr-911.24E-04 Kr-891.13E-02Y-902.28E-07Rb-861.77E-04Y-91M6.84E-05Rb-884.17E-01Y-911.22E-05Sr-897.29E-04Y-936.46E-06Sr-902.08E-05Zr-955.46E-06Sr-911.35E-03Nb-954.55E-06Y-902.50E-06Mo-991.60E-02Y-91M7.50E-04Ru-1034.10E-06Y-911.33E-04Ru-1069.10E-07 CALLAWAY - SPTABLE 12.2-6 (Sheet 4)Rev. OL-135/03Y-937.08E-05Te-125M2.64E-06----------Zr-956.00E-05Te-127M2.55E-05Nb-955.00E-05Te-1277.74E-05Mo-991.75E-01Te-129M1.27E-04Ru-1034.50E-05Te-1291.46E-04 Ru-1061.00E-05Te-131M2.28E-04Te-125M2.90E-05Te-1311.00E-04 Te-127M2.80E-04Te-1322.46E-03Te-1278.50E-04I-1304.37E-04Te-129M1.40E-03I-1315.62E-02Te-1291.60E-03I-1322.08E-02Te-131M2.50E-03I-1337.92E-02Te-1311.10E-03I-1349.79E-03 Te-1322.70E-02I-1353.96E-02I-1304.37E-03Cs-1342.38E-02 I-1315.62E-01Cs-1361.24E-02 I-1322.08E-01Cs-1371.71E-02 I-1337.92E-01Ba-137M3.04E-03 I-1349.79E-02Ba-1404.18E-05 I-1353.96E-01La-1402.85E-05Xe-131M3.98E-02Ce-1411.33E-05 Xe-133M2.17E-01Ce-1437.60E-06Xe-1331.08E+01Ce-1446.27E-06Xe-135M3.00E-02Pr-1439.50E-06Xe-1356.44E-01Pr-1446.27E-06 Xe-1372.04E-02RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities
(&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)Cr-511.90E-03Cr-511.73E-04Kr-83M2.69E-01Mn-543.10E-04Mn-542.82E-05Kr-85M2.01E+00Fe-551.60E-03Fe-551.46E-04Kr-852.40E-01 Fe-591.00E-03Fe-599.10E-05Kr-876.11E-01Co-581.60E-02Co-581.46E-03Kr-883.11E+00Co-602.00E-03Co-601.82E-04Kr-893.19E-03Br-831.00E-02Br-831.00E-03Xe-131M5.65E-01Br-845.42E-03Br-845.42E-04Xe-133M2.96E+00Br-856.25E-04Br-856.25E-05Xe-1331.52E+02Kr-83M4.54E-02Rb-868.08E-05Xe-135M3.88E-02Kr-85M2.25E-01Rb-881.90E-01Xe-1357.16E+00Kr-851.67E-02Sr-896.65E-05Xe-1376.88E-03Kr-871.32E-01Sr-901.90E-06Xe-1381.17E-01Kr-884.23E-01Sr-911.24E-04 Kr-891.13E-02Y-902.28E-07Rb-861.77E-04Y-91M6.84E-05Rb-884.17E-01Y-911.22E-05Sr-897.29E-04Y-936.46E-06Sr-902.08E-05Zr-955.46E-06Sr-911.35E-03Nb-954.55E-06Y-902.50E-06Mo-991.60E-02Y-91M7.50E-04Ru-1034.10E-06Y-911.33E-04Ru-1069.10E-07 CALLAWAY - SPTABLE 12.2-6 (Sheet 4)Rev. OL-135/03Y-937.08E-05Te-125M2.64E-06----------Zr-956.00E-05Te-127M2.55E-05Nb-955.00E-05Te-1277.74E-05Mo-991.75E-01Te-129M1.27E-04Ru-1034.50E-05Te-1291.46E-04 Ru-1061.00E-05Te-131M2.28E-04Te-125M2.90E-05Te-1311.00E-04 Te-127M2.80E-04Te-1322.46E-03Te-1278.50E-04I-1304.37E-04Te-129M1.40E-03I-1315.62E-02Te-1291.60E-03I-1322.08E-02Te-131M2.50E-03I-1337.92E-02Te-1311.10E-03I-1349.79E-03 Te-1322.70E-02I-1353.96E-02I-1304.37E-03Cs-1342.38E-02 I-1315.62E-01Cs-1361.24E-02 I-1322.08E-01Cs-1371.71E-02 I-1337.92E-01Ba-137M3.04E-03 I-1349.79E-02Ba-1404.18E-05 I-1353.96E-01La-1402.85E-05Xe-131M3.98E-02Ce-1411.33E-05 Xe-133M2.17E-01Ce-1437.60E-06Xe-1331.08E+01Ce-1446.27E-06Xe-135M3.00E-02Pr-1439.50E-06Xe-1356.44E-01Pr-1446.27E-06 Xe-1372.04E-02RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities
(&#xb5;Ci/gm)IsotopeActivities
(&#xb5;Ci/gm)IsotopeActivities
(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)
(&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)
CALLAWAY - SPTABLE 12.2-6 (Sheet 5)Rev. OL-135/03Xe-1389.89E-02--------------------Cs-1345.21E-02Cs-1362.71E-02NOTE :
CALLAWAY - SPTABLE 12.2-6 (Sheet 5)Rev. OL-135/03Xe-1389.89E-02--------------------Cs-1345.21E-02Cs-1362.71E-02NOTE :
Cs-1373.75E-02Ba-137M3.33E-02The moderating heat exchanger, chiller heat exchanger, and letdown reheat heat exchanger are the same as the combined liquid and gaseous sources for the volume control tank.Ba-1404.58E-04La-1403.12E-04Ce-1411.46E-04Ce-1438.33E-05Ce-1446.87E-05Pr-1431.04E-04 Pr-1446.87E-05RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities
Cs-1373.75E-02Ba-137M3.33E-02The moderating heat exchanger, chiller heat exchanger, and letdown reheat heat exchanger are the same as the combined liquid and gaseous sources for the volume control tank.Ba-1404.58E-04La-1403.12E-04Ce-1411.46E-04Ce-1438.33E-05Ce-1446.87E-05Pr-1431.04E-04 Pr-1446.87E-05RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities
(&#xb5;Ci/gm)IsotopeActivities
(&#xb5;Ci/gm)IsotopeActivities
(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)
(&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)
CALLAWAY - SPTABLE 12.2-6 (Sheet 6)Rev. OL-135/03Boric Acid TanksIsotopeActivity(&#xb5;Ci/cc)Cr-514.16E-06Mn-548.56E-07Fe-554.49E-06Fe-592.41E-06Co-584.08E-05Co-605.63E-06Br-832.54E-08Br-841.92E-12Br-85negligibleRb-868.40E-06Rb-881.87E-09Rb-895.94E-14Sr-897.32E-06Sr-905.90E-08Sr-916.39E-08Y-89M6.58E-10Y-905.00E-08Y-91M4.27E-08Y-913.54E-07 Y-933.61E-09Zr-951.52E-07Nb-951.35E-07Nb-95M1.18E-07Mo-998.72E-05Tc-991.45E-11Ru-1031.06E-07Ru-1062.77E-08Te-125M7.23E-08Te-127M7.40E-07Te-1277.68E-07Te-129M3.21E-06Te-1292.05E-06Te-131M5.16E-07Te-1319.41E-08Te-1321.59E-05I-1295.08E-15I-1303.06E-07 CALLAWAY - SPTABLE 12.2-6 (Sheet 7)Rev. OL-135/03I-1317.46E-04I-1321.67E-05I-1331.09E-04I-1342.26E-09I-1351.06E-05Cs-1343.56E-03Cs-1353.71E-09Cs-1361.11E-03Cs-1372.59E-03Cs-1387.90E-08Ba-137M2.45E-03Ba-1407.64E-07La-1408.17E-07Ce-1413.31E-07Ce-1431.92E-08Ce-1441.90E-07Pr-1431.93E-07Pr-1441.90E-07Boric Acid TanksIsotopeActivity(&#xb5;Ci/cc)
CALLAWAY - SPTABLE 12.2-6 (Sheet 6)Rev. OL-135/03Boric Acid TanksIsotopeActivity (&#xb5;Ci/cc)Cr-514.16E-06Mn-548.56E-07Fe-554.49E-06Fe-592.41E-06Co-584.08E-05Co-605.63E-06Br-832.54E-08Br-841.92E-12Br-85negligibleRb-868.40E-06Rb-881.87E-09Rb-895.94E-14Sr-897.32E-06Sr-905.90E-08Sr-916.39E-08Y-89M6.58E-10Y-905.00E-08Y-91M4.27E-08Y-913.54E-07 Y-933.61E-09Zr-951.52E-07Nb-951.35E-07Nb-95M1.18E-07Mo-998.72E-05Tc-991.45E-11Ru-1031.06E-07Ru-1062.77E-08Te-125M7.23E-08Te-127M7.40E-07Te-1277.68E-07Te-129M3.21E-06Te-1292.05E-06Te-131M5.16E-07Te-1319.41E-08Te-1321.59E-05I-1295.08E-15I-1303.06E-07 CALLAWAY - SPTABLE 12.2-6 (Sheet 7)Rev. OL-135/03I-1317.46E-04I-1321.67E-05I-1331.09E-04I-1342.26E-09I-1351.06E-05Cs-1343.56E-03Cs-1353.71E-09Cs-1361.11E-03Cs-1372.59E-03Cs-1387.90E-08Ba-137M2.45E-03Ba-1407.64E-07La-1408.17E-07Ce-1413.31E-07Ce-1431.92E-08Ce-1441.90E-07Pr-1431.93E-07Pr-1441.90E-07Boric Acid TanksIsotopeActivity (&#xb5;Ci/cc)
CALLAWAY - SPRev. OL-135/03TABLE 12.2-7  FUEL STORAGE POOL WATER ACTIVITIESIsotope**Other isotopes will be present in much lower concentrations.Activities
CALLAWAY - SPRev. OL-135/03TABLE 12.2-7  FUEL STORAGE POOL WATER ACTIVITIESIsotope**Other isotopes will be present in much lower concentrations.Activities (&#xb5;Ci/cc)Co-58 1.3E-04 Co-60 4.0E-04 Cs-134 2.0E-05 Cs-137 2.0E-04 CALLAWAY - SPRev. OL-135/03TABLE 12.2-8  SECONDARY SYSTEM ACTIVITIESCondensate DemineralizerHigh TDS Regenerant Collector TankSecondary Liquid Waste Drain Collector TankSecondary Liquid Waste Monitor TankIsotopeActivity (&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)Cr-511.85E-06Cr-512.04E-07Cr-513.98E-11Cr-512.38E-12Mn-544.96E-07Mn-546.43E-08Mn-548.99E-12Mn-546.43E-13Fe-552.01E-06Fe-552.61E-07Fe-553.60E-11Fe-552.61E-12 Fe-591.33E-06Fe-591.73E-07Fe-592.67E-11Fe-591.72E-12Co-581.86E-05Co-582.41E-06Co-583.57E-10Co-582.41E-11Co-602.27E-06Co-602.94E-07Co-604.05E-11Co-602.94E-12 Br-834.96E-07Br-836.43E-08Br-838.83E-11Br-835.83E-13Br-842.94E-08Br-843.81E-09Br-845.23E-12Br-847.03E-18Br-854.27E-11Br-855.54E-12Br-857.62E-15Br-856.31E-66 Rb-861.52E-07Rb-861.96E-08Rb-863.95E-12Rb-861.94E-13Rb-887.73E-08Rb-881.00E-08Rb-881.53E-11Rb-883.08E-22Sr-899.40E-07Sr-891.22E-07Sr-891.86E-11Sr-891.21E-12 Sr-902.10E-08Sr-902.73E-09Sr-903.75E-13Sr-902.73E-14Sr-914.21E-08Sr-915.46E-09Sr-917.13E-12Sr-913.01E-14Y-89M8.46E-11Y-89M1.10E-11Y-89M1.67E-15Y-89M1.09E-16 Y-901.73E-08Y-902.25E-09Y-901.36E-13Y-902.29E-14Y-91M2.86E-08Y-91M3.70E-09Y-91M4.84E-12Y-91M2.01E-14Y-911.50E-07Y-911.95E-08Y-912.90E-12Y-911.94E-13 Y-932.19E-09Y-932.84E-10Y-933.68E-13Y-931.61E-15Zr-959.26E-08Zr-951.20E-08Zr-951.79E-12Zr-951.20E-13Nb-959.43E-08Nb-951.22E-08Nb-951.78E-12Nb-951.23E-13 Nb-95M6.62E-08Nb-95M8.59E-09Nb-95M2.61E-13Nb-95M8.80E-14Nb-994.89E-05Nb-996.35E-06Nb-993.14E-09Nb-995.82E-11Tc-995.99E-12Tc-997.77E-13Tc-992.59E-17Tc-997.96E-18 Ru-1034.37E-08Ru-1035.66E-09Ru-1038.89E-13Ru-1035.63E-14Ru-1069.93E-09Ru-1061.29E-09Ru-1061.79E-13Ru-1061.29E-14Te-125M2.29E-08Te-125M2.96E-09Te-125M4.47E-13Te-125M2.95E-14 Te-127M2.39E-07Te-127M3.10E-08Te-127M4.48E-12Te-127M3.10E-13Te-1272.61E-07Te-1273.39E-08Te-1278.06E-12Te-1273.25E-13 CALLAWAY - SPTABLE 12.2-8 (Sheet 2)Rev. OL-135/03Te-129M1.28E-06Te-129M1.66E-07Te-129M2.66E-11Te-129M1.65E-12Te-1298.26E-07Te-1291.07E-07Te-1291.83E-11Te-1291.06E-12Te-131M2.78E-07Te-131M3.61E-08Te-131M3.08E-11Te-131M2.98E-13Te-1315.17E-08Te-1316.70E-09Te-1315.77E-12Te-1315.44E-14 Te-1326.79E-06Te-1328.81E-07Te-1323.86E-10Te-1328.18E-12I-1291.44E-15I-1291.87E-16I-1302.34E-10I-1291.88E-20I-1301.45E-06I-1301.88E-07I-1317.85E-08I-1301.18E-11 I-1312.45E-03I-1313.18E-04I-1322.68E-09I-1313.09E-08I-1321.98E-05I-1322.57E-06I-1336.17E-08I-1322.82E-11I-1334.60E-04I-1335.97E-05I-1342.03E-10I-1334.52E-09 I-1341.14E-06I-1341.48E-07I-1351.19E-08I-1341.91E-14I-1356.77E-05I-1358.77E-06Cs-1341.20E-09I-1353.71E-10Cs-1346.00E-05Cs-1347.78E-06Cs-1357.31E-18Cs-1347.77E-11 Cs-1357.70E-13Cs-1359.99E-14Cs-1365.94E-10Cs-1356.54E-19Cs-1362.04E-05Cs-1362.65E-06Cs-1378.66E-10Cs-1362.60E-11Cs-1374.37E-05Cs-1375.67E-06Ba-137M8.21E-10Cs-1375.67E-11 Ba-137M4.14E-05Ba-137M5.36E-06Ba-1409.02E-12Ba-137M5.36E-11Ba-1403.42E-07Ba-1404.44E-08La-1407.56E-12Ba-1404.36E-13La-1403.63E-07La-1404.71E-08Ce-1413.70E-12La-1404.67E-13 Ce-1411.77E-07Ce-1412.29E-08Ce-1436.63E-13Ce-1412.27E-13Ce-1436.34E-09Ce-1438.23E-10Ce-1441.88E-12Ce-1436.91E-15Ce-1441.03E-07Ce-1441.34E-08Pr-1431.84E-12Ce-1441.34E-13 Pr-1437.36E-08Pr-1439.55E-09Pr-1441.90E-12Pr-1439.40E-14Pr-1441.03E-07Pr-1441.34E-08Pr-1441.34E-13Condensate DemineralizerHigh TDS Regenerant Collector TankSecondary Liquid Waste Drain Collector TankSecondary Liquid Waste Monitor TankIsotopeActivity (&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)IsotopeActivity (&#xb5;Ci/cc)
(&#xb5;Ci/cc)Co-58 1.3E-04 Co-60 4.0E-04 Cs-134 2.0E-05 Cs-137 2.0E-04 CALLAWAY - SPRev. OL-135/03TABLE 12.2-8  SECONDARY SYSTEM ACTIVITIESCondensate DemineralizerHigh TDS Regenerant Collector TankSecondary Liquid Waste Drain Collector TankSecondary Liquid Waste Monitor TankIsotopeActivity(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)Cr-511.85E-06Cr-512.04E-07Cr-513.98E-11Cr-512.38E-12Mn-544.96E-07Mn-546.43E-08Mn-548.99E-12Mn-546.43E-13Fe-552.01E-06Fe-552.61E-07Fe-553.60E-11Fe-552.61E-12 Fe-591.33E-06Fe-591.73E-07Fe-592.67E-11Fe-591.72E-12Co-581.86E-05Co-582.41E-06Co-583.57E-10Co-582.41E-11Co-602.27E-06Co-602.94E-07Co-604.05E-11Co-602.94E-12 Br-834.96E-07Br-836.43E-08Br-838.83E-11Br-835.83E-13Br-842.94E-08Br-843.81E-09Br-845.23E-12Br-847.03E-18Br-854.27E-11Br-855.54E-12Br-857.62E-15Br-856.31E-66 Rb-861.52E-07Rb-861.96E-08Rb-863.95E-12Rb-861.94E-13Rb-887.73E-08Rb-881.00E-08Rb-881.53E-11Rb-883.08E-22Sr-899.40E-07Sr-891.22E-07Sr-891.86E-11Sr-891.21E-12 Sr-902.10E-08Sr-902.73E-09Sr-903.75E-13Sr-902.73E-14Sr-914.21E-08Sr-915.46E-09Sr-917.13E-12Sr-913.01E-14Y-89M8.46E-11Y-89M1.10E-11Y-89M1.67E-15Y-89M1.09E-16 Y-901.73E-08Y-902.25E-09Y-901.36E-13Y-902.29E-14Y-91M2.86E-08Y-91M3.70E-09Y-91M4.84E-12Y-91M2.01E-14Y-911.50E-07Y-911.95E-08Y-912.90E-12Y-911.94E-13 Y-932.19E-09Y-932.84E-10Y-933.68E-13Y-931.61E-15Zr-959.26E-08Zr-951.20E-08Zr-951.79E-12Zr-951.20E-13Nb-959.43E-08Nb-951.22E-08Nb-951.78E-12Nb-951.23E-13 Nb-95M6.62E-08Nb-95M8.59E-09Nb-95M2.61E-13Nb-95M8.80E-14Nb-994.89E-05Nb-996.35E-06Nb-993.14E-09Nb-995.82E-11Tc-995.99E-12Tc-997.77E-13Tc-992.59E-17Tc-997.96E-18 Ru-1034.37E-08Ru-1035.66E-09Ru-1038.89E-13Ru-1035.63E-14Ru-1069.93E-09Ru-1061.29E-09Ru-1061.79E-13Ru-1061.29E-14Te-125M2.29E-08Te-125M2.96E-09Te-125M4.47E-13Te-125M2.95E-14 Te-127M2.39E-07Te-127M3.10E-08Te-127M4.48E-12Te-127M3.10E-13Te-1272.61E-07Te-1273.39E-08Te-1278.06E-12Te-1273.25E-13 CALLAWAY - SPTABLE 12.2-8 (Sheet 2)Rev. OL-135/03Te-129M1.28E-06Te-129M1.66E-07Te-129M2.66E-11Te-129M1.65E-12Te-1298.26E-07Te-1291.07E-07Te-1291.83E-11Te-1291.06E-12Te-131M2.78E-07Te-131M3.61E-08Te-131M3.08E-11Te-131M2.98E-13Te-1315.17E-08Te-1316.70E-09Te-1315.77E-12Te-1315.44E-14 Te-1326.79E-06Te-1328.81E-07Te-1323.86E-10Te-1328.18E-12I-1291.44E-15I-1291.87E-16I-1302.34E-10I-1291.88E-20I-1301.45E-06I-1301.88E-07I-1317.85E-08I-1301.18E-11 I-1312.45E-03I-1313.18E-04I-1322.68E-09I-1313.09E-08I-1321.98E-05I-1322.57E-06I-1336.17E-08I-1322.82E-11I-1334.60E-04I-1335.97E-05I-1342.03E-10I-1334.52E-09 I-1341.14E-06I-1341.48E-07I-1351.19E-08I-1341.91E-14I-1356.77E-05I-1358.77E-06Cs-1341.20E-09I-1353.71E-10Cs-1346.00E-05Cs-1347.78E-06Cs-1357.31E-18Cs-1347.77E-11 Cs-1357.70E-13Cs-1359.99E-14Cs-1365.94E-10Cs-1356.54E-19Cs-1362.04E-05Cs-1362.65E-06Cs-1378.66E-10Cs-1362.60E-11Cs-1374.37E-05Cs-1375.67E-06Ba-137M8.21E-10Cs-1375.67E-11 Ba-137M4.14E-05Ba-137M5.36E-06Ba-1409.02E-12Ba-137M5.36E-11Ba-1403.42E-07Ba-1404.44E-08La-1407.56E-12Ba-1404.36E-13La-1403.63E-07La-1404.71E-08Ce-1413.70E-12La-1404.67E-13 Ce-1411.77E-07Ce-1412.29E-08Ce-1436.63E-13Ce-1412.27E-13Ce-1436.34E-09Ce-1438.23E-10Ce-1441.88E-12Ce-1436.91E-15Ce-1441.03E-07Ce-1441.34E-08Pr-1431.84E-12Ce-1441.34E-13 Pr-1437.36E-08Pr-1439.55E-09Pr-1441.90E-12Pr-1439.40E-14Pr-1441.03E-07Pr-1441.34E-08Pr-1441.34E-13Condensate DemineralizerHigh TDS Regenerant Collector TankSecondary Liquid Waste Drain Collector TankSecondary Liquid Waste Monitor TankIsotopeActivity(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)IsotopeActivity(&#xb5;Ci/cc)
CALLAWAY - SPRev. OL-135/03TABLE 12.2-9  CONSERVATIVE BASIS ACCUMULATED RADIOACTIVITY IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATIONIsotopeActivity atPlant Shutdown(curies)Kr-8515317.5Kr-85m10.9 Kr-871.05 Kr-8810.6 Xe-131m128.5 Xe-13316536 Xe-133m145 Xe-13582 Xe-135m0.035 Xe-1380.04 I-1310.207 I-1320.00095 I-1330.04 I-1340.000187 I-1350.0065This table is based on 40 years' continuous operation with 0.25 percent fuel defect. Power is assumed to be 3,565 MWt. The data are based on a volume control tank purge rate of 0.7 scfm and 100 percent stripping efficiency.
CALLAWAY - SPRev. OL-135/03TABLE 12.2-9  CONSERVATIVE BASIS ACCUMULATED RADIOACTIVITY IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATIONIsotopeActivity atPlant Shutdown(curies)Kr-8515317.5Kr-85m10.9 Kr-871.05 Kr-8810.6 Xe-131m128.5 Xe-13316536 Xe-133m145 Xe-13582 Xe-135m0.035 Xe-1380.04 I-1310.207 I-1320.00095 I-1330.04 I-1340.000187 I-1350.0065This table is based on 40 years' continuous operation with 0.25 percent fuel defect. Power is assumed to be 3,565 MWt. The data are based on a volume control tank purge rate of 0.7 scfm and 100 percent stripping efficiency.
CALLAWAY - SPRev. OL-135/03TABLE 12.2-10  DELETEDTable 12.2-10, Sheets 1 through 3, Deleted.
CALLAWAY - SPRev. OL-135/03TABLE 12.2-10  DELETEDTable 12.2-10, Sheets 1 through 3, Deleted.
Line 102: Line 83:
&deg;F  and 95 percent relative humidity and with pool surface area of 2111.5ft 2 1024From fuel storage pool transfer canal and connecting slots into fuel building atmosphere during power (based on pool temperature of 95
&deg;F  and 95 percent relative humidity and with pool surface area of 2111.5ft 2 1024From fuel storage pool transfer canal and connecting slots into fuel building atmosphere during power (based on pool temperature of 95
&deg;F and building air temperature of 70
&deg;F and building air temperature of 70
&deg;F and 50 percent RH and pool surface area of 2111.5ft 2 409c.Partition FactorsHalogensParticulatesTritiumContainment during power1.0010.35Fuel storage and refueling pool surfaces1Negligible1 CALLAWAY - SPTABLE 12.2-11 (Sheet 2)Rev. OL-135/03d.Ventilation RatesCFMContainment during power4,000Containment during refueling20,000 Fuel building during power20,000 Fuel building during refueling20,000e.Volumes of the RegionsCFContainment2.5 x 10 6Fuel building8.2 x 10 5f.Maximum Annual Individual OccupancyHrs/yrContainment during power 5hr/wk for50 wks/year250Containment during fuel handling 10 hr/day for ~ 6.25 days when the refueling pool is full of water62.5Fuel building during power 5hr/week for 50 weeks/year250Fuel building during refueling 10 hrs/day for ~ 10 days and 8hrs/day for ~ 3 days125g.Miscellaneous Information1.Failed fuel percentage for fission products0.122.Reactor coolant specific activitiesTable 11.1-13.Refueling and fuel storage pool concentration (
&deg;F and 50 percent RH and pool surface area of 2111.5ft 2 409c.Partition FactorsHalogensParticulatesTritiumContainment during power1.0010.35Fuel storage and refueling pool surfaces1Negligible1 CALLAWAY - SPTABLE 12.2-11 (Sheet 2)Rev. OL-135/03d.Ventilation RatesCFMContainment during power4,000Containment during refueling20,000 Fuel building during power20,000 Fuel building during refueling20,000e.Volumes of the RegionsCFContainment2.5 x 10 6Fuel building8.2 x 10 5f.Maximum Annual Individual OccupancyHrs/yrContainment during power 5hr/wk for50 wks/year250Containment during fuel handling 10 hr/day for ~ 6.25 days when the refueling pool is full of water62.5Fuel building during power 5hr/week for 50 weeks/year250Fuel building during refueling 10 hrs/day for ~ 10 days and 8hrs/day for ~ 3 days125g.Miscellaneous Information1.Failed fuel percentage for fission products0.122.Reactor coolant specific activitiesTable 11.1-13.Refueling and fuel storage pool concentration (&#xb5;Ci/gm). (These are the maximum concentrations during refueling)I-131I-1333.21 x 10-52.32 x 10-64.Fuel storage pool cleanup rate during power (gpm) (for conservatism, no cleanup of fuel storage pool or refueling pool is assumed during refueling)300 CALLAWAY - SPTABLE 12.2-11 (Sheet 3)Rev. OL-135/035.Decay of isotopes in the pools during refuelingNot included for conservatism6.Duration of refueling pool evaporation (hrs)1507.Duration of fuel storage pool evaporation during refueling (hrs)3208.Duration of fuel storage pool evaporation during power (hours/year)84409.Tritium release to environment via containment ventilation exhaust during refueling (Ci) (based on a total tritium release of 1000 Ci via gaseous effluents, durations of evaporation quoted in items 6, 7 and 8, and evaporation rates from pools given in Item b)2410.Tritium release to environment via fuel building ventilation exhaust during refueling (Ci) (same bases as given for Item9)10411.Tritium release to environment via fuel building ventilation exhaust during power (Ci) (same bases as given for Item9)1094 CALLAWAY - SPRev. OL-135/03TABLE 12.2-12  AIRBORNE RADIOACTIVITY CONCENTRATIONS (&#xb5;Ci/cc)NuclideContainment(power)Containment(refueling)Fuel Building(power)Fuel Building(refueling)H-32.34E-74.7E-63.81E-069.55E-06Cr-511.25E-12negligiblenegligiblenegligibleMn-542.07E-13negligiblenegligiblenegligibleFe-551.07E-12negligiblenegligiblenegligible Fe-596.63E-13negligiblenegligiblenegligibleCo-581.06E-11negligiblenegligiblenegligibleCo-601.33E-12negligiblenegligiblenegligible Br-831.76E-11negligiblenegligiblenegligibleBr-842.61E-12negligiblenegligiblenegligibleKr-83m6.60E-8negligiblenegligiblenegligible Kr-85m6.02E-7negligiblenegligiblenegligibleKr-851.18E-7negligiblenegligiblenegligibleKr-871.39E-7negligiblenegligiblenegligible Kr-888.34E-7negligiblenegligiblenegligibleKr-895.83E-10negligiblenegligiblenegligibleRb-888.01E-7negligiblenegligiblenegligible Rb-895.63E-10negligiblenegligiblenegligibleSr-893.54E-12negligiblenegligiblenegligibleSr-912.49E-13negligiblenegligiblenegligible Mo-995.06E-11negligiblenegligiblenegligibleTe-127m1.86E-13negligiblenegligiblenegligibleTe-1274.01E-13negligiblenegligiblenegligible Te-129m9.26E-13negligiblenegligiblenegligibleTe-1296.58E-13negligiblenegligiblenegligibleTe-131m1.34E-12negligiblenegligiblenegligible Te-1312.69E-13negligiblenegligiblenegligibleTe-1321.65E-11negligiblenegligiblenegligibleI-1301.95E-11negligiblenegligiblenegligible I-1313.82E-92.14E-10negligible7.6E-10I-1323.66E-10negligiblenegligiblenegligibleI-1334.14E-91.55E-11negligible5.5E-11 I-1347.41E-11negligiblenegligiblenegligibleI-1351.34E-9negligiblenegligiblenegligibleXe-131m2.75E-7negligiblenegligiblenegligible Xe-133m1.35E-6negligiblenegligiblenegligibleXe-1337.24E-5negligiblenegligiblenegligibleXe-135m7.64E-9negligiblenegligiblenegligible CALLAWAY - SPTABLE 12.2-12 (Sheet 2)Rev. OL-135/03Xe-1352.55E-6negligiblenegligiblenegligibleXe-1371.26E-9negligiblenegligiblenegligibleXe-1382.22E-8negligiblenegligiblenegligible Cs-1341.67E-11negligiblenegligiblenegligibleCs-1368.48E-12negligiblenegligiblenegligibleCs-1371.20E-11negligiblenegligiblenegligible Cs-1382.07E-8negligiblenegligiblenegligibleBa-137m1.14E-11negligiblenegligiblenegligibleBa-1401.43E-13negligiblenegligiblenegligible La-1401.07E-13negligiblenegligiblenegligibleNOTES:1.Iodine airborne concentrations during refueling are calculated very conservatively, assuming no purification by fuel pool cleanup system and no decay in the pool and a partition factor of 1 at the water-air interface at pool surface.2.Continuous pool cleanup, decay in the pool, and lower evaporation rates are expected to reduce iodine air concentrations to negligible levels during power operation in the fuel building.3.Xe-133 and I-131 air concentrations in the containment ventilation exhaust duct are expected to be 2 x 10-2 and 1.6x10
&#xb5;Ci/gm). (These are the maximum concentrations during refueling)I-131I-1333.21 x 10
-52.32 x 10
-64.Fuel storage pool cleanup rate during power (gpm) (for conservatism, no cleanup of fuel storage pool or refueling pool is assumed during refueling)300 CALLAWAY - SPTABLE 12.2-11 (Sheet 3)Rev. OL-135/035.Decay of isotopes in the pools during refuelingNot included for conservatism6.Duration of refueling pool evaporation (hrs)1507.Duration of fuel storage pool evaporation during refueling (hrs)3208.Duration of fuel storage pool evaporation during power (hours/year)84409.Tritium release to environment via containment ventilation exhaust during refueling (Ci) (based on a total tritium release of 1000 Ci via gaseous effluents, durations of evaporation quoted in items 6, 7 and 8, and evaporation rates from pools given in Item b)2410.Tritium release to environment via fuel building ventilation exhaust during refueling (Ci) (same bases as given for Item9)10411.Tritium release to environment via fuel building ventilation exhaust during power (Ci) (same bases as given for Item9)1094 CALLAWAY - SPRev. OL-135/03TABLE 12.2-12  AIRBORNE RADIOACTIVITY CONCENTRATIONS (
&#xb5;Ci/cc)NuclideContainment(power)Containment(refueling)Fuel Building(power)Fuel Building(refueling)H-32.34E-74.7E-63.81E-069.55E-06Cr-511.25E-12negligiblenegligiblenegligibleMn-542.07E-13negligiblenegligiblenegligibleFe-551.07E-12negligiblenegligiblenegligible Fe-596.63E-13negligiblenegligiblenegligibleCo-581.06E-11negligiblenegligiblenegligibleCo-601.33E-12negligiblenegligiblenegligible Br-831.76E-11negligiblenegligiblenegligibleBr-842.61E-12negligiblenegligiblenegligibleKr-83m6.60E-8negligiblenegligiblenegligible Kr-85m6.02E-7negligiblenegligiblenegligibleKr-851.18E-7negligiblenegligiblenegligibleKr-871.39E-7negligiblenegligiblenegligible Kr-888.34E-7negligiblenegligiblenegligibleKr-895.83E-10negligiblenegligiblenegligibleRb-888.01E-7negligiblenegligiblenegligible Rb-895.63E-10negligiblenegligiblenegligibleSr-893.54E-12negligiblenegligiblenegligibleSr-912.49E-13negligiblenegligiblenegligible Mo-995.06E-11negligiblenegligiblenegligibleTe-127m1.86E-13negligiblenegligiblenegligibleTe-1274.01E-13negligiblenegligiblenegligible Te-129m9.26E-13negligiblenegligiblenegligibleTe-1296.58E-13negligiblenegligiblenegligibleTe-131m1.34E-12negligiblenegligiblenegligible Te-1312.69E-13negligiblenegligiblenegligibleTe-1321.65E-11negligiblenegligiblenegligibleI-1301.95E-11negligiblenegligiblenegligible I-1313.82E-92.14E-10negligible7.6E-10I-1323.66E-10negligiblenegligiblenegligibleI-1334.14E-91.55E-11negligible5.5E-11 I-1347.41E-11negligiblenegligiblenegligibleI-1351.34E-9negligiblenegligiblenegligibleXe-131m2.75E-7negligiblenegligiblenegligible Xe-133m1.35E-6negligiblenegligiblenegligibleXe-1337.24E-5negligiblenegligiblenegligibleXe-135m7.64E-9negligiblenegligiblenegligible CALLAWAY - SPTABLE 12.2-12 (Sheet 2)Rev. OL-135/03Xe-1352.55E-6negligiblenegligiblenegligibleXe-1371.26E-9negligiblenegligiblenegligibleXe-1382.22E-8negligiblenegligiblenegligible Cs-1341.67E-11negligiblenegligiblenegligibleCs-1368.48E-12negligiblenegligiblenegligibleCs-1371.20E-11negligiblenegligiblenegligible Cs-1382.07E-8negligiblenegligiblenegligibleBa-137m1.14E-11negligiblenegligiblenegligibleBa-1401.43E-13negligiblenegligiblenegligible La-1401.07E-13negligiblenegligiblenegligibleNOTES:1.Iodine airborne concentrations during refueling are calculated very conservatively, assuming no purification by fuel pool cleanup system and no decay in the pool and a partition factor of 1 at the water-air interface at pool surface.2.Continuous pool cleanup, decay in the pool, and lower evaporation rates are expected to reduce iodine air concentrations to negligible levels during power operation in the fuel building.3.Xe-133 and I-131 air concentrations in the containment ventilation exhaust duct are expected to be 2 x 10-2 and 1.6x10
-6 &#xb5;Ci/cc, respectively during reactor head venting. However, the containment airborne concentrations for these isotopes are expected to be significantly less during head venting since the radioactivity is directly piped to the exhaust duct. The maximum value for Xe-133 is based on operating plant measurements normalized for a reactor system which has continuous stripping of the noble gases in the volume control tank. The maximum value for I-131 is based on operating plant measurements.NuclideContainment(power)Containment(refueling)Fuel Building(power)Fuel Building(refueling)
-6 &#xb5;Ci/cc, respectively during reactor head venting. However, the containment airborne concentrations for these isotopes are expected to be significantly less during head venting since the radioactivity is directly piped to the exhaust duct. The maximum value for Xe-133 is based on operating plant measurements normalized for a reactor system which has continuous stripping of the noble gases in the volume control tank. The maximum value for I-131 is based on operating plant measurements.NuclideContainment(power)Containment(refueling)Fuel Building(power)Fuel Building(refueling)
CALLAWAY - SP12.3-1Rev. OL-2211/1612.3RADIATIONPROTECTIONDESIGNFEATURES12.3.1FACILITY DESIGN FEATURESIn this section, specific design features for maintaining personnel exposures ALARA, commensurate with the guidance given in Regulatory Guide 8.8, are discussed.12.3.1.1PlantDesignDescriptionforasLowasis ReasonablyAchievable(ALARA)The equipment and plant design features employed to maintain radiation exposures as low as is reasonably achievable are based upon the design considerations of Section 12.1.2 and are outlined in this section for several general classes of equipment (Section 12.3.1.1.1) and several typical plant layout situations (Section 12.3.1.1.2
CALLAWAY - SP12.3-1Rev. OL-2211/1612.3RADIATIONPROTECTIONDESIGNFEATURES12.3.1FACILITY DESIGN FEATURESIn this section, specific design features for maintaining personnel exposures ALARA, commensurate with the guidance given in Regulatory Guide 8.8, are discussed.12.3.1.1PlantDesignDescriptionforasLowasis ReasonablyAchievable(ALARA)The equipment and plant design features employed to maintain radiation exposures as low as is reasonably achievable are based upon the design considerations of Section 12.1.2 and are outlined in this section for several general classes of equipment (Section 12.3.1.1.1) and several typical plant layout situations (Section 12.3.1.1.2
Line 122: Line 99:
CALLAWAY - SP12.3-9Rev. OL-2211/1612.3.2.2GeneralShieldingDesignShielding is provided, as necessary, to attenuate direct radiation and scattered radiation through walls and penetrations to less than the upper limit of the radiation zone for each area shown in Figure 12.3-2. The minimum shielding requirements for all plant areas are given on scaled layout drawings in Section 1.2. General locations of the plant areas and equipment discussed in this section are also shown in the general arrangement drawings of Section 1.2. Design criteria for penetrations comply with the intent of Regulatory Guide 8.8 and are discussed in Section12.3.1.1.2
CALLAWAY - SP12.3-9Rev. OL-2211/1612.3.2.2GeneralShieldingDesignShielding is provided, as necessary, to attenuate direct radiation and scattered radiation through walls and penetrations to less than the upper limit of the radiation zone for each area shown in Figure 12.3-2. The minimum shielding requirements for all plant areas are given on scaled layout drawings in Section 1.2. General locations of the plant areas and equipment discussed in this section are also shown in the general arrangement drawings of Section 1.2. Design criteria for penetrations comply with the intent of Regulatory Guide 8.8 and are discussed in Section12.3.1.1.2
.The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 147 lb/ft
.The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 147 lb/ft
: 3. Whenever poured-in-place concrete has been replaced by concrete blocks or other material, design assures protection on an equivalent shielding basis, as determined by the characteristics of the material selected (Ref. 1). Compliance of concrete radiation shield design with Regulatory Guide 1.69 is discussed in Appendix 3A. Water is used as the primary shield material for areas above the spent fuel transfer and storage areas.For design basis accidents, the reactor building reduces the plant radiation intensities from fission products inside the containment to acceptable emergency levels, as defined by GDC-19, for the control room (see Sections 12.3.2.2.6
: 3. Whenever poured-in-place concrete has been replaced by concrete blocks or other material, design assures protection on an equivalent shielding basis, as determined by the characteristics of the material selected (Ref. 1). Compliance of concrete radiation shield design with Regulatory Guide 1.69 is discussed in Appendix 3A. Water is used as the primary shield material for areas above the spent fuel transfer and storage areas.For design basis accidents, the reactor building reduces the plant radiation intensities from fission products inside the containment to acceptable emergency levels, as defined by GDC-19, for the control room (see Sections 12.3.2.2.6 , 6.4, and 15.6.5).12.3.2.2.1Reactor Building Interior Shielding DesignDuring reactor operation, most areas inside the reactor building are Zone E and normally inaccessible.The main sources of radiation are the reactor vessel and the primary loop components, consisting of the steam generators, pressurizer, reactor coolant pumps, and associated piping. The reactor vessel is shielded by the concrete primary shield, reactor cavity shield, and the concrete secondary shield, which also surrounds all the other primary loop components. Air cooling is provided to prevent overheating, dehydration, and degradation of the shielding and structural properties of the primary shield.The reactor cavity seal/neutron shield, at the RPV seal ledge, attenuates neutron streaming. This seal/shield has been designed to provide equivalent neutron attentuation to 12 inches of water as was previously provided by water bag shielding. MORSE analyses have been performed which predict that limited personnel access may be allowed at the operating floor during power operation. See the response to NRC Question331.1 for a discussion of the Morse analyses. Components of the letdown system are located in shielded areas which are normally Zone E restricted access areas. Shielding is provided for N-16 delay piping, the excess letdown heat exchanger, and the regenerative heat exchanger.
, 6.4, and 15.6.5).12.3.2.2.1Reactor Building Interior Shielding DesignDuring reactor operation, most areas inside the reactor building are Zone E and normally inaccessible.The main sources of radiation are the reactor vessel and the primary loop components, consisting of the steam generators, pressurizer, reactor coolant pumps, and associated piping. The reactor vessel is shielded by the concrete primary shield, reactor cavity shield, and the concrete secondary shield, which also surrounds all the other primary loop components. Air cooling is provided to prevent overheating, dehydration, and degradation of the shielding and structural properties of the primary shield.The reactor cavity seal/neutron shield, at the RPV seal ledge, attenuates neutron streaming. This seal/shield has been designed to provide equivalent neutron attentuation to 12 inches of water as was previously provided by water bag shielding. MORSE analyses have been performed which predict that limited personnel access may be allowed at the operating floor during power operation. See the response to NRC Question331.1 for a discussion of the Morse analyses. Components of the letdown system are located in shielded areas which are normally Zone E restricted access areas. Shielding is provided for N-16 delay piping, the excess letdown heat exchanger, and the regenerative heat exchanger.
CALLAWAY - SP12.3-10Rev. OL-2211/16After shutdown, most of the containment is accessible for limited periods of time, and all access is controlled. Areas are surveyed to establish allowable working periods. Dose rates are expected to range from 0.5 to 1,000 mrem/hr, depending on the location inside the containment (excluding reactor cavity). These dose rates result from residual fission products, neutron-activated materials, and corrosion products in the reactor coolant system.Spent fuel is the primary source of radiation during refueling due to its high buildup of fission product activity. However, radiation levels are limited in areas outside the refueling pool by the shielding effects of the thick structural walls of the refueling pool. The operators involved in the refueling operations are shielded from the spent fuel by the depth of water maintained above the fuel assemblies.12.3.2.2.2Auxiliary Building Shielding DesignDuring normal operation, the major components in the auxiliary building containing potentially high radioactivity are those in the chemical and volume control system. These include the letdown lines, the volume control tank, purification filters and demineralizers, and the charging pumps.Shielding is provided for each piece of equipment consistent with the access and zoning requirements of adjacent areas (Figure12.3-2
CALLAWAY - SP12.3-10Rev. OL-2211/16After shutdown, most of the containment is accessible for limited periods of time, and all access is controlled. Areas are surveyed to establish allowable working periods. Dose rates are expected to range from 0.5 to 1,000 mrem/hr, depending on the location inside the containment (excluding reactor cavity). These dose rates result from residual fission products, neutron-activated materials, and corrosion products in the reactor coolant system.Spent fuel is the primary source of radiation during refueling due to its high buildup of fission product activity. However, radiation levels are limited in areas outside the refueling pool by the shielding effects of the thick structural walls of the refueling pool. The operators involved in the refueling operations are shielded from the spent fuel by the depth of water maintained above the fuel assemblies.12.3.2.2.2Auxiliary Building Shielding DesignDuring normal operation, the major components in the auxiliary building containing potentially high radioactivity are those in the chemical and volume control system. These include the letdown lines, the volume control tank, purification filters and demineralizers, and the charging pumps.Shielding is provided for each piece of equipment consistent with the access and zoning requirements of adjacent areas (Figure12.3-2
).Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to ZoneC access.Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.Following reactor shutdown, the residual heat removal (RHR) system pumps and heat exchangers are in operation to remove heat from the reactor coolant system. The radiation levels in the vicinity of this equipment will temporarily reach Zone E levels due to corrosion and fission products in the reactor water. Shielding is provided to attenuate radiation from RHR equipment during shutdown cooling operations to levels consistent with the radiation zoning requirements of the adjacent areas.12.3.2.2.3Fuel Building Shielding DesignSpent fuel is the primary source of radiation in the fuel building. Because of the extremely high activity of the fission products contained in the spent fuel elements and the proximity of Zone B and C areas, extensive shielding has been provided for areas surrounding the fuel storage pool, cask loading pit and the fuel transfer canal to ensure CALLAWAY - SP12.3-11Rev. OL-2211/16that radiation levels remain below zone levels specified for adjacent areas. Water provides the shielding above the spent fuel assemblies during fuel handling operations.During spent fuel loading operations, the HI-TRAC VW transfer cask provides shielding through a design employing water jackets, steel barriers and lead shielding which minimizes personnel exposure. For a detailed discussion of shielding during ISFSI operations, refer to UMAX FSAR Section 5.The fuel pool cooling and cleanup system (FPCCS) (Section9.1.3) shielding is based on the maximum activity discussed in Section12.2.1 and the access and zoning requirements of adjacent areas. Equipment in the FPCC system to be shielded includes the FPCC heat exchangers, pumps, piping, filters, and demineralizers.12.3.2.2.4Radwaste Building Shielding DesignShielding is provided, as necessary, around the following equipment in the radwaste building to ensure that the radiation zone and access requirements are met for surrounding areas.a.Liquid waste collection tanks and pumpsb.Liquid waste monitor tanks and pumpsc.Chemical drain tank and pump d.Liquid waste boron recycle and secondary liquid waste evaporatorse.Solid radwaste solidification equipmentf.Solid radwaste drumming and storage areas g.Evaporator bottoms tanksh.Liquid radwaste pipingi.Radwaste filters and demineralizers j.Spent resin storage tank and pumpk.Gaseous radwaste surge tank, recombiners, and compressorsl.Gas decay tanksShielding is based upon operation with maximum activity conditions, as discussed in Sections 11.1
).Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to ZoneC access.Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.Following reactor shutdown, the residual heat removal (RHR) system pumps and heat exchangers are in operation to remove heat from the reactor coolant system. The radiation levels in the vicinity of this equipment will temporarily reach Zone E levels due to corrosion and fission products in the reactor water. Shielding is provided to attenuate radiation from RHR equipment during shutdown cooling operations to levels consistent with the radiation zoning requirements of the adjacent areas.12.3.2.2.3Fuel Building Shielding DesignSpent fuel is the primary source of radiation in the fuel building. Because of the extremely high activity of the fission products contained in the spent fuel elements and the proximity of Zone B and C areas, extensive shielding has been provided for areas surrounding the fuel storage pool, cask loading pit and the fuel transfer canal to ensure CALLAWAY - SP12.3-11Rev. OL-2211/16that radiation levels remain below zone levels specified for adjacent areas. Water provides the shielding above the spent fuel assemblies during fuel handling operations.During spent fuel loading operations, the HI-TRAC VW transfer cask provides shielding through a design employing water jackets, steel barriers and lead shielding which minimizes personnel exposure. For a detailed discussion of shielding during ISFSI operations, refer to UMAX FSAR Section 5.The fuel pool cooling and cleanup system (FPCCS) (Section9.1.3) shielding is based on the maximum activity discussed in Section12.2.1 and the access and zoning requirements of adjacent areas. Equipment in the FPCC system to be shielded includes the FPCC heat exchangers, pumps, piping, filters, and demineralizers.12.3.2.2.4Radwaste Building Shielding DesignShielding is provided, as necessary, around the following equipment in the radwaste building to ensure that the radiation zone and access requirements are met for surrounding areas.a.Liquid waste collection tanks and pumpsb.Liquid waste monitor tanks and pumpsc.Chemical drain tank and pump d.Liquid waste boron recycle and secondary liquid waste evaporatorse.Solid radwaste solidification equipmentf.Solid radwaste drumming and storage areas g.Evaporator bottoms tanksh.Liquid radwaste pipingi.Radwaste filters and demineralizers j.Spent resin storage tank and pumpk.Gaseous radwaste surge tank, recombiners, and compressorsl.Gas decay tanksShielding is based upon operation with maximum activity conditions, as discussed in Sections 11.1 , 11.2 , 11.3, and11.4.
, 11.2, 11.3, and11.4.
CALLAWAY - SP12.3-12Rev. OL-2211/16Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to ZoneC access.Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.12.3.2.2.5Turbine Building Shielding DesignRadiation shielding is not required for any process equipment located in the turbine building. All areas in the turbine building are classified ZoneA.12.3.2.2.6Control Room Shielding Design The design basis LOCA dictates the shielding requirements for the control room. Shielding is provided to permit access and occupancy of the control room under LOCA conditions with radiation doses limited to 5 rem whole body from all contributing modes of exposure for the duration of the accident, in accordance with GDC-19. A complete discussion of control room habitability during a LOCA is provided in Section 6.4. Figure 12.3-3 provides an isometric view of the control room shielding.12.3.2.2.7Diesel Generator Building Shielding Design There are no radiation sources in the diesel generator building. Therefore, no shielding is required within the building.12.3.2.2.8Miscellaneous Plant Areas and Plant Yard AreasSufficient shielding is provided for all plant buildings containing radiation sources so that radiation levels at the accessible outside surfaces of the buildings are maintained below Zone A levels. Plant yard areas which are frequently occupied by plant personnel are fully accessible during normal operation and shutdown. These areas are surrounded by a security fence and closed off from areas accessible to the general public. Access to outside storage tanks which have a contact dose rate greater than 0.5 mrem/hr is restricted by a fence located at a distance at which the dose rate is less than 0.5 mrem/hr.12.3.2.2.9Independent Spent Fuel Storage Installation (ISFSI) Shielding DesignThe ISFSI is designed for interim dry storage of spent nuclear fuel. The shielding is sufficient to comply with the requirements of 10 CFR 72.104 and 10 CFR 72.106. The dose rate on the VVM closure lid is calculated to be 0.25 mrem/hr neutron and 0.69 mrem/hr gamma for a total dose rate of 0.94 mrem/hr. The dose rate on the outlet duct screen is calculated to be 0.80 mrem/hr neutron and 1.32 mrem/hr gamma for a total CALLAWAY - SP12.3-13Rev. OL-2211/16dose rate of 2.12 mrem/hr. The dose rate on the side of the HI-TRAC VW transfer cask is calculated to be 237 mrem/hr neutron and 1,611 mrem/hr gamma, for a total dose rate of 1,848 mrem/hr. The regulatory dose limit at the owner-controlled area boundary is 25 mrem annual whole body dose for Normal and Off-Normal conditions and 5 rem whole body for accident conditions.For a detailed discussion of the VVM and HI-TRAC VW transfer cask shielding properties, refer to the HI-STORM UMAX FSAR Chapter 5.12.3.2.3ShieldingCalculationalMethodsThe shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in Section12.2.1. The thickness of each shield wall surrounding the radioactive equipment is determined by approximating, as closely as possible, the actual geometry and physical condition of the source or sources. The isotopic concentrations are converted to energy group sources, using data from the Table of Isotopes (Ref.2).The geometric model (Ref.3-11), assumed for the shielding evaluation of pipe tanks, heat exchangers, filters, demineralizers, evaporators, and the containment is a finite cylindrical volume source. In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source (Ref.3-11). Typical computer codes that are used for shielding analysis are listed in Table12.3-1 (Ref.12-21).The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area. Shielding requirements are evaluated at the point of maximum radiation dose through any wall. Therefore, the actual anticipated radiation levels in the greater region of each plant areas are less than this maximum dose rate and, therefore, less than the radiation zone upper limit.Where shielded entryways to compartments containing high radiation sources are necessary, labyrinths or mazes are used. The mazes are constructed so that the scattered dose rate plus the transmitted dose rate through the shield wall from all contributing sources are below the upper limit of the radiation zone specified for each plant area.12.3.3VENTILATIONThe plant heating, ventilating, and air-conditioning (HVAC) systems are designed to provide a suitable environment for personnel and equipment during normal operation and anticipated operational occurrences. Parts of the plant HVAC systems perform safety-related functions.
CALLAWAY - SP12.3-12Rev. OL-2211/16Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to ZoneC access.Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.12.3.2.2.5Turbine Building Shielding DesignRadiation shielding is not required for any process equipment located in the turbine building. All areas in the turbine building are classified ZoneA.12.3.2.2.6Control Room Shielding Design The design basis LOCA dictates the shielding requirements for the control room. Shielding is provided to permit access and occupancy of the control room under LOCA conditions with radiation doses limited to 5 rem whole body from all contributing modes of exposure for the duration of the accident, in accordance with GDC-19. A complete discussion of control room habitability during a LOCA is provided in Section 6.4
CALLAWAY - SP12.3-14Rev. OL-2211/1612.3.3.1DesignObjectivesThe plant HVAC systems for normal plant operation and anticipated operational occurrences are designed to meet the requirements of 10 CFR 20, "Standards for Protection Against Radiation," and 10 CFR 50, "Licensing of Production and Utilization Facilities."12.3.3.2DesignCriteriaDesign criteria for the plant HVAC systems include the following:a.During normal operation and anticipated operational occurrences, the average and maximum airborne radioactivity levels to which plant personnel are exposed in the restricted areas of the plant are ALARA and within the limits specified in 10CFR20.b.During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary will be ALARA and within the limits specified in 10 CFR 20 and 10 CFR 50.c.The plant siting dose guidelines of 10 CFR 100 will be satisfied, following those hypothetical accidents described in Chapter15.d.The dose to control room personnel shall not exceed the limits specified in GDC-19, following those hypothetical accidents described in Chapter 15.0 and Section6.4
. Figure 12.3-3 provides an isometric view of the control room shielding.12.3.2.2.7Diesel Generator Building Shielding Design There are no radiation sources in the diesel generator building. Therefore, no shielding is required within the building.12.3.2.2.8Miscellaneous Plant Areas and Plant Yard AreasSufficient shielding is provided for all plant buildings containing radiation sources so that radiation levels at the accessible outside surfaces of the buildings are maintained below Zone A levels. Plant yard areas which are frequently occupied by plant personnel are fully accessible during normal operation and shutdown. These areas are surrounded by a security fence and closed off from areas accessible to the general public. Access to outside storage tanks which have a contact dose rate greater than 0.5 mrem/hr is restricted by a fence located at a distance at which the dose rate is less than 0.5 mrem/hr.12.3.2.2.9Independent Spent Fuel Storage Installation (ISFSI) Shielding DesignThe ISFSI is designed for interim dry storage of spent nuclear fuel. The shielding is sufficient to comply with the requirements of 10 CFR 72.104 and 10 CFR 72.106. The dose rate on the VVM closure lid is calculated to be 0.25 mrem/hr neutron and 0.69 mrem/hr gamma for a total dose rate of 0.94 mrem/hr. The dose rate on the outlet duct screen is calculated to be 0.80 mrem/hr neutron and 1.32 mrem/hr gamma for a total CALLAWAY - SP12.3-13Rev. OL-2211/16dose rate of 2.12 mrem/hr. The dose rate on the side of the HI-TRAC VW transfer cask is calculated to be 237 mrem/hr neutron and 1,611 mrem/hr gamma, for a total dose rate of 1,848 mrem/hr. The regulatory dose limit at the owner-controlled area boundary is 25 mrem annual whole body dose for Normal and Off-Normal conditions and 5 rem whole body for accident conditions.For a detailed discussion of the VVM and HI-TRAC VW transfer cask shielding properties, refer to the HI-STORM UMAX FSAR Chapter 5.12.3.2.3ShieldingCalculationalMethodsThe shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in Section12.2.1. The thickness of each shield wall surrounding the radioactive equipment is determined by approximating, as closely as possible, the actual geometry and physical condition of the source or sources. The isotopic concentrations are converted to energy group sources, using data from the Table of Isotopes (Ref.2).The geometric model (Ref.3-11), assumed for the shielding evaluation of pipe tanks, heat exchangers, filters, demineralizers, evaporators, and the containment is a finite cylindrical volume source. In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source (Ref.3-11). Typical computer codes that are used for shielding analysis are listed in Table12.3-1 (Ref.12-21).The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area. Shielding requirements are evaluated at the point of maximum radiation dose through any wall. Therefore, the actual anticipated radiation levels in the greater region of each plant areas are less than this maximum dose rate and, therefore, less than the radiation zone upper limit.Where shielded entryways to compartments containing high radiation sources are necessary, labyrinths or mazes are used. The mazes are constructed so that the scattered dose rate plus the transmitted dose rate through the shield wall from all contributing sources are below the upper limit of the radiation zone specified for each plant area.12.3.3VENTILATIONThe plant heating, ventilating, and air-conditioning (HVAC) systems are designed to provide a suitable environment for personnel and equipment during normal operation and anticipated operational occurrences. Parts of the plant HVAC systems perform safety-related functions.
CALLAWAY - SP12.3-14Rev. OL-2211/1612.3.3.1DesignObjectivesThe plant HVAC systems for normal plant operation and anticipated operational occurrences are designed to meet the requirements of 10 CFR 20, "Standards for Protection Against Radiation," and 10 CFR 50, "Licensing of Production and Utilization Facilities."12.3.3.2DesignCriteriaDesign criteria for the plant HVAC systems include the following:a.During normal operation and anticipated operational occurrences, the average and maximum airborne radioactivity levels to which plant personnel are exposed in the restricted areas of the plant are ALARA and within the limits specified in 10CFR20.b.During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary will be ALARA and within the limits specified in 10 CFR 20 and 10 CFR 50.c.The plant siting dose guidelines of 10 CFR 100 will be satisfied, following those hypothetical accidents described in Chapter15
.d.The dose to control room personnel shall not exceed the limits specified in GDC-19, following those hypothetical accidents described in Chapter 15.0 and Section6.4
.12.3.3.3DesignGuidelinesIn order to accomplish the design objectives, the following guidelines are followed, wherever practicable.12.3.3.3.1Guidelines to Minimize Airborne Radioactivitya.Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.b.Equipment vents and drains are piped directly to a collection device connected to the collection system, instead of allowing any contaminated fluid to flow across the floor to the floor drain.c.All-welded piping systems are employed on contaminated systems, to the maximum extent practicable, to reduce system leakage. If welded piping systems are not employed, drip trays are provided at the points of potential leakage. Drains from drip trays are piped directly to the collection system.
.12.3.3.3DesignGuidelinesIn order to accomplish the design objectives, the following guidelines are followed, wherever practicable.12.3.3.3.1Guidelines to Minimize Airborne Radioactivitya.Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.b.Equipment vents and drains are piped directly to a collection device connected to the collection system, instead of allowing any contaminated fluid to flow across the floor to the floor drain.c.All-welded piping systems are employed on contaminated systems, to the maximum extent practicable, to reduce system leakage. If welded piping systems are not employed, drip trays are provided at the points of potential leakage. Drains from drip trays are piped directly to the collection system.
CALLAWAY - SP12.3-15Rev. OL-2211/16d.Suitable coatings are applied to the concrete floors of potentially contaminated areas to facilitate decontamination.e.To minimize the amount of airborne radioactivity as a result of valve leakage, all valves 2-1/2 inches and larger in the radioactive systems are provided with double sets of packing with lantern rings and with leak-off connections that are piped to drain headers. Diaphragm or bellows seal valves are used on those systems where essentially no leakage can be tolerated.f.Contaminated equipment has design features that minimize the potential for airborne contamination during maintenance operations. These features include flush connections for draining and flushing the pump prior to maintenance and flush connections on piping systems that could become highly radioactive.12.3.3.3.2Guidelines to Control Airborne Radioactivitya.The airflow is directed from areas with lesser potential for contamination to areas with greater potential for contamination.b.In building compartments with a potential for contamination, a greater volumetric flow is exhausted from the area that is supplied to the area to minimize the amount of uncontrolled exfiltration from the area.c.Consideration is given to the possible disruption of normal airflow patterns by maintenance operations, and provisions are made in the design to prevent adverse air flow direction.d.The air cleaning system's design, maintenance, and testing criteria are discussed in detail in the response to Regulatory Guides 1.52 and 1.140 found in Section 9.4. An illustrative example of the air cleaning system design is given in Section12.3.3.5
CALLAWAY - SP12.3-15Rev. OL-2211/16d.Suitable coatings are applied to the concrete floors of potentially contaminated areas to facilitate decontamination.e.To minimize the amount of airborne radioactivity as a result of valve leakage, all valves 2-1/2 inches and larger in the radioactive systems are provided with double sets of packing with lantern rings and with leak-off connections that are piped to drain headers. Diaphragm or bellows seal valves are used on those systems where essentially no leakage can be tolerated.f.Contaminated equipment has design features that minimize the potential for airborne contamination during maintenance operations. These features include flush connections for draining and flushing the pump prior to maintenance and flush connections on piping systems that could become highly radioactive.12.3.3.3.2Guidelines to Control Airborne Radioactivitya.The airflow is directed from areas with lesser potential for contamination to areas with greater potential for contamination.b.In building compartments with a potential for contamination, a greater volumetric flow is exhausted from the area that is supplied to the area to minimize the amount of uncontrolled exfiltration from the area.c.Consideration is given to the possible disruption of normal airflow patterns by maintenance operations, and provisions are made in the design to prevent adverse air flow direction.d.The air cleaning system's design, maintenance, and testing criteria are discussed in detail in the response to Regulatory Guides 1.52 and 1.140 found in Section 9.4. An illustrative example of the air cleaning system design is given in Section12.3.3.5
Line 147: Line 120:
12.3.4.1.1.1Safety Design BasesThe area radiation monitoring system has no function related to the safe shutdown of the plant or the capability to mitigate the consequences of accidents that could result in offsite exposures comparable to the guideline exposure of 10 CFR 100 and, therefore, has no safety design bases. See Appendix7A for a discussion of Regulatory Guide1.97. 12.3.4.1.1.2Power Generation Design BasesPOWER GENERATION DESIGN BASIS ONE - The ARMS functions continuously to immediately alert plant personnel entering or working in nonradiation or low-radiation areas of increasing or abnormally high radiation levels which, if unnoticed, could possibly result in inadvertent overexposures. POWER GENERATION DESIGN BASIS TWO - The ARMS serves to inform the control room operator of the occurrence and approximate location of an abnormal radiation increase in nonradiation or low-radiation areas.POWER GENERATION DESIGN BASIS THREE - The ARMS complies with the requirements of 10 CFR 50, Appendix A, General Design Criterion 63 for monitoring fuel and waste storage and handling areas.
12.3.4.1.1.1Safety Design BasesThe area radiation monitoring system has no function related to the safe shutdown of the plant or the capability to mitigate the consequences of accidents that could result in offsite exposures comparable to the guideline exposure of 10 CFR 100 and, therefore, has no safety design bases. See Appendix7A for a discussion of Regulatory Guide1.97. 12.3.4.1.1.2Power Generation Design BasesPOWER GENERATION DESIGN BASIS ONE - The ARMS functions continuously to immediately alert plant personnel entering or working in nonradiation or low-radiation areas of increasing or abnormally high radiation levels which, if unnoticed, could possibly result in inadvertent overexposures. POWER GENERATION DESIGN BASIS TWO - The ARMS serves to inform the control room operator of the occurrence and approximate location of an abnormal radiation increase in nonradiation or low-radiation areas.POWER GENERATION DESIGN BASIS THREE - The ARMS complies with the requirements of 10 CFR 50, Appendix A, General Design Criterion 63 for monitoring fuel and waste storage and handling areas.
CALLAWAY - SP12.3-19Rev. OL-2211/16POWER GENERATION DESIGN BASIS FOUR - Certain monitors located near the spent fuel pool and new fuel storage vault act as criticality alarm monitors and conform to the requirements of 10 CFR 70, Regulatory Guides 8.5 and 8.12 and Standards ANSI/
CALLAWAY - SP12.3-19Rev. OL-2211/16POWER GENERATION DESIGN BASIS FOUR - Certain monitors located near the spent fuel pool and new fuel storage vault act as criticality alarm monitors and conform to the requirements of 10 CFR 70, Regulatory Guides 8.5 and 8.12 and Standards ANSI/
ANS-8.3-1979 and USASN2.3-1967.12.3.4.1.1.3Codes and Standards Codes and standards applicable to the area radiation monitors are indicated in Table3.2-1
ANS-8.3-1979 and USASN2.3-1967.12.3.4.1.1.3Codes and Standards Codes and standards applicable to the area radiation monitors are indicated in Table3.2-1. 12.3.4.1.2System Description12.3.4.1.2.1General DescriptionThe ARMS consists of five-decade range GM tube detectors located throughout the plant to warn personnel of abnormal gamma radiation levels. The detector signals are transmitted to the control room over individual cables. The displays, both local and in the control room, are five-decade logarithmic ratemeters. The alarms are both audible and visual, and are located in the control room and near the local detectors. 12.3.4.1.2.2Criteria for Area Monitor Selection The following design criteria are applicable to the area radiation monitoring system. RANGE - The ARMS has a five-decade range from 10
. 12.3.4.1.2System Description12.3.4.1.2.1General DescriptionThe ARMS consists of five-decade range GM tube detectors located throughout the plant to warn personnel of abnormal gamma radiation levels. The detector signals are transmitted to the control room over individual cables. The displays, both local and in the control room, are five-decade logarithmic ratemeters. The alarms are both audible and visual, and are located in the control room and near the local detectors. 12.3.4.1.2.2Criteria for Area Monitor Selection The following design criteria are applicable to the area radiation monitoring system. RANGE - The ARMS has a five-decade range from 10
-1 to 10+4 mrem/hr, except for the reactor coolant sample room radiation monitor which has a range from 10 to 10 5 mrem/hr. The ranges are made sufficiently wide to measure the radiation levels expected in the areas concerned. The system continues to read upscale if exposed to radiation levels above the maximum range. SENSITIVITY - Gamma sensitive to photon energies of 100 keV and above.
-1 to 10+4 mrem/hr, except for the reactor coolant sample room radiation monitor which has a range from 10 to 10 5 mrem/hr. The ranges are made sufficiently wide to measure the radiation levels expected in the areas concerned. The system continues to read upscale if exposed to radiation levels above the maximum range. SENSITIVITY - Gamma sensitive to photon energies of 100 keV and above.
RESPONSE - In any range, the readout indicates at least 90 percent of its end point reading within 5 seconds after a step change in radiation level at the detector. ENERGY DEPENDENCE - The dose rate (mrem/hr) readout is within +/-20 percent of the actual dose rate in each detected area from photon energies between 100 keV and 2.5meV. DIRECTIONAL DEPENDENCE - The dose rate readout does not vary more than 10 percent when exposed to a single point radiation source of approximately 1.0MeV from any point in the frontal hemisphere of a horizontal plane.
RESPONSE - In any range, the readout indicates at least 90 percent of its end point reading within 5 seconds after a step change in radiation level at the detector. ENERGY DEPENDENCE - The dose rate (mrem/hr) readout is within +/-20 percent of the actual dose rate in each detected area from photon energies between 100 keV and 2.5meV. DIRECTIONAL DEPENDENCE - The dose rate readout does not vary more than 10 percent when exposed to a single point radiation source of approximately 1.0MeV from any point in the frontal hemisphere of a horizontal plane.
CALLAWAY - SP12.3-20Rev. OL-2211/16ENVIRONMENTAL DEPENDENCE - The system meets the above requirements for all variations of temperature, pressure, and relative humidity within each area monitored, as listed below: For instruments located outside the containment: For instruments located inside the containment:
CALLAWAY - SP12.3-20Rev. OL-2211/16ENVIRONMENTAL DEPENDENCE - The system meets the above requirements for all variations of temperature, pressure, and relative humidity within each area monitored, as listed below: For instruments located outside the containment: For instruments located inside the containment:
EXPOSURE LIFE - Each monitor located inside the containment maintains its characteristics up to an integrated dose of 10 7rads. Each monitor located outside the containment maintains its characteristics up to an integrated dose of 10 6rads. 12.3.4.1.2.3AlarmsEach monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two-alarm setpoints provide sequential alarms on increasing radiation levels. Loss of power will cause an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected. 12.3.4.1.2.4Check Sources Each monitor is provided with a check source, operated from the control room, which simulates a radiation level in the area for operational and gross calibration checks. 12.3.4.1.2.5Power Supplies The power supplies for all of the monitors are given in Table12.3-4
EXPOSURE LIFE - Each monitor located inside the containment maintains its characteristics up to an integrated dose of 10 7rads. Each monitor located outside the containment maintains its characteristics up to an integrated dose of 10 6rads. 12.3.4.1.2.3AlarmsEach monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two-alarm setpoints provide sequential alarms on increasing radiation levels. Loss of power will cause an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected. 12.3.4.1.2.4Check Sources Each monitor is provided with a check source, operated from the control room, which simulates a radiation level in the area for operational and gross calibration checks. 12.3.4.1.2.5Power Supplies The power supplies for all of the monitors are given in Table12.3-4. 12.3.4.1.2.6Calibration and MaintenanceThe area radiation monitors are calibrated by the manufacturer, using a Cs-137 source. The manufacturer's calibration standards are traceable to The National Bureau of Standards primary calibration standard sources and are accurate to at least 5 percent. Temperature,&deg;F60 to 120Humidity, %RH5 to 95PressureAtmosphericTemperature,&deg;F50 to 150Humidity, %RH5 to 100Pressure+/- 2 psig CALLAWAY - SP12.3-21Rev. OL-2211/16The source-detector geometry during this primary calibration is identical to the source-detector geometry in field calibrations. A secondary standard counted in reproducible geometry during the primary calibration is supplied with each monitoring system. Channel checks and source checks are performed at regular intervals to ensure proper monitor function. The monitors are recalibrated at regular intervals, and following repairs or modifications, using the secondary radionuclide standard.12.3.4.1.2.7SensitivitiesEach area radiation monitor is able to detect radiation levels as low as 0.1 mrem/hr, except for post-accident sample room radiation monitor which detects radiation levels as low as 10mrem/hr. 12.3.4.1.2.8Criteria for Location of Area Monitors Generally, area radiation monitors are provided in areas to which personnel normally have access and for which there is a potential for personnel unknowingly to receive high radiation doses (e.g., in excess of 10 CFR 20 limits) in a short period of time because of system failure or improper personnel action. Any plant area which meets one or more of the following criteria is monitored:  a.ZoneA areas which, during normal plant operations, including refueling, could exceed the radiation limit of 0.5 mrem/hr upon system failure or personnel error or which would be continuously occupied following an accident requiring plant shutdown. b.Zone B areas where personnel could otherwise unknowingly receive high levels of radiation exposure due to system failure or personnel error. c.Areas in which quantities of special nuclear material in excess of 10 CFR70.24 quantities are received, stored, or handled, e.g. spent fuel pool and new fuel storage vault. In the case of new fuel, two detectors are provided at the operating floor of the fuel building to serve as criticality monitors. In the case of spent fuel, two detectors are provided on opposite walls above the spent fuel pool to serve as criticality monitors. In both cases, the monitors will provide a distinct audible and visual personnel alarm at no more than 15mrem/hr to signal the need for personnel evacuation. These monitors are provided in accordance with 10 CFR70.24 requirements and will detect the criticality described in 10CFR70.24(a)(1).d.For areas in which fuel is stored or radioactive waste systems and handling equipment are located, area radiation monitors are provided to detect CALLAWAY - SP12.3-22Rev. OL-2211/16conditions that might result in loss of residual heat removal capability and excessive radiation levels and to alert the operators to initiate appropriate safety action in accordance with GDC-63 of 10 CFR 50, AppendixA. The location of each area radiation detector is indicated on the radiation zoning and access control drawings, Figure 12.3-2, and are listed in Table 12.3-2. Consistent with the above criteria, the following general areas are monitored: a.Main control room b.Radwaste building corridorsc.Auxiliary building corridorsd.Fuel storage and handling area e.Radwaste pipe tunnelf.Railway access (fuel building)g.Hot machine and hot instrument shops h.Containmenti.Radwaste solidification areaj.Sampling rooms and laboratory12.3.4.1.2.9Setpoints The bases for the area radiation monitor setpoints are determined from the need to alert operators to abnormal radiation levels in the area. The setpoints are sufficiently above the normal radiation levels in the measured areas to avoid spurious alarms. The normal setpoints for the individual area radiation monitors are provided in Table12.3-2. To support temporary plant conditions or evolutions, the normal setpoint levels may be increased. At these increased levels, access to the area will be controlled in accordance with the Radiation Protection Program based on actual radiation levels. Area radiation monitor setpoints are administratively controlled when not at the normal setpoints.12.3.4.1.2.10Safety Evaluation The ARMS is designed to operate unattended for extended periods of time, detecting and measuring ambient gamma radiation. Ambient radiation dose rate at the detector is indicated locally at the detector and remotely in the main control room. These monitors CALLAWAY - SP12.3-23Rev. OL-2211/16cause an audible and visual alarm at the detector and in the main control room if the radiation levels exceed preset limits. All components are solid state, and the system is designed for high reliability. The system is not essential for safe shutdown of the plant, and it serves no active emergency function during operation. The system serves to warn plant personnel of high radiation levels in various plant areas. All monitors are independent, and failure of one unit has no effect on any other. 12.3.4.2AirborneRadioactivityMonitoringMonitoring for the presence of airborne radioactivity inside the plant is necessary for the protection of plant personnel, in compliance with Regulatory Guide8.2 and within the limits established by 10CFR 20.1-20.601. The airborne radioactivity monitors provide information necessary to ensure that gaseous, particulate, and iodine radioactivity do not exceed 10 MPC hours in areas occupied by the station personnel. The systems consist of permanently installed, continuous monitoring devices together with a program and provisions for specific sample collections and laboratory analyses. 12.3.4.2.1Design Bases The principal objectives and criteria of the airborne radiological monitoring systems (AiRMS) are provided below. 12.3.4.2.1.1Safety Design BasesSAFETY DESIGN BASES - There are no safety design bases for the monitoring of airborne radioactivity for inplant personnel protection. The control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are required to automatically initiate operation of engineered safety features systems in the event that airborne radioactivity in excess of the allowable limits exists. Additional design bases are given in the following sections:  a.Containment purge isolation system, Sections6.2.4 , 7.3.2 , 9.4, and11.5. b.Fuel building ventilation isolation, Sections7.3.3 , 9.4.2, and 11.5. c.Control room intake isolation, Sections6.4.1 , 7.3.4 , 9.4.1 , and 11.5. These radioactivity monitors are protection system elements and are designed in accordance with IEEE Standard 279 due to safety design bases of 6.2.4 , 6.4.1 , 7.3 , 9.4 , and11.5.
. 12.3.4.1.2.6Calibration and MaintenanceThe area radiation monitors are calibrated by the manufacturer, using a Cs-137 source. The manufacturer's calibration standards are traceable to The National Bureau of Standards primary calibration standard sources and are accurate to at least 5 percent. Temperature,&deg;F60 to 120Humidity, %RH5 to 95PressureAtmosphericTemperature,&deg;F50 to 150Humidity, %RH5 to 100Pressure+/- 2 psig CALLAWAY - SP12.3-21Rev. OL-2211/16The source-detector geometry during this primary calibration is identical to the source-detector geometry in field calibrations. A secondary standard counted in reproducible geometry during the primary calibration is supplied with each monitoring system. Channel checks and source checks are performed at regular intervals to ensure proper monitor function. The monitors are recalibrated at regular intervals, and following repairs or modifications, using the secondary radionuclide standard.12.3.4.1.2.7SensitivitiesEach area radiation monitor is able to detect radiation levels as low as 0.1 mrem/hr, except for post-accident sample room radiation monitor which detects radiation levels as low as 10mrem/hr. 12.3.4.1.2.8Criteria for Location of Area Monitors Generally, area radiation monitors are provided in areas to which personnel normally have access and for which there is a potential for personnel unknowingly to receive high radiation doses (e.g., in excess of 10 CFR 20 limits) in a short period of time because of system failure or improper personnel action. Any plant area which meets one or more of the following criteria is monitored:  a.ZoneA areas which, during normal plant operations, including refueling, could exceed the radiation limit of 0.5 mrem/hr upon system failure or personnel error or which would be continuously occupied following an accident requiring plant shutdown. b.Zone B areas where personnel could otherwise unknowingly receive high levels of radiation exposure due to system failure or personnel error. c.Areas in which quantities of special nuclear material in excess of 10 CFR70.24 quantities are received, stored, or handled, e.g. spent fuel pool and new fuel storage vault. In the case of new fuel, two detectors are provided at the operating floor of the fuel building to serve as criticality monitors. In the case of spent fuel, two detectors are provided on opposite walls above the spent fuel pool to serve as criticality monitors. In both cases, the monitors will provide a distinct audible and visual personnel alarm at no more than 15mrem/hr to signal the need for personnel evacuation. These monitors are provided in accordance with 10 CFR70.24 requirements and will detect the criticality described in 10CFR70.24(a)(1).d.For areas in which fuel is stored or radioactive waste systems and handling equipment are located, area radiation monitors are provided to detect CALLAWAY - SP12.3-22Rev. OL-2211/16conditions that might result in loss of residual heat removal capability and excessive radiation levels and to alert the operators to initiate appropriate safety action in accordance with GDC-63 of 10 CFR 50, AppendixA. The location of each area radiation detector is indicated on the radiation zoning and access control drawings, Figure 12.3-2, and are listed in Table 12.3-2. Consistent with the above criteria, the following general areas are monitored: a.Main control room b.Radwaste building corridorsc.Auxiliary building corridorsd.Fuel storage and handling area e.Radwaste pipe tunnelf.Railway access (fuel building)g.Hot machine and hot instrument shops h.Containmenti.Radwaste solidification areaj.Sampling rooms and laboratory12.3.4.1.2.9Setpoints The bases for the area radiation monitor setpoints are determined from the need to alert operators to abnormal radiation levels in the area. The setpoints are sufficiently above the normal radiation levels in the measured areas to avoid spurious alarms. The normal setpoints for the individual area radiation monitors are provided in Table12.3-2. To support temporary plant conditions or evolutions, the normal setpoint levels may be increased. At these increased levels, access to the area will be controlled in accordance with the Radiation Protection Program based on actual radiation levels. Area radiation monitor setpoints are administratively controlled when not at the normal setpoints.12.3.4.1.2.10Safety Evaluation The ARMS is designed to operate unattended for extended periods of time, detecting and measuring ambient gamma radiation. Ambient radiation dose rate at the detector is indicated locally at the detector and remotely in the main control room. These monitors CALLAWAY - SP12.3-23Rev. OL-2211/16cause an audible and visual alarm at the detector and in the main control room if the radiation levels exceed preset limits. All components are solid state, and the system is designed for high reliability. The system is not essential for safe shutdown of the plant, and it serves no active emergency function during operation. The system serves to warn plant personnel of high radiation levels in various plant areas. All monitors are independent, and failure of one unit has no effect on any other. 12.3.4.2AirborneRadioactivityMonitoringMonitoring for the presence of airborne radioactivity inside the plant is necessary for the protection of plant personnel, in compliance with Regulatory Guide8.2 and within the limits established by 10CFR 20.1-20.601. The airborne radioactivity monitors provide information necessary to ensure that gaseous, particulate, and iodine radioactivity do not exceed 10 MPC hours in areas occupied by the station personnel. The systems consist of permanently installed, continuous monitoring devices together with a program and provisions for specific sample collections and laboratory analyses. 12.3.4.2.1Design Bases The principal objectives and criteria of the airborne radiological monitoring systems (AiRMS) are provided below. 12.3.4.2.1.1Safety Design BasesSAFETY DESIGN BASES - There are no safety design bases for the monitoring of airborne radioactivity for inplant personnel protection. The control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are required to automatically initiate operation of engineered safety features systems in the event that airborne radioactivity in excess of the allowable limits exists. Additional design bases are given in the following sections:  a.Containment purge isolation system, Sections6.2.4
CALLAWAY - SP12.3-24Rev. OL-2211/16The safety evaluation of these systems is discussed in Section7.3. These monitors also serve for inplant worker protection, and this function is discussed at length in this section. 12.3.4.2.1.2Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The airborne radioactivity monitors operate continuously to detect airborne particulates, iodine, and/or noble gases in the air upstream of all filters in the containment, auxiliary building, fuel building, radwaste building, waste gas decay tank rooms, access control area, and control room for the protection of the workers. POWER GENERATION DESIGN BASIS TWO - The airborne radioactivity monitors are designed to detect 10 MPC-hours or better in any compartment or room served by the monitoring system. POWER GENERATION DESIGN BASIS THREE - The containment atmosphere monitors are designed to detect leakage of radioactivity from the reactor coolant system into the containment atmosphere. This function is described in greater detail in Section5.2.5. The containment purge monitors will serve as a backup to the containment atmosphere monitors while the purge is in operation. 12.3.4.2.1.3Codes and Standards Codes and standards applicable to the airborne radioactivity monitors are indicated in Table3.2-1. The monitors listed in Section12.3.4.2.1.1 have additional codes and standards applied due to their safety-related functions discussed in other sections, as noted above. 12.3.4.2.2System Description 12.3.4.2.2.1General Description 12.3.4.2.2.1.1Data Collection The AiRMS consist of particulate, iodine, and noble gas monitors with the attendant controls, alarms, pumps, valves, and indicators required to meet the design objectives in Section 12.3.4.2.1. Each monitor consists of the detector assembly and a local microprocessor. The local microprocessor processes the detector assembly signal in digital form, computes average radioactivity levels, stores data, performs alarm or control functions, and transmits the digital signal to the control room microprocessor. Signal transmission is accomplished via two two-wire daisy-chain loops. Each loop allows data transmission in either direction, ensuring that a single fault in the loop will not prevent the control room microprocessor from receiving the data.
, 7.3.2, 9.4, and11.5. b.Fuel building ventilation isolation, Sections7.3.3
, 9.4.2, and 11.5. c.Control room intake isolation, Sections6.4.1
, 7.3.4, 9.4.1, and 11.5. These radioactivity monitors are protection system elements and are designed in accordance with IEEE Standard 279 due to safety design bases of 6.2.4, 6.4.1, 7.3, 9.4, and11.5.
CALLAWAY - SP12.3-24Rev. OL-2211/16The safety evaluation of these systems is discussed in Section7.3
. These monitors also serve for inplant worker protection, and this function is discussed at length in this section. 12.3.4.2.1.2Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The airborne radioactivity monitors operate continuously to detect airborne particulates, iodine, and/or noble gases in the air upstream of all filters in the containment, auxiliary building, fuel building, radwaste building, waste gas decay tank rooms, access control area, and control room for the protection of the workers. POWER GENERATION DESIGN BASIS TWO - The airborne radioactivity monitors are designed to detect 10 MPC-hours or better in any compartment or room served by the monitoring system. POWER GENERATION DESIGN BASIS THREE - The containment atmosphere monitors are designed to detect leakage of radioactivity from the reactor coolant system into the containment atmosphere. This function is described in greater detail in Section5.2.5. The containment purge monitors will serve as a backup to the containment atmosphere monitors while the purge is in operation. 12.3.4.2.1.3Codes and Standards Codes and standards applicable to the airborne radioactivity monitors are indicated in Table3.2-1. The monitors listed in Section12.3.4.2.1.1 have additional codes and standards applied due to their safety-related functions discussed in other sections, as noted above. 12.3.4.2.2System Description 12.3.4.2.2.1General Description 12.3.4.2.2.1.1Data Collection The AiRMS consist of particulate, iodine, and noble gas monitors with the attendant controls, alarms, pumps, valves, and indicators required to meet the design objectives in Section 12.3.4.2.1. Each monitor consists of the detector assembly and a local microprocessor. The local microprocessor processes the detector assembly signal in digital form, computes average radioactivity levels, stores data, performs alarm or control functions, and transmits the digital signal to the control room microprocessor. Signal transmission is accomplished via two two-wire daisy-chain loops. Each loop allows data transmission in either direction, ensuring that a single fault in the loop will not prevent the control room microprocessor from receiving the data.
CALLAWAY - SP12.3-25Rev. OL-2211/16The local microprocessors for monitors which perform safety functions (control room ventilation, fuel building ventilation, containment atmosphere, and containment purge monitors, refer to Section11.5) are wired directly to individual indicators located on the seismic CategoryI AiRMS cabinets in the control room. The input from the safety-related channels to the daisy-chain loop is an isolated signal to ensure that the safety-related signals will not be affected by signals or conditions existing in the nonsafety portion of the system. The control room microprocessor provides controls and indication for the AiRMS. Indication is via a Visual Display located in the control room. The signals from each monitor may also be recorded on a system printer. The safety-related monitors are also recorded on analog strip chart recorders. 12.3.4.2.2.1.2Selection Criteria for Airborne Monitors12.3.4.2.2.1.2.1IntroductionThe type of fixed instrumentation used for monitoring airborne radioactivity is offline. The offline system extracts a sample from the process stream and transports that sample to the radiation monitoring system which contains the specified equipment to detect particulates, halogens, and/or noble gases. 12.3.4.2.2.1.2.2Sampling Criteria The sampling system for the particulate/halogen/noble gas monitors is designed and installed in accordance with ANSI N13.1-1969 guide to sampling of airborne radioactive materials. Systems whose sensitivity is dependent upon sample flow employ isokinetic nozzles and suitable control of the flow rate. 12.3.4.2.2.1.2.3Detection Criteria Since both radioactive particulates and radioactive noble gases are beta emitters, beta-sensitive scintillation detectors are used to sense radioactivity to minimize the effects due to background radiation and, consequently, obtain a lower minimum detectable concentration. Where spectrometric analysis is required (such as in iodine monitoring) an NaI (Tl) gamma scintillation detector assembly is employed. 12.3.4.2.2.1.2.4Instrumentation Criteria Instrumentation necessary to indicate, alarm, and perform control functions is provided to complete the monitoring system. Since radioactive concentrations may vary substantially, wide-range instruments are utilized. All airborne radiation monitors include provisions for obtaining a gas sample for laboratory isotopic analysis. The particulate and charcoal filters can readily be removed for laboratory analyses.
CALLAWAY - SP12.3-25Rev. OL-2211/16The local microprocessors for monitors which perform safety functions (control room ventilation, fuel building ventilation, containment atmosphere, and containment purge monitors, refer to Section11.5) are wired directly to individual indicators located on the seismic CategoryI AiRMS cabinets in the control room. The input from the safety-related channels to the daisy-chain loop is an isolated signal to ensure that the safety-related signals will not be affected by signals or conditions existing in the nonsafety portion of the system. The control room microprocessor provides controls and indication for the AiRMS. Indication is via a Visual Display located in the control room. The signals from each monitor may also be recorded on a system printer. The safety-related monitors are also recorded on analog strip chart recorders. 12.3.4.2.2.1.2Selection Criteria for Airborne Monitors12.3.4.2.2.1.2.1IntroductionThe type of fixed instrumentation used for monitoring airborne radioactivity is offline. The offline system extracts a sample from the process stream and transports that sample to the radiation monitoring system which contains the specified equipment to detect particulates, halogens, and/or noble gases. 12.3.4.2.2.1.2.2Sampling Criteria The sampling system for the particulate/halogen/noble gas monitors is designed and installed in accordance with ANSI N13.1-1969 guide to sampling of airborne radioactive materials. Systems whose sensitivity is dependent upon sample flow employ isokinetic nozzles and suitable control of the flow rate. 12.3.4.2.2.1.2.3Detection Criteria Since both radioactive particulates and radioactive noble gases are beta emitters, beta-sensitive scintillation detectors are used to sense radioactivity to minimize the effects due to background radiation and, consequently, obtain a lower minimum detectable concentration. Where spectrometric analysis is required (such as in iodine monitoring) an NaI (Tl) gamma scintillation detector assembly is employed. 12.3.4.2.2.1.2.4Instrumentation Criteria Instrumentation necessary to indicate, alarm, and perform control functions is provided to complete the monitoring system. Since radioactive concentrations may vary substantially, wide-range instruments are utilized. All airborne radiation monitors include provisions for obtaining a gas sample for laboratory isotopic analysis. The particulate and charcoal filters can readily be removed for laboratory analyses.
CALLAWAY - SP12.3-26Rev. OL-2211/16The airborne particulate monitors each consist of a fixed filter upon which radioactive particulate matter is deposited by means of a positive displacement pump that draws a continuous sample, using an isokinetic nozzle from the ventilation exhaust duct for the particular area. The fixed filter is located in front of a beta scintillation detector coupled to a photomultiplier tube which responds to the scintillations emitted from the crystal as a result of incident radiation giving up its energy within the crystal. Each airborne iodine monitor consists of a charcoal cartridge upon which iodine is adsorbed. The air sample is prefiltered to remove particulates. The charcoal cartridge is located in front of a gamma scintillation detector coupled to a photomultiplier tube. Each airborne noble gas monitor consists of a fixed volume sample chamber through which prefiltered sample air is passed. A beta scintillation detector is located within the sample chamber to detect the activity level of the air sample. All of the detectors and sample chambers are enclosed in heavily shielded lead pigs. Two motor-operated valves, operated locally, are provided to permit air purging of the sample chamber to facilitate background activity checks. The sensitivities and alarm setpoints are given in Table12.3-3. The high-alarm points are based on the most restrictive isotopes which are expected to be present. 12.3.4.2.2.1.3Criteria for Airborne Radioactivity Monitor LocationsThe criteria for locating airborne radioactivity monitors are dependent upon the point of leakage, the ability to identify the source of radioactivity so that corrective action may be performed, and whether personnel may be exposed to the airborne radioactivity. a.Airborne radioactivity monitors sample normally accessible personnel operating areas for which there is a potential for airborne radioactivity. b.Areas not normally accessible are monitored, prior to personnel entry, with portable monitors or samplers, depending upon the potential for airborne radioactivity and the work to be performed in the area. c.Exhaust ducts servicing an area containing processes which, in the event of major leakage, could result in concentrations within the plant approaching the limits established by 10CFR20.1-20.601 for plant workers are monitored. d.Dilution from other exhaust ducts is considered when locating monitors in exhaust systems to ensure maximum coverage and still be able to detect 10 CFR 20.1-20.601 airborne radioactivity limits in the area with the lowest ventilation flow.
CALLAWAY - SP12.3-26Rev. OL-2211/16The airborne particulate monitors each consist of a fixed filter upon which radioactive particulate matter is deposited by means of a positive displacement pump that draws a continuous sample, using an isokinetic nozzle from the ventilation exhaust duct for the particular area. The fixed filter is located in front of a beta scintillation detector coupled to a photomultiplier tube which responds to the scintillations emitted from the crystal as a result of incident radiation giving up its energy within the crystal. Each airborne iodine monitor consists of a charcoal cartridge upon which iodine is adsorbed. The air sample is prefiltered to remove particulates. The charcoal cartridge is located in front of a gamma scintillation detector coupled to a photomultiplier tube. Each airborne noble gas monitor consists of a fixed volume sample chamber through which prefiltered sample air is passed. A beta scintillation detector is located within the sample chamber to detect the activity level of the air sample. All of the detectors and sample chambers are enclosed in heavily shielded lead pigs. Two motor-operated valves, operated locally, are provided to permit air purging of the sample chamber to facilitate background activity checks. The sensitivities and alarm setpoints are given in Table12.3-3. The high-alarm points are based on the most restrictive isotopes which are expected to be present. 12.3.4.2.2.1.3Criteria for Airborne Radioactivity Monitor LocationsThe criteria for locating airborne radioactivity monitors are dependent upon the point of leakage, the ability to identify the source of radioactivity so that corrective action may be performed, and whether personnel may be exposed to the airborne radioactivity. a.Airborne radioactivity monitors sample normally accessible personnel operating areas for which there is a potential for airborne radioactivity. b.Areas not normally accessible are monitored, prior to personnel entry, with portable monitors or samplers, depending upon the potential for airborne radioactivity and the work to be performed in the area. c.Exhaust ducts servicing an area containing processes which, in the event of major leakage, could result in concentrations within the plant approaching the limits established by 10CFR20.1-20.601 for plant workers are monitored. d.Dilution from other exhaust ducts is considered when locating monitors in exhaust systems to ensure maximum coverage and still be able to detect 10 CFR 20.1-20.601 airborne radioactivity limits in the area with the lowest ventilation flow.
CALLAWAY - SP12.3-27Rev. OL-2211/16e.Outside air intake ducts for the control building are monitored to measure possible introduction of radioactive materials into the control room to ensure the habitability of those areas requiring personnel occupancy for safe shutdown. f.Airborne radioactivity monitors are located so that the actual sample chamber and detector location are in an area where the background radiation is low. Detailed physical locations are provided on the radiation zone drawings, Figure12.3-2
CALLAWAY - SP12.3-27Rev. OL-2211/16e.Outside air intake ducts for the control building are monitored to measure possible introduction of radioactive materials into the control room to ensure the habitability of those areas requiring personnel occupancy for safe shutdown. f.Airborne radioactivity monitors are located so that the actual sample chamber and detector location are in an area where the background radiation is low. Detailed physical locations are provided on the radiation zone drawings, Figure12.3-2. 12.3.4.2.2.1.4AlarmsEach monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two alarm setpoints provide sequential alarms on increasing radioactivity levels. Loss of power will cause an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected. Alarms from the AiRMS are provided in the control room on the plant annunciator (audible and visual), the balance-of-plant computer alarm display (visual), and the AiRMS Visual Display. In addition, the safety-related channels have individual alarm lights on the safety-related indicators on the AiRMS control panel. The balance-of-plant and AiRMS computers also provide printouts of each alarm. The pumping systems are controlled from the control room and are provided with a low-flow alarm to alert the operator of pump failure or any other condition which causes a loss of flow through the sample system, and a flow control valve and flow controller to automatically compensate for filter loading. 12.3.4.2.2.1.5Check Sources Each monitor is provided with a check source, operated from the control room, which simulates a radioactive sample in the detector assembly for operational and gross calibration checks. 12.3.4.2.2.1.6Power Supplies All Class 1E inplant AiRMS are powered from Class 1E motor control centers. The power supplies for all of the monitors are given in Table12.3-4. 12.3.4.2.2.1.7Calibration and MaintenanceThe airborne radioactivity monitors are calibrated by the manufacturer for the principal radionuclides listed in Table 12.3-3. The manufacturer's calibration standards are traceable to National Bureau of Standards primary calibration standard sources and are accurate to at least 5 percent. The source detector geometry during this primary CALLAWAY - SP12.3-28Rev. OL-2211/16calibration is identical to the sample detector geometry in actual use. Secondary standards counted in reproducible geometry during the primary calibration are supplied with each continuous monitor. Channel checks and source checks are performed at regular intervals to ensure proper monitor function. The monitors are recalibrated at regular intervals, and following repairs or modifications, using the secondary radionuclide standard.12.3.4.2.2.1.8Sensitivities The AiRMS is capable of detecting 10 MPC-hours of airborne radioactivity. The most restrictive isotope for each type of monitor is that isotope with the lowest worker maximum permissible concentration (WMPC), as defined in TableI, ColumnI, of Appendix B to 10 CFR 20.1-20.601.For the containment atmosphere and containment purge system monitors, the High and Alert alarm set points are based on the Offsite Dose Calculation Manual. The sensitivities and alarm setpoints are given in Table 12.3-3. The High alarm points are based on the most restrictive isotopes which are expected to be present. The concentration levels are as defined in TableI, ColumnI, of AppendixB to 10 CFR 20.1-20.601 or Technical Specification limits, considering dilution. The sensitivity of the airborne radioactivity monitors is based on a 95 percent confidence level for 0.5MeV beta or gamma radiation in a 1mr/hr gamma radiation background at standard pressure and ambient temperature. The fixed volume noble gas detector assemblies have a minimum detectable concentration of 2x10
. 12.3.4.2.2.1.4AlarmsEach monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two alarm setpoints provide sequential alarms on increasing radioactivity levels. Loss of power will cause an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected. Alarms from the AiRMS are provided in the control room on the plant annunciator (audible and visual), the balance-of-plant computer alarm display (visual), and the AiRMS Visual Display. In addition, the safety-related channels have individual alarm lights on the safety-related indicators on the AiRMS control panel. The balance-of-plant and AiRMS computers also provide printouts of each alarm. The pumping systems are controlled from the control room and are provided with a low-flow alarm to alert the operator of pump failure or any other condition which causes a loss of flow through the sample system, and a flow control valve and flow controller to automatically compensate for filter loading. 12.3.4.2.2.1.5Check Sources Each monitor is provided with a check source, operated from the control room, which simulates a radioactive sample in the detector assembly for operational and gross calibration checks. 12.3.4.2.2.1.6Power Supplies All Class 1E inplant AiRMS are powered from Class 1E motor control centers. The power supplies for all of the monitors are given in Table12.3-4
. 12.3.4.2.2.1.7Calibration and MaintenanceThe airborne radioactivity monitors are calibrated by the manufacturer for the principal radionuclides listed in Table 12.3-3. The manufacturer's calibration standards are traceable to National Bureau of Standards primary calibration standard sources and are accurate to at least 5 percent. The source detector geometry during this primary CALLAWAY - SP12.3-28Rev. OL-2211/16calibration is identical to the sample detector geometry in actual use. Secondary standards counted in reproducible geometry during the primary calibration are supplied with each continuous monitor. Channel checks and source checks are performed at regular intervals to ensure proper monitor function. The monitors are recalibrated at regular intervals, and following repairs or modifications, using the secondary radionuclide standard.12.3.4.2.2.1.8Sensitivities The AiRMS is capable of detecting 10 MPC-hours of airborne radioactivity. The most restrictive isotope for each type of monitor is that isotope with the lowest worker maximum permissible concentration (WMPC), as defined in TableI, ColumnI, of Appendix B to 10 CFR 20.1-20.601.For the containment atmosphere and containment purge system monitors, the High and Alert alarm set points are based on the Offsite Dose Calculation Manual. The sensitivities and alarm setpoints are given in Table 12.3-3. The High alarm points are based on the most restrictive isotopes which are expected to be present. The concentration levels are as defined in TableI, ColumnI, of AppendixB to 10 CFR 20.1-20.601 or Technical Specification limits, considering dilution. The sensitivity of the airborne radioactivity monitors is based on a 95 percent confidence level for 0.5MeV beta or gamma radiation in a 1mr/hr gamma radiation background at standard pressure and ambient temperature. The fixed volume noble gas detector assemblies have a minimum detectable concentration of 2x10
-7 &#xb5;Ci/cc, using Kr-85/Xe-133 as the limiting isotope. The fixed filter particulate detector assemblies have a minimum detectable concentration of 1x10-11 &#xb5;Ci/cc, using Cs-137 as the limiting isotope. The filter assembly has a collection efficiency of 99percent for particles of 0.3micron or larger. The charcoal filter halogen detector assemblies have a minimum detectable concentration of 1 x 10-11 &#xb5;Ci/cc, using I-131 as the limiting isotope. The charcoal filter assembly has a collection efficiency of at least 95percent for iodine. 12.3.4.2.2.1.9Ranges and SetpointsThe ranges of the various airborne radioactivity monitors were chosen based on the detection of radioactivity in concentrations ranging from 10 MPC-hours or lower in compartments served up to those from postulated accidents.
-7 &#xb5;Ci/cc, using Kr-85/Xe-133 as the limiting isotope. The fixed filter particulate detector assemblies have a minimum detectable concentration of 1x10-11 &#xb5;Ci/cc, using Cs-137 as the limiting isotope. The filter assembly has a collection efficiency of 99percent for particles of 0.3micron or larger. The charcoal filter halogen detector assemblies have a minimum detectable concentration of 1 x 10-11 &#xb5;Ci/cc, using I-131 as the limiting isotope. The charcoal filter assembly has a collection efficiency of at least 95percent for iodine. 12.3.4.2.2.1.9Ranges and SetpointsThe ranges of the various airborne radioactivity monitors were chosen based on the detection of radioactivity in concentrations ranging from 10 MPC-hours or lower in compartments served up to those from postulated accidents.
CALLAWAY - SP12.3-29Rev. OL-2211/16The fixed volume noble gas detector assemblies have a range of 10
CALLAWAY - SP12.3-29Rev. OL-2211/16The fixed volume noble gas detector assemblies have a range of 10
-7 to10-2 &#xb5;Ci/cc. The fixed filter particulate detector assemblies have a range of 10-12 to 10-7 &#xb5;Ci/cc. The charcoal filter halogen detector assemblies have a range of 10-11 to 10-6 &#xb5;Ci/cc. The setpoints are chosen to alert the operators to airborne radioactivity that might be present so that 10 CFR 20.1-20.601 limits on worker exposure or Technical Specification limits will not be exceeded. The setpoints for control of ventilation are discussed in Section7.3
-7 to10-2 &#xb5;Ci/cc. The fixed filter particulate detector assemblies have a range of 10-12 to 10-7 &#xb5;Ci/cc. The charcoal filter halogen detector assemblies have a range of 10-11 to 10-6 &#xb5;Ci/cc. The setpoints are chosen to alert the operators to airborne radioactivity that might be present so that 10 CFR 20.1-20.601 limits on worker exposure or Technical Specification limits will not be exceeded. The setpoints for control of ventilation are discussed in Section7.3. The setpoints on the monitors used for reactor coolant pressure boundary leakage detection are discussed in Section5.2.5. The ranges and setpoints for the airborne radioactivity monitors are provided in Table12.3-3. 12.3.4.2.2.1.10Expected System Parameters The expected ranges of system parameters, such as flow rate, composition, and concentrations, are summarized in Table12.3-3. Detailed information on the individual HVAC systems can be found in Section9.4. 12.3.4.2.2.2Monitoring SystemsThe systems discussed in the following sections are summarized in Table12.3-3. 12.3.4.2.2.2.1Access Control Area Ventilation Exhaust Radioactivity MonitorThe access control area ventilation exhaust radioactivity monitor, 0-GK-RE-41, continuously monitors for particulate radioactivity in the access control area ventilation exhaust upstream of the HVAC filters. The sample is extracted from the duct through an isokinetic nozzle, in accordance with ANSI Standard N13.1-1969, to ensure that a representative sample of the system is obtained. After passing through the fixed filter detector assembly and the pumping system, the sample is discharged back to the duct. The high and high-high alarms function to alert the operator to airborne particulate radioactivity in the access control area. Noble gases are monitored by the unit vent monitor described in Section11.5. The airborne iodine radioactivity in the access control area can be inferred from the particulate and noble gas radioactivity levels. Since no radioactive equipment is in the area, there is no source for release of iodine or noble gas. Particulates may be present due to dust on the protective clothing becoming airborne.
. The setpoints on the monitors used for reactor coolant pressure boundary leakage detection are discussed in Section5.2.5
. The ranges and setpoints for the airborne radioactivity monitors are provided in Table12.3-3
. 12.3.4.2.2.1.10Expected System Parameters The expected ranges of system parameters, such as flow rate, composition, and concentrations, are summarized in Table12.3-3. Detailed information on the individual HVAC systems can be found in Section9.4
. 12.3.4.2.2.2Monitoring SystemsThe systems discussed in the following sections are summarized in Table12.3-3
. 12.3.4.2.2.2.1Access Control Area Ventilation Exhaust Radioactivity MonitorThe access control area ventilation exhaust radioactivity monitor, 0-GK-RE-41, continuously monitors for particulate radioactivity in the access control area ventilation exhaust upstream of the HVAC filters. The sample is extracted from the duct through an isokinetic nozzle, in accordance with ANSI Standard N13.1-1969, to ensure that a representative sample of the system is obtained. After passing through the fixed filter detector assembly and the pumping system, the sample is discharged back to the duct. The high and high-high alarms function to alert the operator to airborne particulate radioactivity in the access control area. Noble gases are monitored by the unit vent monitor described in Section11.5. The airborne iodine radioactivity in the access control area can be inferred from the particulate and noble gas radioactivity levels. Since no radioactive equipment is in the area, there is no source for release of iodine or noble gas. Particulates may be present due to dust on the protective clothing becoming airborne.
CALLAWAY - SP12.3-30Rev. OL-2211/16Further determination of the source and the corrective action to be taken will be based on monitoring with portable detection and sampling equipment. Indication for this system is provided on the AiRMS Visual Display in the control room. 12.3.4.2.2.2.2Deleted 12.3.4.2.2.2.3Waste Gas Decay Tank Area Ventilation Exhaust Radioactivity MonitorThe waste gas decay tank area ventilation radioactivity monitor, 0-GH-RE-23, continuously monitors for gaseous radioactivity in the discharge duct from the waste gas decay tank area upstream of the radwaste building exhaust filter adsorber. The sample point provides rapid detection of a leak in the waste gas processing system and, in conjunction with the radwaste building exhaust radioactivity monitor and the radwaste building effluent monitor, helps localize the affected area in the event of an alarm on either monitor.The sample is extracted from the exhaust duct and passed through the fixed volume noble gas detector assembly and the pumping system. Then the sample is discharged back to the duct. The high alarm provides indication of a leak in the decay tanks, compressors, piping, or valves. The high-high alarm indicates that concentrations in the decay tank room are at or near 10 MPC for the most restrictive isotope expected to be present (Kr-85 or Xe-133).Back up for this monitor is provided by the radwaste building exhaust and effluent monitors.Indication of this monitor is provided on the AiRMS Visual Display in the control room.12.3.4.2.2.2.4Auxiliary Building Ventilation Exhaust Radioactivity Monitor The auxiliary building ventilation exhaust radioactivity monitor, 0-GL-RE-60, continuously monitors for particulate radioactivity in the auxiliary building ventilation system upstream of the filter-adsorber units. The sample point is located to monitor between the last point of possible radioactivity entry to the ventilation system from the areas served and the filter adsorber unit. The sample is extracted through an isokinetic nozzle, in accordance with ANSI Standard N13.1-1969, to ensure that a representative sample is provided to the fixed filter particulate detector assembly. Then the sample is discharged through the pumping system back to the duct. The cartridge filter will be removed for laboratory isotopic analysis upon activation of high activity alarm. The high alarm alerts the operator to high airborne particulate radioactivity levels in the auxiliary building. Indication of this monitor is provided on the AiRMS Visual Display in the control room.
CALLAWAY - SP12.3-30Rev. OL-2211/16Further determination of the source and the corrective action to be taken will be based on monitoring with portable detection and sampling equipment. Indication for this system is provided on the AiRMS Visual Display in the control room. 12.3.4.2.2.2.2Deleted 12.3.4.2.2.2.3Waste Gas Decay Tank Area Ventilation Exhaust Radioactivity MonitorThe waste gas decay tank area ventilation radioactivity monitor, 0-GH-RE-23, continuously monitors for gaseous radioactivity in the discharge duct from the waste gas decay tank area upstream of the radwaste building exhaust filter adsorber. The sample point provides rapid detection of a leak in the waste gas processing system and, in conjunction with the radwaste building exhaust radioactivity monitor and the radwaste building effluent monitor, helps localize the affected area in the event of an alarm on either monitor.The sample is extracted from the exhaust duct and passed through the fixed volume noble gas detector assembly and the pumping system. Then the sample is discharged back to the duct. The high alarm provides indication of a leak in the decay tanks, compressors, piping, or valves. The high-high alarm indicates that concentrations in the decay tank room are at or near 10 MPC for the most restrictive isotope expected to be present (Kr-85 or Xe-133).Back up for this monitor is provided by the radwaste building exhaust and effluent monitors.Indication of this monitor is provided on the AiRMS Visual Display in the control room.12.3.4.2.2.2.4Auxiliary Building Ventilation Exhaust Radioactivity Monitor The auxiliary building ventilation exhaust radioactivity monitor, 0-GL-RE-60, continuously monitors for particulate radioactivity in the auxiliary building ventilation system upstream of the filter-adsorber units. The sample point is located to monitor between the last point of possible radioactivity entry to the ventilation system from the areas served and the filter adsorber unit. The sample is extracted through an isokinetic nozzle, in accordance with ANSI Standard N13.1-1969, to ensure that a representative sample is provided to the fixed filter particulate detector assembly. Then the sample is discharged through the pumping system back to the duct. The cartridge filter will be removed for laboratory isotopic analysis upon activation of high activity alarm. The high alarm alerts the operator to high airborne particulate radioactivity levels in the auxiliary building. Indication of this monitor is provided on the AiRMS Visual Display in the control room.
CALLAWAY - SP12.3-31Rev. OL-2211/16If required, the portable monitor described in Section12.3.4.2.2.2.9 and/or grab samples will be utilized to determine airborne radioactivity levels and iodine concentrations in specific areas to aid in the determination of the source of the release. 12.3.4.2.2.2.5Containment Atmosphere Radioactivity MonitorsThe containment atmosphere radioactivity monitors, 0-GT-RE-31 and 0-GT-RE-32, continuously monitor the containment atmosphere for particulate, iodine, and gaseous radioactivity which could result in personnel exposure during periods of containment access. Other functions of these monitors are covered in Sections5.2.5
CALLAWAY - SP12.3-31Rev. OL-2211/16If required, the portable monitor described in Section12.3.4.2.2.2.9 and/or grab samples will be utilized to determine airborne radioactivity levels and iodine concentrations in specific areas to aid in the determination of the source of the release. 12.3.4.2.2.2.5Containment Atmosphere Radioactivity MonitorsThe containment atmosphere radioactivity monitors, 0-GT-RE-31 and 0-GT-RE-32, continuously monitor the containment atmosphere for particulate, iodine, and gaseous radioactivity which could result in personnel exposure during periods of containment access. Other functions of these monitors are covered in Sections5.2.5 , 7.3 , 9.4 , and 11.5.Samples are extracted from the operating deck level (El. 2047'-6") of the containment through the monitoring system sample lines.The monitors are located as close as possible to the containment penetrations to minimize the length of the sample tubing and the effects of sample plateout. The sample points are located in areas which ensure that representative samples are obtained. Each sample passes through the penetration and then through the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detector assemblies. After passing through the pumping system, the sample is discharged back to the containment through a separate penetration.Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and through isolated signals, on the AiRMS Visual Display in the control room. The containment atmosphere radioactivity monitors are seismic CategoryI systems and completely redundant. The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere. 12.3.4.2.2.2.6Containment Purge System Radioactivity MonitorsThe containment purge system radiation monitors, 0-GT-RE-22 and 0-GT-RE-33, continuously monitor the containment purge exhaust duct during normal purge operations for particulate, iodine, and gaseous radioactivity for worker protection as backup monitors for the containment atmosphere monitors. Other functions for the containment purge monitors are given in Sections5.2.5 , 7.3 , 9.4, and 11.5. The purge monitors are seismic CategoryI and completely redundant.
, 7.3, 9.4, and 11.5.Samples are extracted from the operating deck level (El. 2047'-6") of the containment through the monitoring system sample lines.The monitors are located as close as possible to the containment penetrations to minimize the length of the sample tubing and the effects of sample plateout. The sample points are located in areas which ensure that representative samples are obtained. Each sample passes through the penetration and then through the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detector assemblies. After passing through the pumping system, the sample is discharged back to the containment through a separate penetration.Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and through isolated signals, on the AiRMS Visual Display in the control room. The containment atmosphere radioactivity monitors are seismic CategoryI systems and completely redundant. The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere. 12.3.4.2.2.2.6Containment Purge System Radioactivity MonitorsThe containment purge system radiation monitors, 0-GT-RE-22 and 0-GT-RE-33, continuously monitor the containment purge exhaust duct during normal purge operations for particulate, iodine, and gaseous radioactivity for worker protection as backup monitors for the containment atmosphere monitors. Other functions for the containment purge monitors are given in Sections5.2.5
, 7.3, 9.4, and 11.5. The purge monitors are seismic CategoryI and completely redundant.
The sample points are located outside the containment between the containment isolation dampers and the containment purge filter adsorber unit.
The sample points are located outside the containment between the containment isolation dampers and the containment purge filter adsorber unit.
CALLAWAY - SP12.3-32Rev. OL-2211/16Each monitor is provided with two isokinetic nozzles to ensure that representative samples are obtained from both normal purge and minipurge. Isokinetic nozzle selection is accomplished by sample selector valves which automatically align the correct nozzle to the monitor, based on operation of the minipurge and normal purge exhaust fans. The sample is extracted through the selected nozzle and then passes through the selector valve, the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detectors. The sample then passes through the pumping system and is discharged back to the duct. Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS Visual Display in the control room. The containment purge radiation monitors provide backup for the containment atmosphere radiation monitors. The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere. For plant conditions during CORE ALTERATIONS and during movement of irradiated fuel within containment, the function of the monitors is to alarm only and the trip signals for automatic actuation of CPIS may be bypassed. One instrumentation channel at a minimum is required for the alarm only function during plant refueling activities.12.3.4.2.2.2.7Control Room Ventilation Radioactivity MonitorsThe control room ventilation radioactivity monitors, 0-GK-RE-04 and 0-GK-RE-05, continuously monitor the supply air of the normal heating, ventilation, and air-conditioning system for particulate, iodine, and gaseous radioactivity to provide protection for the control room operators in the event of high airborne radioactivity in the control room HVAC supply duct. This seismic CategoryI system is completely redundant. Samples are extracted through individual isokinetic nozzles, in accordance with ANSI Standard N13.1-1969, and flow through the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detector assemblies prior to passing through the pumping system for discharge to the auxiliary building atmosphere. The high and high-high alarms alert operators to high airborne radioactivity in the control room supply duct. The safety control functions are described in Sections6.4
CALLAWAY - SP12.3-32Rev. OL-2211/16Each monitor is provided with two isokinetic nozzles to ensure that representative samples are obtained from both normal purge and minipurge. Isokinetic nozzle selection is accomplished by sample selector valves which automatically align the correct nozzle to the monitor, based on operation of the minipurge and normal purge exhaust fans. The sample is extracted through the selected nozzle and then passes through the selector valve, the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detectors. The sample then passes through the pumping system and is discharged back to the duct. Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS Visual Display in the control room. The containment purge radiation monitors provide backup for the containment atmosphere radiation monitors. The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere. For plant conditions during CORE ALTERATIONS and during movement of irradiated fuel within containment, the function of the monitors is to alarm only and the trip signals for automatic actuation of CPIS may be bypassed. One instrumentation channel at a minimum is required for the alarm only function during plant refueling activities.12.3.4.2.2.2.7Control Room Ventilation Radioactivity MonitorsThe control room ventilation radioactivity monitors, 0-GK-RE-04 and 0-GK-RE-05, continuously monitor the supply air of the normal heating, ventilation, and air-conditioning system for particulate, iodine, and gaseous radioactivity to provide protection for the control room operators in the event of high airborne radioactivity in the control room HVAC supply duct. This seismic CategoryI system is completely redundant. Samples are extracted through individual isokinetic nozzles, in accordance with ANSI Standard N13.1-1969, and flow through the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detector assemblies prior to passing through the pumping system for discharge to the auxiliary building atmosphere. The high and high-high alarms alert operators to high airborne radioactivity in the control room supply duct. The safety control functions are described in Sections6.4 , 7.3, and 11.5. Indication for these monitors is provided on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS Visual Display in the control room.
, 7.3, and 11.5. Indication for these monitors is provided on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS Visual Display in the control room.
CALLAWAY - SP12.3-33Rev. OL-2211/1612.3.4.2.2.2.8Fuel Building Ventilation Exhaust Radioactivity MonitorsThe fuel building ventilation exhaust radioactivity monitors, 0-GG-RE-27 and 0-GG-RE-28, continuously monitor for particulate, iodine, and gaseous radioactivity in the fuel building ventilation exhaust system for the protection of the workers in the fuel building. The other functions for these monitors are described in Sections9.4 and 11.5. During normal operation, each of the monitors extracts a sample from the normal exhaust duct through individual isokinetic nozzles and sample selector valves. This normal sample point is upstream of the fuel building normal exhaust filter adsorber unit. The high and high-high alarms alert operators to high airborne particulate, iodine, or gaseous radioactivity in the fuel building. The ventilation control functions are described in Sections7.3 and 11.5. Indication is provided by individual indicators on the AiRMS control panel and through isolated signals by the AiRMS Visual Display in the control room. 12.3.4.2.2.2.9Portable MonitorPortable air radioactivity samplers are provided to allow periodic localized monitoring of specific air volumes of interest independent of the fixed monitor systems. The samplers are used to verify that airborne radioactivity levels within the plant operating spaces are within allowable limits and also to verify the proper operation of fixed monitor systems. A portable particulate, iodine, and gaseous continuous air monitor or a particulate only continuous air monitor will be used to monitor the local areas where there is a possibility of airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally, using the portable monitor. 12.3.4.2.2.3Safety EvaluationDue to their safety-related functions, discussed in other sections, the control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are redundant, independent, seismic CategoryI with Class 1E power supplies. These monitors all have safety-related control functions which are described in Section7.3. The following monitors are located upstream of filters and therefore, are effective for inplant personnel protection: a.Containment atmosphereb.Containment purge CALLAWAY - SP12.3-34Rev. OL-2211/16c.Control room supplyd.Fuel building exhaust e.Auxiliary building exhaustf.Access control area exhaustg.Waste gas decay tank room exhaust h.Portable monitorAll the inplant areas where the potential for airborne radioactivity exists are, therefore, monitored. The process and effluent radioactivity monitors are discussed in Section11.5. When the dilution factors of the ventilation system are large enough to preclude direct detection of halogens and noble gases at MPC levels, the release of these isotopes will be accounted for because the particulate activity, present along with halogens and noble gases, will be detected at much lower concentrations. Corresponding ratios among the various species of the airborne radioactivity are periodically determined by laboratory analysis so that the leakage of halogen and gaseous radioactivity can be accounted for. The waste gas decay tank room exhaust radioactivity monitor is designed to detect noble gas radioactivity, because this is the predominant type of activity present. Particulates and the iodine may be present in very small amounts, but will not be significant in this case since they will be removed by the waste gas processing system. A portable particulate, iodine, and gaseous monitor or a particulate only continuous air monitor is used to monitor the local areas where there is a possibility of airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally, using the portable monitor. On the basis of the above discussion, the AiRMS is adequate and sufficient to ensure personnel protection from airborne radioactivity. The system will provide indication to the operator that airborne radioactivity exists. The location of the airborne radioactivity can then be further identified by using portable air samplers that can be manually connected to various selected sample points in the building. Use of these sample points will direct radiation protection personnel to the particular area of concern. The exact location of the airborne radioactivity can then be found by sampling the particular subcompartments in that area. The combination of the AiRMS, in conjunction with administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns CALLAWAY - SP12.3-35Rev. OL-2211/16throughout the plant, plant equipment layout, and restricted radiation Zone E areas, is sufficient to ensure that airborne radioactivity levels are safe in terms of the required duration of personnel access throughout all areas of the plant. A general review of these concepts follows: a.Equipment location is such that hot piping and equipment are located in Radiation Zone D and E areas, which are restricted, and entry is limited by administrative control. Radiation Zone B and C areas do not contain piping and components that would result in significant airborne radioactivity sources. This reduces the possibility of airborne radioactivity exposure to occupants of Radiation Zone B and C areas where general entry is permitted.b.Air flow patterns are consistent with the basic ventilation design criteria of the plant. Clean filtered outside air is supplied to Zone B areas (corridors, clean areas); these areas are exhausted by drawing air into the rooms and areas of successively higher potential for airborne contamination. Air flow is such that air flow reversal or exfiltration from potentially contaminated areas is precluded. This ventilation arrangement eliminates the possibility of personnel exposure to airborne radioactivity in continuous occupancy areas, such as Radiation Zone B areas. c.High radiation areas (Radiation Zone E areas) where dose levels exceed 100 mrem/hr are restricted and conspicuously posted in accordance with 10 CFR 20.100-20.2401. These areas are not normally entered. d.Radiation protection programs are discussed in Section12.5. 12.
CALLAWAY - SP12.3-33Rev. OL-2211/1612.3.4.2.2.2.8Fuel Building Ventilation Exhaust Radioactivity MonitorsThe fuel building ventilation exhaust radioactivity monitors, 0-GG-RE-27 and 0-GG-RE-28, continuously monitor for particulate, iodine, and gaseous radioactivity in the fuel building ventilation exhaust system for the protection of the workers in the fuel building. The other functions for these monitors are described in Sections9.4 and 11.5. During normal operation, each of the monitors extracts a sample from the normal exhaust duct through individual isokinetic nozzles and sample selector valves. This normal sample point is upstream of the fuel building normal exhaust filter adsorber unit. The high and high-high alarms alert operators to high airborne particulate, iodine, or gaseous radioactivity in the fuel building. The ventilation control functions are described in Sections7.3 and 11.5. Indication is provided by individual indicators on the AiRMS control panel and through isolated signals by the AiRMS Visual Display in the control room. 12.3.4.2.2.2.9Portable MonitorPortable air radioactivity samplers are provided to allow periodic localized monitoring of specific air volumes of interest independent of the fixed monitor systems. The samplers are used to verify that airborne radioactivity levels within the plant operating spaces are within allowable limits and also to verify the proper operation of fixed monitor systems. A portable particulate, iodine, and gaseous continuous air monitor or a particulate only continuous air monitor will be used to monitor the local areas where there is a possibility of airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally, using the portable monitor. 12.3.4.2.2.3Safety EvaluationDue to their safety-related functions, discussed in other sections, the control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are redundant, independent, seismic CategoryI with Class 1E power supplies. These monitors all have safety-related control functions which are described in Section7.3
. The following monitors are located upstream of filters and therefore, are effective for inplant personnel protection: a.Containment atmosphereb.Containment purge CALLAWAY - SP12.3-34Rev. OL-2211/16c.Control room supplyd.Fuel building exhaust e.Auxiliary building exhaustf.Access control area exhaustg.Waste gas decay tank room exhaust h.Portable monitorAll the inplant areas where the potential for airborne radioactivity exists are, therefore, monitored. The process and effluent radioactivity monitors are discussed in Section11.5
. When the dilution factors of the ventilation system are large enough to preclude direct detection of halogens and noble gases at MPC levels, the release of these isotopes will be accounted for because the particulate activity, present along with halogens and noble gases, will be detected at much lower concentrations. Corresponding ratios among the various species of the airborne radioactivity are periodically determined by laboratory analysis so that the leakage of halogen and gaseous radioactivity can be accounted for. The waste gas decay tank room exhaust radioactivity monitor is designed to detect noble gas radioactivity, because this is the predominant type of activity present. Particulates and the iodine may be present in very small amounts, but will not be significant in this case since they will be removed by the waste gas processing system. A portable particulate, iodine, and gaseous monitor or a particulate only continuous air monitor is used to monitor the local areas where there is a possibility of airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally, using the portable monitor. On the basis of the above discussion, the AiRMS is adequate and sufficient to ensure personnel protection from airborne radioactivity. The system will provide indication to the operator that airborne radioactivity exists. The location of the airborne radioactivity can then be further identified by using portable air samplers that can be manually connected to various selected sample points in the building. Use of these sample points will direct radiation protection personnel to the particular area of concern. The exact location of the airborne radioactivity can then be found by sampling the particular subcompartments in that area. The combination of the AiRMS, in conjunction with administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns CALLAWAY - SP12.3-35Rev. OL-2211/16throughout the plant, plant equipment layout, and restricted radiation Zone E areas, is sufficient to ensure that airborne radioactivity levels are safe in terms of the required duration of personnel access throughout all areas of the plant. A general review of these concepts follows: a.Equipment location is such that hot piping and equipment are located in Radiation Zone D and E areas, which are restricted, and entry is limited by administrative control. Radiation Zone B and C areas do not contain piping and components that would result in significant airborne radioactivity sources. This reduces the possibility of airborne radioactivity exposure to occupants of Radiation Zone B and C areas where general entry is permitted.b.Air flow patterns are consistent with the basic ventilation design criteria of the plant. Clean filtered outside air is supplied to Zone B areas (corridors, clean areas); these areas are exhausted by drawing air into the rooms and areas of successively higher potential for airborne contamination. Air flow is such that air flow reversal or exfiltration from potentially contaminated areas is precluded. This ventilation arrangement eliminates the possibility of personnel exposure to airborne radioactivity in continuous occupancy areas, such as Radiation Zone B areas. c.High radiation areas (Radiation Zone E areas) where dose levels exceed 100 mrem/hr are restricted and conspicuously posted in accordance with 10 CFR 20.100-20.2401. These areas are not normally entered. d.Radiation protection programs are discussed in Section12.5
. 12.


==3.5REFERENCES==
==3.5REFERENCES==


1.R. L. Walker and M. Grotenhuis, ASummaryofShielding ConstantsforConcrete
1.R. L. Walker and M. Grotenhuis, ASummaryofShielding ConstantsforConcrete , ANL-6443 (November 1961).2.C. M. Lederer, et. al., TableofIsotopes, Lawrence Radiation Laboratory, University of California (March 1968).3.R. G. Jaeger, et. al., EngineeringCompendiumon RadiationShielding, Shielding Fundamentals and Methods, I (1968).4.G. W. Goldstein, X-RayAttenuationCoefficientsfrom 10keVto100MeV , National Bureau of Standards Circular 583 (issued April 10, 1957).5.ReactorPhysicsConstants, Argonne National Laboratory, ANL-5800 (July 1963).
, ANL-6443 (November 1961).2.C. M. Lederer, et. al., TableofIsotopes, Lawrence Radiation Laboratory, University of California (March 1968).3.R. G. Jaeger, et. al., EngineeringCompendiumon RadiationShielding, Shielding Fundamentals and Methods, I (1968).4.G. W. Goldstein, X-RayAttenuationCoefficientsfrom 10keVto100MeV
CALLAWAY - SP12.3-36Rev. OL-2211/166.J. F. Kircher and R. E. Bowman, EffectsofRadiation on Materials and Components (March 1964). 7.T. Rockwell, ReactorShieldingDesignManual, D. Van Nostrand Co., New York (1956). 8.C. R. Tipton, Jr., ReactorHandbook, Vol. I, Materials, second edition (1962). 9.H. Soodak, ReactorHandbook, Vol. III, Part A, Physics, second edition (1962). 10.E. P. Blizzard and L. S. Abbott, ReactorHandbook Vol.III, Part B, Shielding, second edition (1962).11.N. M. Schaeffer, ReactorShieldingforNuclearEngineers, TID-25951 (1967). 12.D. S. Duncan and A. B. Spear, GraceI-AnIBM704-709 Program Design for Computing GammaRayAttenuationandHeatinginReactorShields, Atomics International (June 1959). 13.D. S. Duncan and A. B. Spear, GraceII-AnIBM709 Programfor Computing Gamma RayAttenuationandHeatinginCylindricalandSphericalGeometries , Atomics International (November 1959). 14.W. W. Engle, Jr., AUser'sManualforANISN:AOne Dimensional Discrete Ordinates TransportCodewith AnisotropicScattering, Union Carbide Corporation, Report No. K-1693 (1967). 15.E. D. Arnold and B. F. Maskewitz, SDC,AShielding DesignCodefor Fuel Handling Facilities, ORNL-3041 (March 1966). 16.Richard E. Malenfant, "QAD,ASeriesofPoint-Kernel General-Purpose Shielding Programs," Los Alamos Scientific Laboratory, LA 3573 (October 1966).17.D. A. Klopp, NAP-MultigroupTime-DependentNeutron Activation Prediction Code, IITRI-A6088-21 (January 1966). 18.E. A. Straker, P. N. Stevens, D. C. Irving, and V. R. Crain, MORSE-AMultigroup Neutron andGamma-Ray MonteCarloTransportCode, ORNL-4585 (September 1970). 19.W. A. Rhoades and F. R. Mynatt, TheDOTIII Two-DimensionalDiscrete Ordinates TransportCode, ORNL-TM-4280 (1973). 20.R. E. Malefant, "G 3:AGeneralPurposeGamma-Ray ScatteringProgram," Los Alamos Scientific Laboratory, LA 5176 (June 1973).
, National Bureau of Standards Circular 583 (issued April 10, 1957).5.ReactorPhysicsConstants, Argonne National Laboratory, ANL-5800 (July 1963).
CALLAWAY - SP12.3-36Rev. OL-2211/166.J. F. Kircher and R. E. Bowman, EffectsofRadiation on Materials and Components (March 1964). 7.T. Rockwell, ReactorShieldingDesignManual, D. Van Nostrand Co., New York (1956). 8.C. R. Tipton, Jr., ReactorHandbook, Vol. I, Materials, second edition (1962). 9.H. Soodak, ReactorHandbook, Vol. III, Part A, Physics, second edition (1962). 10.E. P. Blizzard and L. S. Abbott, ReactorHandbook Vol.III, Part B, Shielding, second edition (1962).11.N. M. Schaeffer, ReactorShieldingforNuclearEngineers, TID-25951 (1967). 12.D. S. Duncan and A. B. Spear, GraceI-AnIBM704-709 Program Design for Computing GammaRayAttenuationandHeatinginReactorShields, Atomics International (June 1959). 13.D. S. Duncan and A. B. Spear, GraceII-AnIBM709 Programfor Computing Gamma RayAttenuationandHeatinginCylindricalandSphericalGeometries
, Atomics International (November 1959). 14.W. W. Engle, Jr., AUser'sManualforANISN:AOne Dimensional Discrete Ordinates TransportCodewith AnisotropicScattering, Union Carbide Corporation, Report No. K-1693 (1967). 15.E. D. Arnold and B. F. Maskewitz, SDC,AShielding DesignCodefor Fuel Handling Facilities, ORNL-3041 (March 1966). 16.Richard E. Malenfant, "QAD,ASeriesofPoint-Kernel General-Purpose Shielding Programs," Los Alamos Scientific Laboratory, LA 3573 (October 1966).17.D. A. Klopp, NAP-MultigroupTime-DependentNeutron Activation Prediction Code, IITRI-A6088-21 (January 1966). 18.E. A. Straker, P. N. Stevens, D. C. Irving, and V. R. Crain, MORSE-AMultigroup Neutron andGamma-Ray MonteCarloTransportCode, ORNL-4585 (September 1970). 19.W. A. Rhoades and F. R. Mynatt, TheDOTIII Two-DimensionalDiscrete Ordinates TransportCode, ORNL-TM-4280 (1973). 20.R. E. Malefant, "G 3:AGeneralPurposeGamma-Ray ScatteringProgram," Los Alamos Scientific Laboratory, LA 5176 (June 1973).
CALLAWAY - SP12.3-37Rev. OL-2211/1621.M. J. Bell, "ORIGEN-TheORNLGenerationand DepletionCode," Oak Ridge National Laboratory, ORNL-4628 (May 1973).
CALLAWAY - SP12.3-37Rev. OL-2211/1621.M. J. Bell, "ORIGEN-TheORNLGenerationand DepletionCode," Oak Ridge National Laboratory, ORNL-4628 (May 1973).
CALLAWAY - SPRev. OL-135/03TABLE 12.3-1  LIST OF COMPUTER CODES USED IN SHIELDING DESIGN CALCULATIONSGRACE IMultigroup, multiregion, gamma-ray attenuation code used to compute gamma heating and gamma dose rates in slab geometry (Ref. 12). GRACE IIMultigroup, multiregion, gamma-ray attenuation code used to compute the dose rate or heat generation rate for a spherical or a cylindrical source with slab or concentric shields (Ref. 13). ANISNMultigroup, multiregion code solving the Boltzman transport equation for neutrons or gamma-rays in one dimensional slab, cylindrical, or spherical geometry (Ref. 14). SDCMultigroup, multiregion, Kernal integration gamma-ray, shield design code which calculates dose rates for 13 geometry options (Ref. 15). QADMultigroup, multiregion, three-dimensional, point Kernal code which calculates fast neutron and gamma-ray dose and heat generation rates (Ref. 16). NAPDetermines activation emission source strengths as a function of neutron exposure and decay time (Ref. 17). MORSE-CGThree-dimensional Monte Carlo neutron and gamma ray general transport code (Ref. 18). DOT IIITwo-dimensional neutron, gamma ray, discrete ordinate, transport code (Ref. 19). ORIGENIsotope generation and depletion code which solves equations of radioactive growth and decay for isotopes of arbitrary coupling (Ref. 21). G3A general purpose gamma-ray scattering code (Ref. 20).
CALLAWAY - SPRev. OL-135/03TABLE 12.3-1  LIST OF COMPUTER CODES USED IN SHIELDING DESIGN CALCULATIONSGRACE IMultigroup, multiregion, gamma-ray attenuation code used to compute gamma heating and gamma dose rates in slab geometry (Ref. 12). GRACE IIMultigroup, multiregion, gamma-ray attenuation code used to compute the dose rate or heat generation rate for a spherical or a cylindrical source with slab or concentric shields (Ref. 13). ANISNMultigroup, multiregion code solving the Boltzman transport equation for neutrons or gamma-rays in one dimensional slab, cylindrical, or spherical geometry (Ref. 14). SDCMultigroup, multiregion, Kernal integration gamma-ray, shield design code which calculates dose rates for 13 geometry options (Ref. 15). QADMultigroup, multiregion, three-dimensional, point Kernal code which calculates fast neutron and gamma-ray dose and heat generation rates (Ref. 16). NAPDetermines activation emission source strengths as a function of neutron exposure and decay time (Ref. 17). MORSE-CGThree-dimensional Monte Carlo neutron and gamma ray general transport code (Ref. 18). DOT IIITwo-dimensional neutron, gamma ray, discrete ordinate, transport code (Ref. 19). ORIGENIsotope generation and depletion code which solves equations of radioactive growth and decay for isotopes of arbitrary coupling (Ref. 21). G 3A general purpose gamma-ray scattering code (Ref. 20).
CALLAWAY - SPRev. OL-1610/07TABLE 12.3-2  AREA RADIATION MONITORSInstrumentNumberLocationRadiation ZoneRange (mrem/hr)Hi Alarm (mrem/hr)Hi-Hi Alarm (mrem/hr)0-SD-RE-1Radwaste Building Corridor, BasementB0.1-1042.5 (1)(8)15 (2)(8)0-SD-RE-2Radwaste Building Corridor, BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-3Radwaste Building Corridor, BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-4Radwaste Building Corridor, Ground FloorB0.1-1042.5 (1)(8)15 (2)(8)0 SD-RE-5Radwaste Building Corridor, Ground FloorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-6Solid Radwaste AreaB0.1-1042.5 (1)(8)15 (2)(8)0-SD-RE-7Truck SpaceB0.1 - 10 415 (8)100 (8)0-SD-RE-8Sample LaboratoryB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-9RW Bldg Valve Room CorridorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-10RW Bldg Valve Room CorridorC0.1-10415 (1)(8)100 (2)(8)0-SD-RE-11RW Bldg HVAC Filter UnitB0.1-1042.5 (1)(8)15 (2)(8)0-SD-RE 12Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-13Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-14Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-15Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-16Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-17Pipe Tunnel & Personnel AccessB0.1-1042.5 (1)(8)15 (2)(8)0-SD-RE-18Aux Bldg Ground Floor CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-19Aux Bldg Ground Floor CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-20Aux Bldg Valve Room Corridor Ground FloorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-21Aux Bldg Valve Room Corridor Ground FloorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-22Aux Bldg Corridor Ground FloorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-23Aux Bldg Corridor Ground FloorB0.1-1042.5 (1)(8)15 (2)(8)0-SD-RE-24RC Sample RoomC10 - 10515 (1)(8)100 (1)(8)0-SD-RE-25Filter Unit Aux BldgA0.1 - 10 40.5 (1)(8)2.5 (2)(8)0-SD-RE-26RHR Heat Exchanger OutsideB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-27Ctmt Purge Filter UnitC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-28Personnel HatchB0.1-1042.5 (1)(8)15 (2)(8)0-SD-RE-29Laundry Decontamination FacilityB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-30Laundry Decontamination FacilityB0.1-1042.5 (1)(8)15 (2)(8)
CALLAWAY - SPRev. OL-1610/07TABLE 12.3-2  AREA RADIATION MONITORSInstrumentNumberLocationRadiation ZoneRange (mrem/hr)Hi Alarm (mrem/hr)Hi-Hi Alarm (mrem/hr)0-SD-RE-1Radwaste Building Corridor, BasementB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-2Radwaste Building Corridor, BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-3Radwaste Building Corridor, BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-4Radwaste Building Corridor, Ground FloorB0.1-10 42.5 (1)(8)15 (2)(8)0 SD-RE-5Radwaste Building Corridor, Ground FloorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-6Solid Radwaste AreaB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-7Truck SpaceB0.1 - 10 415 (8)100 (8)0-SD-RE-8Sample LaboratoryB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-9RW Bldg Valve Room CorridorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-10RW Bldg Valve Room CorridorC0.1-10 415 (1)(8)100 (2)(8)0-SD-RE-11RW Bldg HVAC Filter UnitB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE 12Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-13Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-14Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-15Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-16Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-17Pipe Tunnel & Personnel AccessB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-18Aux Bldg Ground Floor CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-19Aux Bldg Ground Floor CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-20Aux Bldg Valve Room Corridor Ground FloorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-21Aux Bldg Valve Room Corridor Ground FloorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-22Aux Bldg Corridor Ground FloorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-23Aux Bldg Corridor Ground FloorB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-24RC Sample RoomC10 - 10 515 (1)(8)100 (1)(8)0-SD-RE-25Filter Unit Aux BldgA0.1 - 10 40.5 (1)(8)2.5 (2)(8)0-SD-RE-26RHR Heat Exchanger OutsideB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-27Ctmt Purge Filter UnitC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-28Personnel HatchB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-29Laundry Decontamination FacilityB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-30Laundry Decontamination FacilityB0.1-10 42.5 (1)(8)15 (2)(8)
CALLAWAY - SPTABLE 12.3-2 (Sheet 2)Rev. OL-1610/070-SD-RE-31Hot LaboratoryB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-32Control Bldg CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-33Control RoomA0.1 - 10 40.5 (1)(8)2.5 (2)(8)0-SD-RE-34Cask Handling Area    B0.1 - 10 42.5 (1)15 (2)0-SD-RE-35New Fuel Storage Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-36New Fuel Storage Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-37Fuel Storage Pool Area  (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-38Fuel Storage Pool Area  (3)B0.1 -1042.5 (1)15 (2)0-SD-RE-39Seal Table AreaE0.1 - 10 41000 (4)(8)10000 (5)(8)0-SD-RE-40Personnel Access Hatch AreaE0.1 - 10 41000 (4)(8)10000 (5)(8)0-SD-RE-41Manipulator Bridge Crane (6)E0.1 -10430 (7)100 (7)0-SD-RE-42Containment BuildingE0.1 -1041000 (4)(8)10000 (5)(8)0-SD-RE-47Boron Concentration Monitoring System RoomE0.1 - 10 415 (1)(8)100 (2)(8)(1)High alarm set for maximum radiation level for the radiation zone for that area.
CALLAWAY - SPTABLE 12.3-2 (Sheet 2)Rev. OL-1610/070-SD-RE-31Hot LaboratoryB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-32Control Bldg CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-33Control RoomA0.1 - 10 40.5 (1)(8)2.5 (2)(8)0-SD-RE-34Cask Handling Area    B0.1 - 10 42.5 (1)15 (2)0-SD-RE-35New Fuel Storage Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-36New Fuel Storage Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-37Fuel Storage Pool Area  (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-38Fuel Storage Pool Area  (3)B0.1 -10 42.5 (1)15 (2)0-SD-RE-39Seal Table AreaE0.1 - 10 41000 (4)(8)10000 (5)(8)0-SD-RE-40Personnel Access Hatch AreaE0.1 - 10 41000 (4)(8)10000 (5)(8)0-SD-RE-41Manipulator Bridge Crane (6)E0.1 -10 430 (7)100 (7)0-SD-RE-42Containment BuildingE0.1 -10 41000 (4)(8)10000 (5)(8)0-SD-RE-47Boron Concentration Monitoring System RoomE0.1 - 10 415 (1)(8)100 (2)(8)(1)High alarm set for maximum radiation level for the radiation zone for that area.
(2)High-high alarm set for the maximum radiation level for the radiation zone above the one for that area.
(2)High-high alarm set for the maximum radiation level for the radiation zone above the one for that area.
(3)These monitors serve as criticality monitors.
(3)These monitors serve as criticality monitors.
Line 205: Line 156:
&#xb5;Ci/ccControlling IsotopeAlert (15) Alarm
&#xb5;Ci/ccControlling IsotopeAlert (15) Alarm


&#xb5;Ci/ccOGTRE31Particulate (3) 10-12 to 10-71 x 10-11Cs1371.0 x 10-9 (16) OGTRE32Iodine (4) 10-11 to 10-61 x 10-10I1311.0 x 10-8Containment atmosphere monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-1331.0 x 10-4OGTRE22Particulate (3) 10-12 to 10-71 x 10-11Cs1375 x 10-8OGTRE33Iodine (4) 10-11 to 10-61 x 10-10I1315 x 10-8Containment purge system monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-133(10)OGGRE27Particulate (3) 10-12 to 10-71 x 10-11Cs1371 x 10-8OGGRE28Iodine (4) 10-11 to 10-61 x 10-10I1319 x 10-9Fuel building exhaust monitors (2)Gaseous (3) 10-7 to 10-22 x 10-7Xe-1331.6 x 10-3OGKRE04Particulate (3) 10-12 to 10-71 x 10-11Cs1371 x 10-8OGKRE05Iodine (4) 10-11 to 10-61 x 10-10I1319 x 10-9Control room air supply monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-1331.1 x 10-3OGLRE60Particulate (3) 10-12 to 10-71 x 10-11Cs1371 x 10-8Auxiliary Building ventilation exhaust monitorOGKRE41Particulate (3) 10-12 to 10-71 x 10-11Cs1371 x 10-9(8)Access control area ventilation exhaust monitorOGHRE23Gaseous (3) 10-7 to 10-22 x 10-7Kr85Xe-1331 x 10-5(8)Waste gas decay tank area ventilation exhaust monitorPortable monitorParticulate (3) 10-12 to 10-71 x 10-11Cs137NAIodine (4) 10-11 to 10-61 x 10-10I131NAGaseous (3) 10-7 to 10-22 x 10-7Kr85NA CALLAWAY - SPTABLE 12.3-3 (Sheet 2)Rev. OL-174/09IHigh (15)Alarm&#xb5;Ci/ccFlowVentilationFlow (cfm)SubcompartmentFlow Rate(cfm)DilutionFactorMinimum RequiredSensitivity  
&#xb5;Ci/ccOGTRE31Particulate (3) 10-12 to 10-71 x 10-11Cs1371.0 x 10-9 (16) OGTRE32Iodine (4) 10-11 to 10-61 x 10-10I1311.0 x 10-8Containment atmosphere monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-1331.0 x 10-4OGTRE22Particulate (3) 10-12 to 10-71 x 10-11Cs137 5 x 10-8OGTRE33Iodine (4) 10-11 to 10-61 x 10-10I131 5 x 10-8Containment purge system monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-133(10)OGGRE27Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-8OGGRE28Iodine (4) 10-11 to 10-61 x 10-10I131 9 x 10-9Fuel building exhaust monitors (2)Gaseous (3) 10-7 to 10-22 x 10-7Xe-1331.6 x 10-3OGKRE04Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-8OGKRE05Iodine (4) 10-11 to 10-61 x 10-10I131 9 x 10-9Control room air supply monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-1331.1 x 10-3OGLRE60Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-8Auxiliary Building ventilation exhaust monitorOGKRE41Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-9(8)Access control area ventilation exhaust monitorOGHRE23Gaseous (3) 10-7 to 10-22 x 10-7Kr85Xe-133 1 x 10-5(8)Waste gas decay tank area ventilation exhaust monitorPortable monitorParticulate (3) 10-12 to 10-71 x 10-11Cs137NAIodine (4) 10-11 to 10-61 x 10-10I131NAGaseous (3) 10-7 to 10-22 x 10-7Kr85NA CALLAWAY - SPTABLE 12.3-3 (Sheet 2)Rev. OL-174/09 IHigh (15)Alarm&#xb5;Ci/ccFlowVentilationFlow (cfm)SubcompartmentFlow Rate(cfm)DilutionFactorMinimum RequiredSensitivity  
(&#xb5;Ci/cc)Monitor Control Function1 x 10-7420,000NANA1 x 10-7(6)(6)NA9 x 10-7420,000NANA9 x 10-8(6)(6)NA6.0 x 10-4420,000NANA1 x 10-4(6)(6)NA1 x 10-720,000/4,000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and control functions.9 x 10-820,000/4,000NANA9 x 10-8(6)(6)(11) (14)20,000/4,000NANA1 x 10-4(6)(6)1 x 10-720,000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and  control functions.9 x 10-820,000NANA9 x 10-8(6)(6)3.2 x 10-3(13)20,000NANA1 x 10-4(6)(6)1 x 10-72000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and  control functions.9 x 10-72000NANA9 x 10-8(6)(6)2.2 x 10-3(12)2000NANA1 x 10-4(6)(6)1 x 10-712,000100 8 x 10-3(5)8 x 10-10(6),(9)Alarms1 x 10-8(7)6,0001001.67 x 10
(&#xb5;Ci/cc)Monitor Control Function1 x 10-7420,000NANA1 x 10-7(6)(6)NA9 x 10-7420,000NANA9 x 10-8(6)(6)NA6.0 x 10-4420,000NANA1 x 10-4(6)(6)NA1 x 10-720,000/4,000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and control functions.9 x 10-820,000/4,000NANA9 x 10-8(6)(6)(11) (14)20,000/4,000NANA1 x 10-4(6)(6)1 x 10-720,000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and  control functions.9 x 10-820,000NANA9 x 10-8(6)(6)3.2 x 10-3(13)20,000NANA1 x 10-4(6)(6)1 x 10-72000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and  control functions.9 x 10-72000NANA9 x 10-8(6)(6)2.2 x 10-3(12)2000NANA1 x 10-4(6)(6)1 x 10-712,000100 8 x 10-3(5)8 x 10-10(6),(9)Alarms1 x 10-8(7)6,0001001.67 x 10-2(5)1.67 x 10-9(6),(9)Alarms1 x 10-4(7)5002500.5(5)5 x 10-5(6),(9)AlarmsNANANANAAlarmsNANANANANANANANASample Flow for each channel is 3 cfm.
-2(5)1.67 x 10
-9(6),(9)Alarms1 x 10-4(7)5002500.5(5)5 x 10-5(6),(9)AlarmsNANANANAAlarmsNANANANANANANANASample Flow for each channel is 3 cfm.
(1)MDC = minimum detectable concentration.(2)When fuel is in the building.(3)Beta scintillation detector.
(1)MDC = minimum detectable concentration.(2)When fuel is in the building.(3)Beta scintillation detector.
(4)Gamma scintillation detector.(5)Dilution factor = (6)Minimum required sensitivity of monitor in  
(4)Gamma scintillation detector.(5)Dilution factor = (6)Minimum required sensitivity of monitor in  
Line 216: Line 165:
(14)High alarm setpoint is established to ensure that ODCM limits are not exceeded.(15)Alert and High alarm setpoint values do not include instrument loop uncertainty estimates.(16)Alert alarm value is set to meet the criteria of Note 10 and to meet RCS leakage detection requirements described in FSAR Section 5.2.5.2.3
(14)High alarm setpoint is established to ensure that ODCM limits are not exceeded.(15)Alert and High alarm setpoint values do not include instrument loop uncertainty estimates.(16)Alert alarm value is set to meet the criteria of Note 10 and to meet RCS leakage detection requirements described in FSAR Section 5.2.5.2.3
.
.
CALLAWAY - SPRev. OL-195/12TABLE 12.3-4  POWER SUPPLIES FOR AREA AND IN-PLANT AIRBORNE MONITORSArea Radiation MonitorsMonitorNumberNormalPowerSupplyRestored AfterLoss of OffsitePowerRemarks0-SD-RE-1 0-SD-RE-20-SD-RE-30-SD-RE-4 0-SD-RE-50-SD-RE-60-SD-RE-7 0-SD-RE-80-SD-RE-90-SD-RE-10 0-SD-RE-110-SD-RE-120-SD-RE-13 0-SD-RE-140-SD-RE-150-SD-RE-16 0-SD-RE-170-SD-RE-180-SD-RE-19 0-SD-RE-200-SD-RE-210-SD-RE-22 0-SD-RE-230-SD-RE-240-SD-RE-25 0-SD-RE-260-SD-RE-270-SD-RE-28 0-SD-RE-290-SD-RE-300-SD-RE-31 0-SD-RE-320-SD-RE-330-SD-RE-34 0-SD-RE-350-SD-RE-36Supplied by regulated  instrumentation ac power which is supplied from the diesel generators on loss of offsite power.Yes CALLAWAY - SPTABLE 12.3-4 (Sheet 2)Rev. OL-195/120-SD-RE-370-SD-RE-38 0-SD-RE-390-SD-RE-400-SD-RE-41 0-SD-RE-420-SD-RE-47Containmentatmosphere0-GT-RE-31 0-GT-RE-32Class 1E MCCsYesContainmentpurge system0-GT-RE-220-GT-RE-33Class 1E MCCsYesFuel buildingexhaust 0-GG-RE-270-GG-RE-28Class 1E MCCsYes  Control room air supply0-GK-RE-04 0-GK-RE-05Class 1E MCCsYes  In-Plant Airborne Radioactivity Monitors (Non-1E)Auxiliary building ventilation exhaust 0-GL-RE-60Non-1E MCCsNoPower is lost to system also, so monitor reading is not meaningful.Access control area ventilation exhaust0-GK-RE-41Non-1E MCCSNoPower is lost to system also, so monitor reading is not meaningful.Waste gas decay tank area  ventilation exhaust 0-GH-RE-23Non-1E MCCsNoPower is lost to system also, so monitor reading is not meaningful.MonitorNumberNormalPowerSupplyRestored AfterLoss of OffsitePowerRemarks CALLAWAY - SP12.4-1Rev. OL-2211/1612.4DOSEASSESSMENTRadiation exposures in the plant are primarily due to direct radiation from components and equipment containing radioactive fluids. In addition, in some plant radiation areas there can be radiation exposure to personnel due to the presence of airborne radionuclides. In-plant radiation exposures during normal operation and anticipated operational occurrences are discussed in Section 12.4.1. Radiation exposures due to direct radiation at locations outside the plant structures, such as the boundary of the restricted area, are a function of the plant layout, equipment selection, and detailed system and shielding designs and are expected to be negligible. Radiation exposures due to the airborne radioactive effluent plume at these locations are expected to be insignificant. The radiation exposures at these locations are discussed in Section 12.4.2.2.Radiation exposures to operating personnel will be within 10CFR 20 limits. Radiation protection design features described in Section 12.3 and the radiation protection program outlined in Section 12.5 will assure that the occupational radiation exposures (ORE) to operating personnel during operation and anticipated operational concurrences will be as low as is reasonably achievable (ALARA).12.4.1EXPOSURES WITHIN THE PLANT 12.4.1.1DirectRadiationDoseEstimatesAnnual man-rem doses from direct radiation during the performance of routine functions, such as operation and surveillance, normal maintenance, radwaste handling, refueling, and inservice inspection, have been estimated, using the following bases:a.Radiation exposure data from operating PWRs are given in Tables 12.4-1 through 12.4-11 (Ref. 1, 2, 3) and Figure 12.4-1
CALLAWAY - SPRev. OL-195/12TABLE 12.3-4  POWER SUPPLIES FOR AREA AND IN-PLANT AIRBORNE MONITORSArea Radiation MonitorsMonitorNumberNormalPowerSupplyRestored AfterLoss of OffsitePowerRemarks0-SD-RE-1 0-SD-RE-20-SD-RE-30-SD-RE-4 0-SD-RE-50-SD-RE-60-SD-RE-7 0-SD-RE-80-SD-RE-90-SD-RE-10 0-SD-RE-110-SD-RE-120-SD-RE-13 0-SD-RE-140-SD-RE-150-SD-RE-16 0-SD-RE-170-SD-RE-180-SD-RE-19 0-SD-RE-200-SD-RE-210-SD-RE-22 0-SD-RE-230-SD-RE-240-SD-RE-25 0-SD-RE-260-SD-RE-270-SD-RE-28 0-SD-RE-290-SD-RE-300-SD-RE-31 0-SD-RE-320-SD-RE-330-SD-RE-34 0-SD-RE-350-SD-RE-36Supplied by regulated  instrumentation ac power which is supplied from the diesel generators on loss of offsite power.Yes CALLAWAY - SPTABLE 12.3-4 (Sheet 2)Rev. OL-195/120-SD-RE-370-SD-RE-38 0-SD-RE-390-SD-RE-400-SD-RE-41 0-SD-RE-420-SD-RE-47Containmentatmosphere0-GT-RE-31 0-GT-RE-32Class 1E MCCsYesContainmentpurge system0-GT-RE-220-GT-RE-33Class 1E MCCsYesFuel buildingexhaust 0-GG-RE-270-GG-RE-28Class 1E MCCsYes  Control room air supply0-GK-RE-04 0-GK-RE-05Class 1E MCCsYes  In-Plant Airborne Radioactivity Monitors (Non-1E)Auxiliary building ventilation exhaust 0-GL-RE-60Non-1E MCCsNoPower is lost to system also, so monitor reading is not meaningful.Access control area ventilation exhaust0-GK-RE-41Non-1E MCCSNoPower is lost to system also, so monitor reading is not meaningful.Waste gas decay tank area  ventilation exhaust 0-GH-RE-23Non-1E MCCsNoPower is lost to system also, so monitor reading is not meaningful.MonitorNumberNormalPowerSupplyRestored AfterLoss of OffsitePowerRemarks CALLAWAY - SP12.4-1Rev. OL-2211/1612.4DOSEASSESSMENTRadiation exposures in the plant are primarily due to direct radiation from components and equipment containing radioactive fluids. In addition, in some plant radiation areas there can be radiation exposure to personnel due to the presence of airborne radionuclides. In-plant radiation exposures during normal operation and anticipated operational occurrences are discussed in Section 12.4.1. Radiation exposures due to direct radiation at locations outside the plant structures, such as the boundary of the restricted area, are a function of the plant layout, equipment selection, and detailed system and shielding designs and are expected to be negligible. Radiation exposures due to the airborne radioactive effluent plume at these locations are expected to be insignificant. The radiation exposures at these locations are discussed in Section 12.4.2.2.Radiation exposures to operating personnel will be within 10CFR 20 limits. Radiation protection design features described in Section 12.3 and the radiation protection program outlined in Section 12.5 will assure that the occupational radiation exposures (ORE) to operating personnel during operation and anticipated operational concurrences will be as low as is reasonably achievable (ALARA).12.4.1EXPOSURES WITHIN THE PLANT 12.4.1.1DirectRadiationDoseEstimatesAnnual man-rem doses from direct radiation during the performance of routine functions, such as operation and surveillance, normal maintenance, radwaste handling, refueling, and inservice inspection, have been estimated, using the following bases:a.Radiation exposure data from operating PWRs are given in Tables 12.4-1 through 12.4-11 (Ref. 1, 2, 3) and Figure 12.4-1. b.Expected average dose rates in plant radiation areas are discussed in this section.c.Expected occupancy times for various work function personnel in the different plant radiation areas are listed in Table 12.4-12
. b.Expected average dose rates in plant radiation areas are discussed in this section.c.Expected occupancy times for various work function personnel in the different plant radiation areas are listed in Table 12.4-12
.d.Anticipated manpower requirements for each unit are given in Table 12.4-12.e.Table 12.4-12 provides an estimate of the distribution of the annual man-rem according to work function.
.d.Anticipated manpower requirements for each unit are given in Table 12.4-12.e.Table 12.4-12 provides an estimate of the distribution of the annual man-rem according to work function.
CALLAWAY - SP12.4-2Rev. OL-2211/16The maximum and expected average dose rates in the plant radiation areas are given below:The maximum dose rates are determined by shielding calculations based on conservative assumptions regarding sources, self-shielding, locations, etc. The expected dose rates are estimated by assuming a failed fuel percentage of 0.12 and that stringent water chemistry control and improved design will minimize crud buildup and hence the expected dose rates in various radiation zones and recognition that real maximum doses in a given zone are localized effects. The expected average doses given above are used in computing the doses for personnel involved in all operations, except inservice inspection and special maintenance. For personnel involved in the performance of inservice inspection (ISI) and special maintenance tasks, an expected average dose rate of 200 mR/hr in the E Zone is used since these personnel will generally be working on reactor coolant system components.Direct radiation exposures to plant personnel can result from the performance of special maintenance functions. In view of the radiation protection design features described in Section12.3 and the radiation protection program outlined in Section 12.5, it is expected that exposures due to special maintenance will be minimized. However, an annual exposure of 150 man-rem/unit is realistically estimated, based on experience at operating PWRs.Exposure to plant personnel from direct radiation during the performance of routine functions is estimated to be approximately 220 man-rem/yr/unit. Details of the man-rem estimates are given in Table 12.4-12. A breakdown of the exposures (including the special maintenance category) by work functions is provided in Table 12.4-12
CALLAWAY - SP12.4-2Rev. OL-2211/16The maximum and expected average dose rates in the plant radiation areas are given below:The maximum dose rates are determined by shielding calculations based on conservative assumptions regarding sources, self-shielding, locations, etc. The expected dose rates are estimated by assuming a failed fuel percentage of 0.12 and that stringent water chemistry control and improved design will minimize crud buildup and hence the expected dose rates in various radiation zones and recognition that real maximum doses in a given zone are localized effects. The expected average doses given above are used in computing the doses for personnel involved in all operations, except inservice inspection and special maintenance. For personnel involved in the performance of inservice inspection (ISI) and special maintenance tasks, an expected average dose rate of 200 mR/hr in the E Zone is used since these personnel will generally be working on reactor coolant system components.Direct radiation exposures to plant personnel can result from the performance of special maintenance functions. In view of the radiation protection design features described in Section12.3 and the radiation protection program outlined in Section 12.5, it is expected that exposures due to special maintenance will be minimized. However, an annual exposure of 150 man-rem/unit is realistically estimated, based on experience at operating PWRs.Exposure to plant personnel from direct radiation during the performance of routine functions is estimated to be approximately 220 man-rem/yr/unit. Details of the man-rem estimates are given in Table 12.4-12. A breakdown of the exposures (including the special maintenance category) by work functions is provided in Table 12.4-12. Table 12.4-13 provides the percentage of the annual total man-rem associated with each work function.Total occupancy in various radiation zones:  Total occupancy time for personnel involved in different work functions in various radiation zones is as follows:ZoneMaximum Dose Rate Expected Average Dose Rate (mrem/hr)(mrem/hr)A        0.5    0.1 B        2.5    0.5C    15    2.5 D  100 15E>100100 CALLAWAY - SP12.4-3Rev. OL-2211/16a.Routine operation and surveillance - The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 2,080 hrs/yr. The major portion of the occupancy is expected to be in radiation Zones A and B. Combined occupancy in radiation Zones C, D, and E is expected to be approximately 4 percent of the total annual occupancy. The average annual dose for an individual involved in this work category is expected to be about 1 rem. The distribution of occupany times in various radiation zones is listed in Table 12.4-12.b.Routine maintenance:  The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 2,080 hrs/yr.
. Table 12.4-13 provides the percentage of the annual total man-rem associated with each work function.Total occupancy in various radiation zones:  Total occupancy time for personnel involved in different work functions in various radiation zones is as follows:ZoneMaximum Dose Rate Expected Average Dose Rate (mrem/hr)(mrem/hr)A        0.5    0.1 B        2.5    0.5C    15    2.5 D  100 15E>100100 CALLAWAY - SP12.4-3Rev. OL-2211/16a.Routine operation and surveillance - The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 2,080 hrs/yr. The major portion of the occupancy is expected to be in radiation Zones A and B. Combined occupancy in radiation Zones C, D, and E is expected to be approximately 4 percent of the total annual occupancy. The average annual dose for an individual involved in this work category is expected to be about 1 rem. The distribution of occupany times in various radiation zones is listed in Table 12.4-12.b.Routine maintenance:  The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 2,080 hrs/yr.
Personnel involved in this work category are expected to spend more time in high radiation areas. However, by following maintenance procedures, such as flushing equipment in high radiation areas before performing maintenance and also by removing smaller equipment from high radiation areas to lower radiation areas for maintenance, the dose rate for personnel can be minimized. Consequently, the annual average dose rate for an individual involved in this work function is expected to be about 3.7 rem. The distribution of occupancy time in various radiation zones is listed in Table 12.4-12
Personnel involved in this work category are expected to spend more time in high radiation areas. However, by following maintenance procedures, such as flushing equipment in high radiation areas before performing maintenance and also by removing smaller equipment from high radiation areas to lower radiation areas for maintenance, the dose rate for personnel can be minimized. Consequently, the annual average dose rate for an individual involved in this work function is expected to be about 3.7 rem. The distribution of occupancy time in various radiation zones is listed in Table 12.4-12
.c.Inservice inspection:  The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 240 hrs/yr  
.c.Inservice inspection:  The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 240 hrs/yr  
Line 226: Line 173:
(Table 12.4-12) at the rate of 40 hrs/wk for 8 wks/yr. The annual average dose for an individual involved in this work function is approximately 0.5 rem.d.Refueling:  The refueling work is performed by personnel involved in routine operation and surveillance, radiation protection and rad chemistry, and routine maintenance. The expected occupancy time for personnel involved in the refueling operation is about 160hrs/yr. The average dose per individual in this category is estimated to be 0.5 rem over the occupancy period. The personnel involved in the refueling operation are expected to spend more time in the high radiation areas. The initial crud burst in the refueling pool and fuel storage pool regions will initially result in occupancies in radiation Zone C. However, the operation of the cleanup systems will result in significant reduction of the dose rates in the regions where occupancy is expected. In view of the above considerations, a combined occupancy of 70 percent in radiation Zones B, C, D, and E is considered reasonable. A realistic distribution of personnel occupancies in various radiation zones is provided in Table 12.4-12
(Table 12.4-12) at the rate of 40 hrs/wk for 8 wks/yr. The annual average dose for an individual involved in this work function is approximately 0.5 rem.d.Refueling:  The refueling work is performed by personnel involved in routine operation and surveillance, radiation protection and rad chemistry, and routine maintenance. The expected occupancy time for personnel involved in the refueling operation is about 160hrs/yr. The average dose per individual in this category is estimated to be 0.5 rem over the occupancy period. The personnel involved in the refueling operation are expected to spend more time in the high radiation areas. The initial crud burst in the refueling pool and fuel storage pool regions will initially result in occupancies in radiation Zone C. However, the operation of the cleanup systems will result in significant reduction of the dose rates in the regions where occupancy is expected. In view of the above considerations, a combined occupancy of 70 percent in radiation Zones B, C, D, and E is considered reasonable. A realistic distribution of personnel occupancies in various radiation zones is provided in Table 12.4-12
.e.Special maintenance:  The total occupancy time for personnel involved in special maintenance problems, generally unanticipated, in various radiation zones is expected to be 320 hrs/yr, and the dose rate per CALLAWAY - SP12.4-4Rev. OL-2211/16individual involved in this work function is estimated to be about 0.7 rem/yr. The distribution of occupancy time in various radiation zones is provided in Table 12.4-12
.e.Special maintenance:  The total occupancy time for personnel involved in special maintenance problems, generally unanticipated, in various radiation zones is expected to be 320 hrs/yr, and the dose rate per CALLAWAY - SP12.4-4Rev. OL-2211/16individual involved in this work function is estimated to be about 0.7 rem/yr. The distribution of occupancy time in various radiation zones is provided in Table 12.4-12
.f.Radwaste processing:  The radwaste operations will be performed by personnel involved in work functions such as routine operation and surveillance, radiation protection, rad chemistry, and routine maintenance. The total times expended by personnel in various radiation levels for the category are given in Table 12.4-12. The expected occupancy time for personnel involved in radwaste processing is about 2080 hrs/yr. The annual average dose per individual in this category is estimated to be about 500 mrem.12.4.1.2AirborneRadioactivityDoseEstimates12.4.1.2.1Exposures Due to Airborne Radioactivity As already discussed in Section 12.2.2, negligible airborne concentrations and consequently negligible airborne radio-activity exposures are expected in those areas of the auxiliary, radwaste, and turbine buildings which are accessible (Ref.4). Exposures due to airborne radioactivity are possible in the containment and fuel building both during power operation and refueling. However, the design of the plant operating procedures described in Sections 12.3.3
.f.Radwaste processing:  The radwaste operations will be performed by personnel involved in work functions such as routine operation and surveillance, radiation protection, rad chemistry, and routine maintenance. The total times expended by personnel in various radiation levels for the category are given in Table 12.4-12. The expected occupancy time for personnel involved in radwaste processing is about 2080 hrs/yr. The annual average dose per individual in this category is estimated to be about 500 mrem.12.4.1.2AirborneRadioactivityDoseEstimates12.4.1.2.1Exposures Due to Airborne Radioactivity As already discussed in Section 12.2.2, negligible airborne concentrations and consequently negligible airborne radio-activity exposures are expected in those areas of the auxiliary, radwaste, and turbine buildings which are accessible (Ref.4). Exposures due to airborne radioactivity are possible in the containment and fuel building both during power operation and refueling. However, the design of the plant operating procedures described in Sections 12.3.3 , 12.3.4, and 12.5.3, and expected limited occupancies in these buildings will minimize exposures to airborne activity and ensure that the doses to an individual from airborne radioactivity are small fractions of the 10 CFR 20 limits for occupational workers and that annual man-rem exposures comply with the ALARA criteria within the plant. The annual man-rem exposures from airborne radioactivity will be a small fraction of the annual man-rem exposures due to direct radiation. Annual occupancy (man-hours), dose rates, and man-rem due to airborne radioactivity in these areas are given in Tables 12.4-14 and 12.4-15, respectively.12.4.1.2.2Model for Calculating Exposures Due to Airborne Radioactive SourcesThyroid and inhalation doses via inhalation pathway are calculated using the following equation:where C i=Airborne concentration of the i th nuclide in pci/cm 3 D oi C i BR ()tDF oi=
, 12.3.4, and 12.5.3, and expected limited occupancies in these buildings will minimize exposures to airborne activity and ensure that the doses to an individual from airborne radioactivity are small fractions of the 10 CFR 20 limits for occupational workers and that annual man-rem exposures comply with the ALARA criteria within the plant. The annual man-rem exposures from airborne radioactivity will be a small fraction of the annual man-rem exposures due to direct radiation. Annual occupancy (man-hours), dose rates, and man-rem due to airborne radioactivity in these areas are given in Tables 12.4-14 and 12.4-15, respectively.12.4.1.2.2Model for Calculating Exposures Due to Airborne Radioactive SourcesThyroid and inhalation doses via inhalation pathway are calculated using the following equation:where Ci=Airborne concentration of the i th nuclide in pci/cm 3DoiCiBR()tDFoi=
CALLAWAY - SP12.4-5Rev. OL-2211/16andTotal body submersion doses are calculated, using a finite cloud model.Annual man-rem exposures due to airborne radioactivity are calculated using the following equation:where and 12.4.1.3IllustrativeExamplesofDoseAssessmentDose assessments for various operations were based on actual operating plant data. A number of typical examples are provided in Table 12.4-1. Note that the maximum/minimum values for number of personnel do not correspond to the maximum/minimum number of days required or dose rates so that the man-rem totals are not simple multiplications of the maximum/minimum factors. The dose assessments in Table 12.4-1 were derived from the average number of personnel, average length of time, and average dose rate to perform each particular operation.a.Operation and SurveillanceThe dose rates in the corridors and other normally occupied areas are expected to be much lower than the maximum radiation levels for each BR =Breathing rate for occupational worker in cm 3/sec=347t=Time duration of inhalation of radioactivity contaminated air in seconds DF oi =Dose factor for adult for organ O (thyroid or lung) via inhalation in mrem/pCi inhaled for the i th isotope (These dose factors are taken from Regulatory Guide 1.109, Rev. 1)
CALLAWAY - SP12.4-5Rev. OL-2211/16andTotal body submersion doses are calculated, using a finite cloud model.Annual man-rem exposures due to airborne radioactivity are calculated using the following equation:where and 12.4.1.3IllustrativeExamplesofDoseAssessmentDose assessments for various operations were based on actual operating plant data. A number of typical examples are provided in Table 12.4-1. Note that the maximum/minimum values for number of personnel do not correspond to the maximum/minimum number of days required or dose rates so that the man-rem totals are not simple multiplications of the maximum/minimum factors. The dose assessments in Table 12.4-1 were derived from the average number of personnel, average length of time, and average dose rate to perform each particular operation.a.Operation and SurveillanceThe dose rates in the corridors and other normally occupied areas are expected to be much lower than the maximum radiation levels for each BR =Breathing rate for occupational worker in cm 3/sec=347t=Time duration of inhalation of radioactivity contaminated air in secondsDFoi =Dose factor for adult for organ O (thyroid or lung) via inhalation in mrem/pCi inhaled for the i th isotope (These dose factors are taken from Regulatory Guide 1.109, Rev. 1)
D o=Dose in millirems to organ O due to inhalation(DR)o=Dose rate for organ in mrem/hrh=Annual occupancy in man-hours/yr D o=Annual exposure in man-rem/yr D o DR ()o 10 3-h=
Do=Dose in millirems to organ O due to inhalation(DR)o=Dose rate for organ in mrem/hrh=Annual occupancy in man-hours/yr Do=Annual exposure in man-rem/yr DoDR()o103-h=
CALLAWAY - SP12.4-6Rev. OL-2211/16zone shown in the radiation zone drawings (Figure 12.3-2). The expected radiation levels are provided in Section 12.4.1.1. Based on these expected dose rates and typical time periods for operation and surveillance, exposures were calculated. Typical examples from operating plants are given in Table 12.4-1. The total exposures for this category range from 13-30 man-rem.b.Routine MaintenanceA number of examples of normal man-rem associated with maintenance are provided in Table 12.4-1. The total number of annual man-rems of exposure associated with this category depend upon many variables, such as equipment run times and breakdowns, number of skilled personnel available, schedule, and crud trapping. Typically, routine maintenance can account for a large percentage of the annual man-rem.c.Radwaste ProcessingAnnual exposures for radwaste processing were determined as a result of the system evaluations described in Section 12.1.2.4. Average expected dose rates and personnel stay times were used. Typical data from operating plants for various tasks are presented in Table 12.4-1
CALLAWAY - SP12.4-6Rev. OL-2211/16zone shown in the radiation zone drawings (Figure 12.3-2). The expected radiation levels are provided in Section 12.4.1.1. Based on these expected dose rates and typical time periods for operation and surveillance, exposures were calculated. Typical examples from operating plants are given in Table 12.4-1. The total exposures for this category range from 13-30 man-rem.b.Routine MaintenanceA number of examples of normal man-rem associated with maintenance are provided in Table 12.4-1. The total number of annual man-rems of exposure associated with this category depend upon many variables, such as equipment run times and breakdowns, number of skilled personnel available, schedule, and crud trapping. Typically, routine maintenance can account for a large percentage of the annual man-rem.c.Radwaste ProcessingAnnual exposures for radwaste processing were determined as a result of the system evaluations described in Section 12.1.2.4. Average expected dose rates and personnel stay times were used. Typical data from operating plants for various tasks are presented in Table 12.4-1
.d.RefuelingBased on operating plant data, refueling operations have required 15 to 30 days, using 8 to 61 personnel who receive from 32 to 347 mrem/day.
.d.RefuelingBased on operating plant data, refueling operations have required 15 to 30 days, using 8 to 61 personnel who receive from 32 to 347 mrem/day.
Line 241: Line 187:


1.NUREG-0109, Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1975.2.NUREG-75/032, Occupational Radiation Exposure At Light Water Cooled Power Reactors, 1969-1974.3.NUREG-0463, Occupational Radiation Exposure, Tenth Annual Report, 1977.4.NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station Unit 1.
1.NUREG-0109, Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1975.2.NUREG-75/032, Occupational Radiation Exposure At Light Water Cooled Power Reactors, 1969-1974.3.NUREG-0463, Occupational Radiation Exposure, Tenth Annual Report, 1977.4.NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station Unit 1.
CALLAWAY - SPRev. OL-215/15TABLE 12.4-1  ILLUSTRATIVE EXAMPLES OF DOSE ASSESSMENTOperation and Surveillance:Description of TaskNo. ofDays (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-remFuel bldg.Refueling2910-1136 Aux. bldg.Entire yr5-99-1862Aux. bldg. equip. test1837-650.2Ctmt. (initial survey after SD)1-35-170.1-0.5 Total for oper. & surv.13-30Routine MaintenanceMisc. instrument calibration for pressurizer7319-920.2Aux. bldg. valves19-224-1613-702Aux. bldg. general maintenance8-458-133-2420.3-2 Aux. bldg instrument calibration41115-480.5Ctmt. decon.Refueling3321-1707Fuel bldg. decon.4652-200.3 Radwaste bldg. general maintenance11928-1370.8Radwaste bldg. decon.21-2439-1434Aux. bldg. decon.1143-190.1Ctmt. valves8-7410-2733-1850.9-6Ctmt. instrument calibration8-654-309-750.3-5 Replace oil in reactor coolant pumps5512-450.3Remove charcoal filters for ctmt. cleanup4-57-360.1-0.5Replace charcoal filters for ctmt. cleanup6-452-200.3-0.8Repair dampers & duct84-50.2-0.3General decon. & relamping29-6833-717-11Decon. Refueling canal2-184-230.1-6 CALLAWAY - SPTABLE 12.4-1 (Sheet 2)Rev. OL-215/15Overhaul reactor coolant pump seals21-3018-34Reactor coolant pump motor work3518-306Uncouple/couple reactor coolant pumps2215-190.8Install/remove reactor coolant pump scaffolding127-100.3Repair ctmt. sump pumps & level Indicators9-759-300.4-4Repair pressurizer relief valves5-314-90.2-0.3Replace excore detectors23-4215-301-8Polar crane maintenance9-789-220.3-3Ctmt. pressure test & valve repair6026-322Remove & clean SI system strainers314-250.6Check & repair snubbers3-436-210.1-2Remove equip. hatch5-102-90.002-0.1Replace equip. hatch2-33-70.03-0.04Radwaste ProcessingWaste drum loadingRefueling762-2806Loading spent resin casksRefueling1078-7604Dry waste drum handlingRefueling3-590.2-4 Spent fuel pool filter replacement240.6Transferring resin14-50.1RefuelingInstall and remove reactor cavity seal ring7-167-300.4-0.9Remove transfer tube flange12-172-100.3-0.4Install & remove upper guide structure lifting rig3-316-250.2-3Rig underwater lights2-184-130.02-0.3Description of TaskNo. ofDays (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPTABLE 12.4-1 (Sheet 3)Rev. OL-215/15Flood refueling canal45-90.3Fuel handling (ctmt.)14-1649-923-7Fuel handling (fuel bldg.)5-2825-770.5-1Incore instrumentation removal3-3613-242-6
CALLAWAY - SPRev. OL-215/15TABLE 12.4-1  ILLUSTRATIVE EXAMPLES OF DOSE ASSESSMENTOperation and Surveillance:Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-remFuel bldg.Refueling2910-1136 Aux. bldg.Entire yr5-99-1862Aux. bldg. equip. test1837-650.2Ctmt. (initial survey after SD)1-35-170.1-0.5 Total for oper. & surv.13-30Routine MaintenanceMisc. instrument calibration for pressurizer7319-920.2Aux. bldg. valves19-224-1613-702Aux. bldg. general maintenance8-458-133-2420.3-2 Aux. bldg instrument calibration41115-480.5Ctmt. decon.Refueling3321-1707Fuel bldg. decon.4652-200.3 Radwaste bldg. general maintenance11928-1370.8Radwaste bldg. decon.21-2439-1434Aux. bldg. decon.1143-190.1Ctmt. valves8-7410-2733-1850.9-6Ctmt. instrument calibration8-654-309-750.3-5 Replace oil in reactor coolant pumps5512-450.3Remove charcoal filters for ctmt. cleanup4-57-360.1-0.5Replace charcoal filters for ctmt. cleanup6-452-200.3-0.8Repair dampers & duct84-50.2-0.3General decon. & relamping29-6833-717-11Decon. Refueling canal2-184-230.1-6 CALLAWAY - SPTABLE 12.4-1 (Sheet 2)Rev. OL-215/15Overhaul reactor coolant pump seals21-3018-34Reactor coolant pump motor work3518-306Uncouple/couple reactor coolant pumps2215-190.8Install/remove reactor coolant pump scaffolding127-100.3Repair ctmt. sump pumps & level Indicators9-759-300.4-4Repair pressurizer relief valves5-314-90.2-0.3Replace excore detectors23-4215-301-8Polar crane maintenance9-789-220.3-3Ctmt. pressure test & valve repair6026-322Remove & clean SI system strainers314-250.6Check & repair snubbers3-436-210.1-2Remove equip. hatch5-102-90.002-0.1Replace equip. hatch2-33-70.03-0.04Radwaste ProcessingWaste drum loadingRefueling762-2806Loading spent resin casksRefueling1078-7604Dry waste drum handlingRefueling3-590.2-4 Spent fuel pool filter replacement240.6Transferring resin14-50.1RefuelingInstall and remove reactor cavity seal ring7-167-300.4-0.9Remove transfer tube flange12-172-100.3-0.4Install & remove upper guide structure lifting rig3-316-250.2-3Rig underwater lights2-184-130.02-0.3Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPTABLE 12.4-1 (Sheet 3)Rev. OL-215/15Flood refueling canal45-90.3Fuel handling (ctmt.)14-1649-923-7Fuel handling (fuel bldg.)5-2825-770.5-1Incore instrumentation removal3-3613-242-6


Remove & replace cable to reactor head11-139-270.9-3Remove tool access flanges & uncouple control rods10-139-390.8-9Install bullet noses1-96-150.7-2Remove bullet noses4-52-40.1-0.4Remove & replace head insulation7-104-50.3-0.4RPV studs - remove, replace & clean10-1926-403-13Install & remove alignment pins2-113-120.1-1Connect Polar Crane to Integrated Head Assembly (IHA) tripod & remove RPV head and IHA2-1011-370.6-3Replace RPV head and IHA1-36-180.4-7Decon. PRV head & vessel flanges & replace "O" rings7-1410-243-5Remove & replace holddown ring4-614-313-8Clean stud plugs & holes and install & remove plugs4-1112-261-6Clean incore instrument flanges, studs, & nuts9-2812-322-6RPV head gasket replacement6-3219-351-4Drain, fill, and vent reactor coolant system2-95-100.1-0.7Install handrails around refueling canal146-80.1Description of TaskNo. ofDays (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPTABLE 12.4-1 (Sheet 4)Rev. OL-215/15Total for refueling15-308-6132-34713-66Inservice InspectionSteam generator tubes7-5113-3829-25004-22 Spent fuel sipping201714-2423 Spent fuel inspection5722-301-2Remove & replace primary manway covers4-1412-373-7Remove & replace secondary manway covers7-84-160.4-0.8Steam generator secondary side2-411-170.2-2Install tent around primary manway660.9Install primary loop dams231Clean steam generator manway studs64-120.6Total for ISI10-24Special MaintenanceReplace upper guide structure on control rods697817-78513Sludge lancing of steam generator secondary side51619-1052Steam generator tube plugging7-305-1020-200NOTES:(1)The number of days indicates the period over which the particular task was performed.It does not mean that personnel were working full time for that period.(2)The number of personnel indicates the total number of personnel who worked on aparticular task. It does not mean that all of the personnel were working simultaneouslyor full time on that task.Description of TaskNo. ofDays (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPRev. OL-135/03TABLE 12.4-2  AVERAGE NUMBER OF PERSONNEL PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1, 2, 3)YearNumberofUnitsAverageNumberofPersonnel/Unit196921311970249319713250 1972440119735774197410602 197514548197625593197728642 1969-1977Overall average580NOTES:1.Only PWRs at power levels 450 MWe have been considered, with the exception ofSan Onofre 1 (430 MWe)2.Only PWRs that have been in commercial operation for at least 18 months at the endof the year have been considered. Multiple units have been considered for any yearduring which all units have been in commercial operation for at least 18 months priorto the end of the year. 3.Units considered for both tables are the same, with one exception. For the year 1972,Point Beach Unit 1 was used in Table 12.4-3, but not in Table 12.4-2 since no dataon the number of personnel were available for 1972.
Remove & replace cable to reactor head11-139-270.9-3Remove tool access flanges & uncouple control rods10-139-390.8-9Install bullet noses1-96-150.7-2Remove bullet noses4-52-40.1-0.4Remove & replace head insulation7-104-50.3-0.4RPV studs - remove, replace & clean10-1926-403-13Install & remove alignment pins2-113-120.1-1Connect Polar Crane to Integrated Head Assembly (IHA) tripod & remove RPV head and IHA2-1011-370.6-3Replace RPV head and IHA1-36-180.4-7Decon. PRV head & vessel flanges & replace "O" rings7-1410-243-5Remove & replace holddown ring4-614-313-8Clean stud plugs & holes and install & remove plugs4-1112-261-6Clean incore instrument flanges, studs, & nuts9-2812-322-6RPV head gasket replacement6-3219-351-4Drain, fill, and vent reactor coolant system2-95-100.1-0.7Install handrails around refueling canal146-80.1Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPTABLE 12.4-1 (Sheet 4)Rev. OL-215/15Total for refueling15-308-6132-34713-66Inservice InspectionSteam generator tubes7-5113-3829-25004-22 Spent fuel sipping201714-2423 Spent fuel inspection5722-301-2Remove & replace primary manway covers4-1412-373-7Remove & replace secondary manway covers7-84-160.4-0.8Steam generator secondary side2-411-170.2-2Install tent around primary manway660.9Install primary loop dams231Clean steam generator manway studs64-120.6Total for ISI10-24Special MaintenanceReplace upper guide structure on control rods697817-78513Sludge lancing of steam generator secondary side51619-1052Steam generator tube plugging7-305-1020-200NOTES:(1)The number of days indicates the period over which the particular task was performed.It does not mean that personnel were working full time for that period.(2)The number of personnel indicates the total number of personnel who worked on aparticular task. It does not mean that all of the personnel were working simultaneouslyor full time on that task.Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPRev. OL-135/03TABLE 12.4-2  AVERAGE NUMBER OF PERSONNEL PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1, 2, 3)YearNumberofUnitsAverageNumberofPersonnel/Unit196921311970249319713250 1972440119735774197410602 197514548197625593197728642 1969-1977Overall average580NOTES:1.Only PWRs at power levels 450 MWe have been considered, with the exception ofSan Onofre 1 (430 MWe)2.Only PWRs that have been in commercial operation for at least 18 months at the endof the year have been considered. Multiple units have been considered for any yearduring which all units have been in commercial operation for at least 18 months priorto the end of the year. 3.Units considered for both tables are the same, with one exception. For the year 1972,Point Beach Unit 1 was used in Table 12.4-3, but not in Table 12.4-2 since no dataon the number of personnel were available for 1972.
CALLAWAY - SPRev. OL-135/03TABLE 12.4-3  AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE) PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1, 2, 3)YearNumberofUnitsAverage Man-Rem Dose/Unit19692741970242219713274 1972548219735610197410439 197514460197625461197728422 1969-1977Overall average441NOTES:1.Only PWRs at power levels 450 MWe have been considered, with the exception ofSan Onofre 1 (430 MWe)2.Only PWRs that have been in commercial operation for at least 18 months at the endof the year have been considered. Multiple units have been considered for any yearduring which all units have been in commercial operation for at least 18 months priorto the end of the year. 3.Units considered for both tables are the same, with one exception. For the year 1972,Point Beach Unit 1 was used in Table 12.4-3, but not in Table 12.4-2 since no dataon the number of personnel were available for 1972.
CALLAWAY - SPRev. OL-135/03TABLE 12.4-3  AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE) PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1, 2, 3)YearNumberofUnitsAverage Man-Rem Dose/Unit19692741970242219713274 1972548219735610197410439 197514460197625461197728422 1969-1977Overall average441NOTES:1.Only PWRs at power levels 450 MWe have been considered, with the exception ofSan Onofre 1 (430 MWe)2.Only PWRs that have been in commercial operation for at least 18 months at the endof the year have been considered. Multiple units have been considered for any yearduring which all units have been in commercial operation for at least 18 months priorto the end of the year. 3.Units considered for both tables are the same, with one exception. For the year 1972,Point Beach Unit 1 was used in Table 12.4-3, but not in Table 12.4-2 since no dataon the number of personnel were available for 1972.
CALLAWAY - SPRev. OL-135/03TABLE 12.4-4  AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE) BASED ON PWR PLANT AGE (REF 1, 2, 3)Initial Operation DateUnits ConsideredMan-RemperUnit123330 664 571 775 481 341 465 665 745No. of Units1415 13 9
CALLAWAY - SPRev. OL-135/03TABLE 12.4-4  AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE) BASED ON PWR PLANT AGE (REF 1, 2, 3)Initial Operation DateUnits ConsideredMan-RemperUnit123330 664 571 775 481 341 465 665 745No. of Units 14 15 13 9
8 5
8 5
4 3
4 3
Line 256: Line 202:
(93-102)
(93-102)
(105-114)
(105-114)
(117-126) 12 3
(117-126)1 2 3
4 5
4 5
6 7
6 7
Line 269: Line 215:
Waste processing3.5%6.9%5.0%5.9%
Waste processing3.5%6.9%5.0%5.9%
Refueling14.0%7.7%7.9%6.6%
Refueling14.0%7.7%7.9%6.6%
Notes on Tables 12.4-5
Notes on Tables 12.4-5 , 6 and 7:1.PWRs and BWRs operating at all power levels were considered in compiling thesetables.2.Percentages for 1974 and 1975 are based on approximately 39 percent and 50percent of the total exposures reported in the appropriate year for light water reactors which had been in commercial operation for at least one full year, as of 12/31/74 and 12/31/75, respectively.3.Percentages for 1976 and 1977 are on only those facilities which have been incommercial operation for at least one full year, as of 12/31/76 and 12/31/77, respectively.4.Distributions of personnel receiving an annual exposure of greater than 100millirems, according to either work function or employee category, are not availablefor 1976 or 1977.
, 6 and 7:1.PWRs and BWRs operating at all power levels were considered in compiling thesetables.2.Percentages for 1974 and 1975 are based on approximately 39 percent and 50percent of the total exposures reported in the appropriate year for light water reactors which had been in commercial operation for at least one full year, as of 12/31/74 and 12/31/75, respectively.3.Percentages for 1976 and 1977 are on only those facilities which have been incommercial operation for at least one full year, as of 12/31/76 and 12/31/77, respectively.4.Distributions of personnel receiving an annual exposure of greater than 100millirems, according to either work function or employee category, are not availablefor 1976 or 1977.
CALLAWAY - SPRev. OL-135/03TABLE 12.4-8  ANNUAL OCCUPATIONAL EXPOSURES FOR VARIOUS PWR VENDOR'S UNITS (REF. 1,2,3)Notes - On Table 12.4-8:1.Only PWRs operating at power levels >450 MWe have been considered with the exception of San Onofre 1 (430 MWe).2.The power plants for this table are the same as those appearing on Table 12.4-4
CALLAWAY - SPRev. OL-135/03TABLE 12.4-8  ANNUAL OCCUPATIONAL EXPOSURES FOR VARIOUS PWR VENDOR'S UNITS (REF. 1,2,3)Notes - On Table 12.4-8:1.Only PWRs operating at power levels >450 MWe have been considered with the exception of San Onofre 1 (430 MWe).2.The power plants for this table are the same as those appearing on Table 12.4-4
.WestinghouseC-EB&WApprox Yr. ofOperationNo. of ReactorsMan-Rem Yr-UnitNo. of ReactorsMan-Rem Yr-UnitNo. of    Man-Rem Reactors  Yr-UnitComments171495117247---------------2824345223 211CE: Palisades - 1,1313760834193 335W:  Ginna - 1,032 CE: Palisades produces very little power.467413229------W:  Surry 1 & 2 - 3,165 568762471------W:  Ginna 1 - 1,224Robinson 2 - 1,142 Surry 1 & 2 - 2,307 645531100---------------------74341---------------------------83443---------------------------92665------------W:  San Onofre - 880 102745manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------
.WestinghouseC-EB&WApprox Yr. ofOperationNo. of ReactorsMan-Rem Yr-UnitNo. of ReactorsMan-Rem Yr-UnitNo. of    Man-Rem Reactors  Yr-UnitComments171495117247---------------2824345223 211CE: Palisades - 1,1313760834193 335W:  Ginna - 1,032 CE: Palisades produces very little power.467413229------W:  Surry 1 & 2 - 3,165 568762471------W:  Ginna 1 - 1,224Robinson 2 - 1,142 Surry 1 & 2 - 2,307 645531100---------------------74341---------------------------83443---------------------------92665------------W:  San Onofre - 880 102745manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------
Line 278: Line 223:
CALLAWAY - SPRev. OL-135/03TABLE 12.4-10  AVERAGE INDIVIDUAL EXPOSURE BASED ON PWR PLANT AGE (REF. 1,2,3)Notes on Table 12.4-10:1.Units considered are the same as those shown on Table 6, except that Palisades has been omitted from the first year average since no data is available for the number of personnel for that year.2.The high exposure average for years 3-5 include the following:Year of Operation (approx)No. ofUnitsAve. No. of Personnel/UnitAve. Exposure man-rem  yr-unitAve. Exposure per Individual (rem-yr)1133621230.342155323300.62 3136616641.00 496235710.92 586917751.12 656274810.77 745403410.63 835834650.80 929876650.671029407450.79PlantYearGinna Average ExposureSurry 1 & 2 Average ExposureRobinson Average Exposure31.52------4---1.15---
CALLAWAY - SPRev. OL-135/03TABLE 12.4-10  AVERAGE INDIVIDUAL EXPOSURE BASED ON PWR PLANT AGE (REF. 1,2,3)Notes on Table 12.4-10:1.Units considered are the same as those shown on Table 6, except that Palisades has been omitted from the first year average since no data is available for the number of personnel for that year.2.The high exposure average for years 3-5 include the following:Year of Operation (approx)No. ofUnitsAve. No. of Personnel/UnitAve. Exposure man-rem  yr-unitAve. Exposure per Individual (rem-yr)1133621230.342155323300.62 3136616641.00 496235710.92 586917751.12 656274810.77 745403410.63 835834650.80 929876650.671029407450.79PlantYearGinna Average ExposureSurry 1 & 2 Average ExposureRobinson Average Exposure31.52------4---1.15---
51.391.241.35Rem/yrRem/yrRem/yr CALLAWAY - SPRev. OL-135/03TABLE 12.4-11  CUMULATIVE AVERAGE OF ANNUAL EXPOSURE BY YEARS OF OPERATION - PWRS (REF. 1, 2, 3)Note: 1.Plants considered are the same as those shown on Table 12.4-4
51.391.241.35Rem/yrRem/yrRem/yr CALLAWAY - SPRev. OL-135/03TABLE 12.4-11  CUMULATIVE AVERAGE OF ANNUAL EXPOSURE BY YEARS OF OPERATION - PWRS (REF. 1, 2, 3)Note: 1.Plants considered are the same as those shown on Table 12.4-4
.Years of OperationPlant12345678910Conn. Yankee-106398379366432394438439462San Onofre 1-429982126171155174226327Ginna207319556478627613616589 Robinson 2364290425484618634608Palisades78606613536568490Maine Yankee117269285235237 Surry 1 & 21525188954631631Fort Calhoun 171183226244Kewaunee28149146 Three Mile Island 173180240Arkansas 121155189Rancho Seco-58225 Cook116208Millstone Pt 2168205Trojan174 St. Lucie152Total17213288429739044173234017731201701789No. of Units1415139854322 Average123219331434522468433400351395 CALLAWAY - SPRev. OL-135/03TABLE 12.4-12  ESTIMATES OF OCCUPANCY TIMES IN PLANT RADIATION AREAS AND GAMMA DOSES TO PLANT PERSONNELOperationZonePercentage ofOccupancyHrs/yrHourly DoseRate(Rems/Hr)Yearly DoseRate(Rems/Yr) No. of men Annual Exposures(Man-Rem/yr-Unit)Operation & SurveillanceA751,5601 x 10-40.15385.7B214375 x 10-40.21388.3C3622.5 x 10-30.15385.9D0.9191.5 x 10-20.283810.8E0.121 x 10-10.20387.6Total1002,0800.9938.3Routine  MaintenanceA751,5601 x 10-40.15284.4B112295 x 10-40.11283.2C10.52182.5 x 10-30.542815.3D2.5521.5 x 10-20.782821.8E1211 x 10-12.12858.8Total1002,0803.68104Radwaste ProcessingA701,4561 x 10-40.1481.16B204165 x 10-40.280.68C81662.5 x 10-30.4183.28D1211.5 x 10-20.3182.52E1211 x 10-12.1816.8Total1002,0803.1624.4RefuelingA3048 1 x 10-40.00460.22B40645 x 10-40.0361.47C24392.5 x 10-30.0964.48 CALLAWAY - SPTABLE 12.4-12 (Sheet 2)Rev. OL-135/03OperationZonePercentage ofOccupancyHrs/yrHourly DoseRate(Rems/Hr)Yearly DoseRate(Rems/Yr) No. of men Annual Exposures (Man-Rem/yr-Unit)D461.5 x 10-20.0964.14E231 x 10-10.30613.8Total1001600.5124.1Inservice Inspection (ISI)A551321 x 10-40.01460.6B36865 x 10-40.04461.96C7.0172.5 x 10-30.04461.96D1.02.51.5 x 10-20.03461.72E12.52 x 10-10.54623Total1002400.6229.2Special MaintenanceA75240 1 x 10-40.0242004.8B20645 x 10-40.032006.4C3102.5 x 10-30.0252005D131.5 x 10-20.042009E132 x 10-10.6200120Total1003200.71145 CALLAWAY - SPRev. OL-135/03TABLE 12.4-13  DISTRIBUTION OF DIRECT RADIATION MAN-REM DOSES ACCORDING TO WORK FUNCTIONSOperationAnnual Exposures (Man-rem/year Unit)PercentageOperation and surveillance38.310.5Routine maintenance10428.5 Radwaste processing24.46.7 Refueling24.16.6 Inservice inspection29.28 Special maintenance14539.7 Total365100 CALLAWAY - SPRev. OL-135/03TABLE 12.4-14  ANNUAL OCCUPANCY IN PLANT AREAS CONTAINING AIRBORNE RADIOACTIVITYBuildinghr/yrper manNo. menman-hours yrContainment-power2503750Containment-refueling62.5322,000Fuel building-power2502500Fuel building-refueling125141,750 CALLAWAY - SPRev. OL-135/03TABLE 12.4-15  DOSES TO PLANT PERSONNEL CAUSED BY AIRBORNE RADIOACTIVITYLocationDose Rate mrem/hrAnnual Occupancy (man-hr/yr)Annual Expose (man-rem/yr)ThyroidLungWhole BodyOperation
.Years of OperationPlant12345678910Conn. Yankee-106398379366432394438439462San Onofre 1-429982126171155174226327Ginna207319556478627613616589 Robinson 2364290425484618634608Palisades78606613536568490Maine Yankee117269285235237 Surry 1 & 21525188954631631Fort Calhoun 171183226244Kewaunee28149146 Three Mile Island 173180240Arkansas 121155189Rancho Seco-58225 Cook116208Millstone Pt 2168205Trojan174 St. Lucie152Total17213288429739044173234017731201701789No. of Units1415139854322 Average123219331434522468433400351395 CALLAWAY - SPRev. OL-135/03TABLE 12.4-12  ESTIMATES OF OCCUPANCY TIMES IN PLANT RADIATION AREAS AND GAMMA DOSES TO PLANT PERSONNELOperationZonePercentage ofOccupancyHrs/yrHourly DoseRate(Rems/Hr)Yearly DoseRate(Rems/Yr) No. of men Annual Exposures(Man-Rem/yr-Unit)Operation & SurveillanceA751,560 1 x 10-40.15385.7B21437 5 x 10-40.21388.3C3622.5 x 10-30.15385.9D0.9191.5 x 10-20.283810.8E0.12 1 x 10-10.20387.6Total1002,0800.9938.3Routine  MaintenanceA751,560 1 x 10-40.15284.4B11229 5 x 10-40.11283.2C10.52182.5 x 10-30.542815.3D2.5521.5 x 10-20.782821.8E121 1 x 10-12.12858.8Total1002,0803.68104Radwaste ProcessingA701,456 1 x 10-40.1481.16B20416 5 x 10-40.280.68C81662.5 x 10-30.4183.28D1211.5 x 10-20.3182.52E121 1 x 10-12.1816.8Total1002,0803.1624.4RefuelingA3048 1 x 10-40.00460.22B4064 5 x 10-40.0361.47C24392.5 x 10-30.0964.48 CALLAWAY - SPTABLE 12.4-12 (Sheet 2)Rev. OL-135/03OperationZonePercentage ofOccupancyHrs/yrHourly DoseRate(Rems/Hr)Yearly DoseRate(Rems/Yr) No. of men Annual Exposures (Man-Rem/yr-Unit)D461.5 x 10-20.0964.14E23 1 x 10-10.30613.8Total1001600.5124.1Inservice Inspection (ISI)A55132 1 x 10-40.01460.6B3686 5 x 10-40.04461.96C7.0172.5 x 10-30.04461.96D1.02.51.5 x 10-20.03461.72E12.5 2 x 10-10.54623Total1002400.6229.2Special MaintenanceA75240 1 x 10-40.0242004.8B2064 5 x 10-40.032006.4C3102.5 x 10-30.0252005D131.5 x 10-20.042009E13 2 x 10-10.6200120Total1003200.71145 CALLAWAY - SPRev. OL-135/03TABLE 12.4-13  DISTRIBUTION OF DIRECT RADIATION MAN-REM DOSES ACCORDING TO WORK FUNCTIONSOperationAnnual Exposures (Man-rem/year Unit)PercentageOperation and surveillance38.310.5Routine maintenance10428.5 Radwaste processing24.46.7 Refueling24.16.6 Inservice inspection29.28 Special maintenance14539.7 Total365100 CALLAWAY - SPRev. OL-135/03TABLE 12.4-14  ANNUAL OCCUPANCY IN PLANT AREAS CONTAINING AIRBORNE RADIOACTIVITYBuilding hr/yrper manNo. menman-hours yrContainment-power2503750Containment-refueling62.5322,000Fuel building-power2502500Fuel building-refueling125141,750 CALLAWAY - SPRev. OL-135/03TABLE 12.4-15  DOSES TO PLANT PERSONNEL CAUSED BY AIRBORNE RADIOACTIVITYLocationDose Rate mrem/hrAnnual Occupancy (man-hr/yr)Annual Expose (man-rem/yr)ThyroidLungWhole BodyOperation& SurveillanceMaintenanceRefuelingTotalThyroidLungWhole BodyContainment - power8.640.050.48100650---7506.480.040.36Containment - refueling1.340.93N---1000100020002.681.86NFuel building - power0.620.62N400100---5000.310.31NFuel building - refueling4.713.28N------175017508.255.74NTotal1880.4N  =  Negligible CALLAWAY - SP12.5-1Rev. OL-155/0612.5RADIATION PROTECTION PROGRAM For descriptions of the radiation protection program, refer to Section 12.5 of the Site Addendum.
& SurveillanceMaintenanceRefuelingTotalThyroidLungWhole BodyContainment - power8.640.050.48100650---7506.480.040.36Containment - refueling1.340.93N---1000100020002.681.86NFuel building - power0.620.62N400100---5000.310.31NFuel building - refueling4.713.28N------175017508.255.74NTotal1880.4N  =  Negligible CALLAWAY - SP12.5-1Rev. OL-155/0612.5RADIATION PROTECTION PROGRAM For descriptions of the radiation protection program, refer to Section 12.5 of the Site Addendum.
CALLAWAY - SA12.0-iTABLE OF CONTENTSCHAPTER 12.0RADIATION PROTECTION Section Page12.1ENSURING THAT OCCUPATIONAL RADIATION EXPOSURE S ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA).........
CALLAWAY - SA12.0-iTABLE OF CONTENTSCHAPTER 12.0RADIATION PROTECTION SectionPage12.1ENSURING THAT OCCUPATIONAL RADIATION EXPOSURE S ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA).........
................
................
........................12.1-112.1.1POLICY CONSIDERATIONS.....
........................12.1-112.1.1POLICY CONSIDERATIONS.....
Line 354: Line 298:
.................
.................
..........................12.5-7 CALLAWAY - SATABLE OF CONTENTS (Continued)
..........................12.5-7 CALLAWAY - SATABLE OF CONTENTS (Continued)
SectionPage12.0-ii12.5.2.2.6Laundry Decontamination Facility ....
Section Page12.0-ii12.5.2.2.6Laundry Decontamination Facility ....
..............
..............
..............
..............
Line 416: Line 360:
This policy is promul gated in general terms by the Senior Vice President and Chief Nuclear Officer and is implemented by specific procedures and directives at the plant level by the Senior Director, Nuclear Operations. Implementation of the ALARA program will be in accordance with the requirements of 10CFR19 and 10CFR20 and will follow the intent of the guidance prov ided in applicable sections of Regulatory Guides 1.8, 8.8 and 8.10.12.1.1.2Organizational StructureThe Senior Director, Nuclear Operations, th rough the Radiation Protection Department staff, is ultimately responsible for development, implementation, and continuing conduct of the overall plant Radiation Protection program of which the ALARA program is a part. The Manager, Radiation Protection directs the Radiation Protection Department and is responsible to the Plant Di rector for the ALARA program. The Manager, Radiation Protection has the responsi bility and authority for de veloping, impl ementing and managing the plant ALARA program.
This policy is promul gated in general terms by the Senior Vice President and Chief Nuclear Officer and is implemented by specific procedures and directives at the plant level by the Senior Director, Nuclear Operations. Implementation of the ALARA program will be in accordance with the requirements of 10CFR19 and 10CFR20 and will follow the intent of the guidance prov ided in applicable sections of Regulatory Guides 1.8, 8.8 and 8.10.12.1.1.2Organizational StructureThe Senior Director, Nuclear Operations, th rough the Radiation Protection Department staff, is ultimately responsible for development, implementation, and continuing conduct of the overall plant Radiation Protection program of which the ALARA program is a part. The Manager, Radiation Protection directs the Radiation Protection Department and is responsible to the Plant Di rector for the ALARA program. The Manager, Radiation Protection has the responsi bility and authority for de veloping, impl ementing and managing the plant ALARA program.
The organization of the Radiaton Protection Department is described in Section 12.5.1
The organization of the Radiaton Protection Department is described in Section 12.5.1
.ALARA reviews can be performed by qualif ied nuclear engineeri ng personnel (i.e., Design Engineering, Technica l Support Engineering, Systems Engineering, or Radiation Protection) for plant modifications or maintenance activities.12.1.1.3Application of the Union Electric ALARA Po licy to Plant Operations The Callaway Plant ALARA program is implemented throug h the adherence to Administrative and Radiation Pr otection procedures which cover, but are not limited to, the following subjects:a.Qualifications of and the conduct of operations of the plant radiation protection department.b.ALARA program administration - re viewing job planni ng, job exposure data, radiation work permit comparisons wit h personnel exposure, CALLAWAY - SA12.1-2Rev. OL-21 5/15procedure reviews, review of design change packages for installation and long term operations and maintenance concerns, and trends of dose data for potential progr ammatic changes.c.Training of personnel in the fundamentals of ra diation protection and in health physics exposur e control procedures.d.Radiation Work Permit Programe.Internal Pers onnel Monitoringf.External Personnel Monitoring g.Area Postingh.Radioactive Material Controli.Radiological Surveys j.Instrumentationk.Radiological Incidentsl.Radiation Work Practices m.Environmental Monitoringn.Respiratory Protection CALLAWAY - SA12.4-1Rev. OL-13 5/0312.4DOSE ASSESSMENTThis section contained the dose assessment to the constructi on workers during the construction of Callaway Plant Un it 2. With the ca ncellation of Callaway Plant Unit 2 in October 1981, this secti on is no longer applicable.
.ALARA reviews can be performed by qualif ied nuclear engineeri ng personnel (i.e., Design Engineering, Technica l Support Engineering, Systems Engineering, or Radiation Protection) for plant modifications or maintenance activities.12.1.1.3Application of the Union Electric ALARA Po licy to Plant Operations The Callaway Plant ALARA program is implemented throug h the adherence to Administrative and Radiation Pr otection procedures which cover, but are not limited to, the following subjects:a.Qualifications of and the conduct of operations of the plant radiation protection department.b.ALARA program administration - re viewing job planni ng, job exposure data, radiation work permit comparisons wit h personnel exposure, CALLAWAY - SA12.1-2Rev. OL-21 5/15 procedure reviews, review of design change packages for installation and long term operations and maintenance concerns, and trends of dose data for potential progr ammatic changes.c.Training of personnel in the fundamentals of ra diation protection and in health physics exposur e control procedures.d.Radiation Work Permit Programe.Internal Pers onnel Monitoringf.External Personnel Monitoring g.Area Postingh.Radioactive Material Controli.Radiological Surveys j.Instrumentationk.Radiological Incidentsl.Radiation Work Practices m.Environmental Monitoringn.Respiratory Protection CALLAWAY - SA12.4-1Rev. OL-13 5/0312.4DOSE ASSESSMENTThis section contained the dose assessment to the constructi on workers during the construction of Callaway Plant Un it 2. With the ca ncellation of Callaway Plant Unit 2 in October 1981, this secti on is no longer applicable.
CALLAWAY - SA12.5-1Rev. OL-21 5/1512.5RADIATION PROTECTION PROGRAM12.5.1ORGANIZATION12.5.1.1Radiation Protecti on Program ObjectivesThe objective of the radiation protection program for the Callaway Plant is to maintain the occupational radiation exposure of personnel working at the pl ant as low as is reasonably achievable (ALARA). This objecti ve is accomplished by adherence to the requirements of Title 10 of t he Code of Federal R egulations, Part 20, by following the intent of guidance given in USNRC Regulatory Guides 8.2, 8.8, 8.10, 8.34, 8.35, 8.36 and 1.8, and by using industry acce pted radiation protection practices.
CALLAWAY - SA12.5-1Rev. OL-21 5/1512.5RADIATION PROTECTION PROGRAM12.5.1ORGANIZATION12.5.1.1Radiation Protecti on Program ObjectivesThe objective of the radiation protection program for the Callaway Plant is to maintain the occupational radiation exposure of personnel working at the pl ant as low as is reasonably achievable (ALARA). This objecti ve is accomplished by adherence to the requirements of Title 10 of t he Code of Federal R egulations, Part 20, by following the intent of guidance given in USNRC Regulatory Guides 8.2, 8.8, 8.10, 8.34, 8.35, 8.36 and 1.8, and by using industry acce pted radiation protection practices.
The radiation protection pr ogram is administered by t he organization described in subsections 12.5.1.2 and 13.1.2.2.2 to accomplish the following:a.Identify and review radiation protection training requirements for personnel assigned to work in radiological cont rolled areas commens urate with their duties, responsibilities, and the degree of radiation hazards anticipated.
The radiation protection pr ogram is administered by t he organization described in subsections 12.5.1.2 and 13.1.2.2.2 to accomplish the following:a.Identify and review radiation protection training requirements for personnel assigned to work in radiological cont rolled areas commens urate with their duties, responsibilities, and the degree of radiation hazards anticipated.
Inform these personnel of methods for maintaining occupational radiation exposure ALARA and assist them in carrying out t heir radiation safety responsibilities.b.Evaluate and review the radiological status of the plant by monitoring radiation, contamination and airborne radioactivity levels to control or eliminate radiological hazards.c.Control external and in ternal radiation exposure of perso nnel through the implementation of radiological controls during operations and maintenance.d.Review and evaluate radiation prot ection records and appropriate plant operating and maintenance procedures fo r methods to reduce radiation exposure to personnel.e.Maintain and evaluate personnel exposure reco rds to ensure that occupational radiation exposures are maintained ALARA.f.Maintain control of radioactive mate rials on-site and maintain releases of radioactive materials in effluent s to unrestricted areas ALARA.g.Evaluate and review t he swipe samples from instruments containing radioactive sources that other Union Electric fossil power plants have provided.
Inform these personnel of methods for maintaining occupational radiation exposure ALARA and assist them in carrying out t heir radiation safety responsibilities.b.Evaluate and review the radiological status of the plant by monitoring radiation, contamination and airborne radioactivity levels to control or eliminate radiological hazards.c.Control external and in ternal radiation exposure of perso nnel through the implementation of radiological controls during operations and maintenance.d.Review and evaluate radiation prot ection records and appropriate plant operating and maintenance procedures fo r methods to reduce radiation exposure to personnel.e.Maintain and evaluate personnel exposure reco rds to ensure that occupational radiation exposures are maintained ALARA.f.Maintain control of radioactive mate rials on-site and maintain releases of radioactive materials in effluent s to unrestricted areas ALARA.g.Evaluate and review t he swipe samples from instruments containing radioactive sources that other Union Electric fossil power plants have provided.
CALLAWAY - SA12.5-2Rev. OL-21 5/1512.5.1.2Organization of the Radiation Protection Department The experience and qualificatio ns of the personnel respons ible for administering the radiation protection pr ogram are presented in Section 13.1.3.1
CALLAWAY - SA12.5-2Rev. OL-21 5/1512.5.1.2Organization of the Radiation Protection Department The experience and qualificatio ns of the personnel respons ible for administering the radiation protection pr ogram are presented in Section 13.1.3.1. The responsibilities of the personnel are discussed in Section 13.1.2.2.2
. The responsibilities of the personnel are discussed in Section 13.1.2.2.2
.The Radiation Protection  Department is comprised of personnel in the following classifications:a.Manager, Radiation Protectionb.Supervising Health Physicist c.Health Physicists/Nuclear Scientists d.Radiation Protection Supervisorse.Radiation Protection Techniciansf.Radiation/Chemical Helpers 12.5.1.3Personnel Training The training of plant per sonnel in radiation prot ection is described in Section 12.5.3.4. 12.5.2EQUIPMENT, INSTRUMENTATION AND FACILITIES12.5.2.1Equipment and InstrumentationRadiation Protection  instrumentation and equipment will be availabl e for the assessment of plant radiological conditions and to support the oper ation of the Ca llaway Plant Radiation Protection Program.
.The Radiation Protection  Department is comprised of personnel in the following classifications:a.Manager, Radiation Protectionb.Supervising Health Physicist c.Health Physicists/Nuclear Scientists d.Radiation Protection Supervisorse.Radiation Protection Techniciansf.Radiation/Chemical Helpers 12.5.1.3Personnel Training The training of plant per sonnel in radiation prot ection is described in Section 12.5.3.4
The three classifications of equipment that will be utilized in radiological monitoring and su rveillance activities include:a.Installed Equipmentb.Laboratory Equipmentc.Portable EquipmentInstalled equipment includes bot h the in-plant area radiat ion monitoring system and the process and effluent radioacti vity monitoring system. Laboratory equipment consists of analytical instrumentation used to analyze and quantify radioactivity. Portable equipment includes portable radiation detection instrumentation, ai r samplers and personnel dosimetry used for performing radiation and contaminat ion surveys, airborne radioactivity monito ring and sampling and personnel monitoring.
. 12.5.2EQUIPMENT, INSTRUMENTATION AND FACILITIES12.5.2.1Equipment and InstrumentationRadiation Protection  instrumentation and equipment will be availabl e for the assessment of plant radiological conditions and to support the oper ation of the Ca llaway Plant Radiation Protection Program.
CALLAWAY - SA12.5-3Rev. OL-21 5/15Portable and laboratory instrumentation used in the radiation protecti on program will be selected in accordance with the following criteria:a.Ability to measure the quantity of interest to an acceptable degree of precision and accuracy.b.Ease of operation, ma intenance, and calibration.c.Appropriate sensitivity and range for various operati onal situations, including normal operations, anticipated operati onal occurrences and accident conditions, as determined by the requirements of applicable regulations.d.Operational reliability.12.5.2.1.1Installed Radiat ion Monitoring EquipmentInstalled area radiation monito rs and continuous air monitors are described in Section 12.3. Process and effluent monitors are described in Section 11.5.12.5.2.1.2Radiation Protection Laboratory Instrumentation Radiation protection laboratory instrumentation will in clude a gamma spectroscopy system, a  liquid scintillation counter, a gas flow pr oportional c ounter and a geiger-mueller type counting system. Additional info rmation on radiation protection laboratory equipment is provided in Table 12.5-1. The primary loca tion of radiation protection laboratory equipment will be the count room, 1984' elevati on of the Control Building. However, additi onal counting equipment will be available at other plant locations including the Emergency Operations Facility. 12.5.2.1.3Portable Radiation Detection InstrumentationPortable radiation detection instrumentation is described in Table 12.5-2 which includes information on the type, range, accuracy and typical quantities of instruments. Survey instruments will be available in-plant to facilitate access. Instruments requiring calibration or repair will be identified as out-of-service and segregated from in-use equipment.12.5.2.1.4Portable Ai r Sampling EquipmentPortable air sampling equipment includes high and low-volume air samplers capable of accepting particulate filters and charcoal cartridges for grab samp les of radioactive particulates and iodine. In addition, portable continuous air monitors, typical of that described in Table 12.5-2 , will be used for continuous surveill ance of airborne radioactivity levels. Air sampling instruments are calibrat ed periodically in accordance with an established calibration program.
The three classifications of equipment that will be utilized in radiological monitoring and su rveillance activities include:a.Installed Equipmentb.Laboratory Equipmentc.Portable EquipmentInstalled equipment includes bot h the in-plant area radiat ion monitoring system and the process and effluent radioacti vity monitoring system
. Laboratory equipment consists of analytical instrumentation used to analyze and quantify radioactivity. Portable equipment includes portable radiation detection instrumentation, ai r samplers and personnel dosimetry used for performing radiation and contaminat ion surveys, airborne radioactivity monito ring and sampling and personnel monitoring.
CALLAWAY - SA12.5-3Rev. OL-21 5/15Portable and laboratory instrumentation used in the radiation protecti on program will be selected in accordance with the following criteria:a.Ability to measure the quantity of interest to an acceptable degree of precision and accuracy.b.Ease of operation, ma intenance, and calibration.c.Appropriate sensitivity and range for various operati onal situations, including normal operations, anticipated operati onal occurrences and accident conditions, as determined by the requirements of applicable regulations.d.Operational reliability.12.5.2.1.1Installed Radiat ion Monitoring EquipmentInstalled area radiation monito rs and continuous air monitors are described in Section 12.3. Process and effluent monitors are described in Section 11.5.12.5.2.1.2Radiation Protection Laboratory Instrumentation Radiation protection laboratory instrumentation will in clude a gamma spectroscopy system, a  liquid scintillation counter, a gas flow pr oportional c ounter and a geiger-mueller type counting system. Additional info rmation on radiation protection laboratory equipment is provided in Table 12.5-1
. The primary loca tion of radiation protection laboratory equipment will be the count room, 1984' elevati on of the Control Building. However, additi onal counting equipment will be available at other plant locations including the Emergency Operations Facility. 12.5.2.1.3Portable Radiation Detection InstrumentationPortable radiation detection instrumentation is described in Table 12.5-2 which includes information on the type, range, accuracy and typical quantities of instruments. Survey instruments will be available in-plant to facilitate access. Instruments requiring calibration or repair will be identified as out-of-service and segregated from in-use equipment.12.5.2.1.4Portable Ai r Sampling EquipmentPortable air sampling equipment includes high and low-volume air samplers capable of accepting particulate filters and charcoal cartridges for grab samp les of radioactive particulates and iodine. In addition, portable continuous air monitors, typical of that described in Table 12.5-2
, will be used for continuous surveill ance of airborne radioactivity levels. Air sampling instruments are calibrat ed periodically in accordance with an established calibration program.
CALLAWAY - SA12.5-4Rev. OL-21 5/1512.5.2.1.5Instrument CalibrationCalibration of portable and laboratory radi ation protection instrumentation will be performed in accordance with an established calibration program. Requirements of this program include as a minimum:a.written procedures for operation a nd calibration of each instrument.b.a mechanism for tracking and recall of instruments for calibration.c.established calibration frequencies fo r each instrument and recalibration following maintenance and repair that could change the instrument performance characteristics establishe d during the prev ious calibration.d.use of calibration standar ds which are traceable to the National Institute of Standards and Tec hnology (NIST).e.required calibration a ccuracies and tolerances.f.periodic response checking of instruments to verify continuing proper operation.g.maintenance of records document ing calibration activities.h.tagging and labeling of instruments.12.5.2.1.6Protective Clothing Protective clothing is prescribed and issued to plant personnel by the radiation protection staff based upon the actual or potential radiological condit ions expected for the job assignment. Protective clothing stations are established at strategic loca tions within the plant as required to ensure efficient operations and to preclude the spreading of contamination. Protective clothing avai lable at the plant in cludes the following:a.Coveralls.b.Caps and hoods.c.Shoe covers.
CALLAWAY - SA12.5-4Rev. OL-21 5/1512.5.2.1.5Instrument CalibrationCalibration of portable and laboratory radi ation protection instrumentation will be performed in accordance with an established calibration program. Requirements of this program include as a minimum:a.written procedures for operation a nd calibration of each instrument.b.a mechanism for tracking and recall of instruments for calibration.c.established calibration frequencies fo r each instrument and recalibration following maintenance and repair that could change the instrument performance characteristics establishe d during the prev ious calibration.d.use of calibration standar ds which are traceable to the National Institute of Standards and Tec hnology (NIST).e.required calibration a ccuracies and tolerances.f.periodic response checking of instruments to verify continuing proper operation.g.maintenance of records document ing calibration activities.h.tagging and labeling of instruments.12.5.2.1.6Protective Clothing Protective clothing is prescribed and issued to plant personnel by the radiation protection staff based upon the actual or potential radiological condit ions expected for the job assignment. Protective clothing stations are established at strategic loca tions within the plant as required to ensure efficient operations and to preclude the spreading of contamination. Protective clothing avai lable at the plant in cludes the following:a.Coveralls.b.Caps and hoods.c.Shoe covers.
d.Plastic, rubber, and cotton gloves.
d.Plastic, rubber, and cotton gloves.
Line 454: Line 393:
The plant arrangement is such that traffi c between radiol ogical controlled areas and uncontrolled areas is r outed through access c ontrol. The access control facilities for normal traffic are located in the control building where peopl e entering and leaving all controlled areas will check for contamination. Within t he controlled area, access is controlled by utilizing lock ed and/or annunciated doors and gates, fences, alarms, rope barriers, and the posting of signs.
The plant arrangement is such that traffi c between radiol ogical controlled areas and uncontrolled areas is r outed through access c ontrol. The access control facilities for normal traffic are located in the control building where peopl e entering and leaving all controlled areas will check for contamination. Within t he controlled area, access is controlled by utilizing lock ed and/or annunciated doors and gates, fences, alarms, rope barriers, and the posting of signs.
High  radiation  areas  in  whic h  the  radiation  field  is greater than  100 mrem/hr  but less than 1000 mrem/hr  wil l be  posted  as  a high  radiation  area and entrance  into  the area  will be controlled by the radiation work  permit  program. High  radiation  areas in  which the  radiation  fiel d  is  greater  t han or equal to 1000 mr em/hr will be provided with locked doors to prevent unauthorized entry.
High  radiation  areas  in  whic h  the  radiation  field  is greater than  100 mrem/hr  but less than 1000 mrem/hr  wil l be  posted  as  a high  radiation  area and entrance  into  the area  will be controlled by the radiation work  permit  program. High  radiation  areas in  which the  radiation  fiel d  is  greater  t han or equal to 1000 mr em/hr will be provided with locked doors to prevent unauthorized entry.
For individual areas accessible to personnel that are lo cated within large areas, such as the Containment, where no enclosure exists for purpose of locking
For individual areas accessible to personnel that are lo cated within large areas, such as the Containment, where no enclosure exists for purpose of locking , and where no enclosure can be reasonably constructed, the individual area shall be barricaded, conspicuously posted and a flashing CALLAWAY - SA12.5-11Rev. OL-21 5/15 light activated as a warning de vice. The flashing light may be omitted if positive control of access to the area is provided.Very high radiation areas with radiation levels in excess of 500 rads in one hour at one meter from the radiation sour ce will be conspicuously posted as a very high radiation area. Entrances to these areas will be barr icaded and locked.
, and where no enclosure can be reasonably constructed, the individual area shall be barricaded, conspicuously posted and a flashing CALLAWAY - SA12.5-11Rev. OL-21 5/15light activated as a warning de vice. The flashing light may be omitted if positive control of access to the area is provided.Very high radiation areas with radiation levels in excess of 500 rads in one hour at one meter from the radiation sour ce will be conspicuously posted as a very high radiation area. Entrances to these areas will be barr icaded and locked.
Administrative measures for access cont rol in high radiat ion areas include the requirement for issuance of a radiation work permit for any operation to take place in such an area. Depending upon the operation, a radiation protection technician may be assigned to supervise stay-tim e and make appropriate surve ys while the operation is in progress.The radiation work permit (RWP) program will be established as an integral part of the ALARA policy implementation.
Administrative measures for access cont rol in high radiat ion areas include the requirement for issuance of a radiation work permit for any operation to take place in such an area. Depending upon the operation, a radiation protection technician may be assigned to supervise stay-tim e and make appropriate surve ys while the operation is in progress.The radiation work permit (RWP) program will be established as an integral part of the ALARA policy implementation.
This program will be the responsibility of the Manager, Radiation Protection. In addi tion to entries into and work performed in high radiation areas, RWP's may be required for the following type activities:a.Entries into designated airborne radi oactivity areas, r adiation areas, contamination areas and radioacti ve material storage areas. b.Work involving maintenance or other adjustments to any system or component which contains, stores, transports or collects radioactive materials.c.Other activities which in the judgement of Radiation Protection Supervision warrant the issuance of a radiation wo rk permit prior to initiating the operations.
This program will be the responsibility of the Manager, Radiation Protection. In addi tion to entries into and work performed in high radiation areas, RWP's may be required for the following type activities:a.Entries into designated airborne radi oactivity areas, r adiation areas, contamination areas and radioacti ve material storage areas. b.Work involving maintenance or other adjustments to any system or component which contains, stores, transports or collects radioactive materials.c.Other activities which in the judgement of Radiation Protection Supervision warrant the issuance of a radiation wo rk permit prior to initiating the operations.
The radiation work permit st ipulates the purpose of entry
The radiation work permit st ipulates the purpose of entry , work location, radiation conditions, surveillance and dosimetry requirements, stay-t ime, protective clothing and equipment, and other procedural requirement s and precautions.
, work location, radiation conditions, surveillance and dosimetry requirements, stay-t ime, protective clothing and equipment, and other procedural requirement s and precautions.
The primary objectives of the radiation wo rk permit program are to insure that:a.Radiological conditions are k nown as accurately as possible.b.Proper protective measures are taken to safely perform the r equired duties.c.Each person involved in these operations acknowledges his understanding of the radiation condition s, the protective and sa fety measures required, and his willingness to follow the radiation work permit requirements.d.Appropriate supervisors are aware of the task being performed, the radiation conditions, and the prescribed pr otective measures.
The primary objectives of the radiation wo rk permit program are to insure that:a.Radiological conditions are k nown as accurately as possible.b.Proper protective measures are taken to safely perform the r equired duties.c.Each person involved in these operations acknowledges his understanding of the radiation condition s, the protective and sa fety measures required, and his willingness to follow the radiation work permit requirements.d.Appropriate supervisors are aware of the task being performed, the radiation conditions, and the prescribed pr otective measures.
CALLAWAY - SA12.5-12Rev. OL-21 5/15e.A means for maintaining the account ability of personnel is provided.12.5.3.3Contamination ControlAreas that may be contaminat ed with radioactive material will be decontaminated to a level that is reasonably achi evable using available met hods and techniques. Since the complete removal of surface contamination from parts of the plant is a practical impossibility, certain plant areas may be designated as "contamination areas". These areas will be posted with the proper warni ng placards and barricaded.
CALLAWAY - SA12.5-12Rev. OL-21 5/15e.A means for maintaining the account ability of personnel is provided.12.5.3.3Contamination ControlAreas that may be contaminat ed with radioactive material will be decontaminated to a level that is reasonably achi evable using available met hods and techniques. Since the complete removal of surface contamination from parts of the plant is a practical impossibility, certain plant areas may be designated as "contamination areas". These areas will be posted with the proper warni ng placards and barricaded.
Line 468: Line 405:
CALLAWAY - SA12.5-13Rev. OL-21 5/1512.5.3.5Personnel Dosimetry12.5.3.5.1External Radiation Dosimetry See Section 12.5.2.1.8
CALLAWAY - SA12.5-13Rev. OL-21 5/1512.5.3.5Personnel Dosimetry12.5.3.5.1External Radiation Dosimetry See Section 12.5.2.1.8
.12.5.3.5.2Internal Ra diation Dosimetry See Section 12.5.2.1.9
.12.5.3.5.2Internal Ra diation Dosimetry See Section 12.5.2.1.9
.12.5.3.6Airborne Radiati on Evaluation and Control Airborne radioactivity will be routinely assessed using local sampling, portable continuous air monitors
.12.5.3.6Airborne Radiati on Evaluation and Control Airborne radioactivity will be routinely assessed using local sampling, portable continuous air monitors , and the fixed radiation monitoring system. Airborne radioactive materials (particulates, noble gases, halogens, tritium) will be sampled and analyzed using appropriate techniques. Since local sampling will provide better estimates of airborne contamination levels existing in a work area than wi ll a monitor reading, such special air sampling will be us ed in the radiation work per mit program to keep radiation exposure due to airborne radioactivity ALARA. Portable continuous air monitors and fixed airborne radioactivity monitors will be used to provide alarm indications and additional information which will be used wit h local sampling for the assessment of airborne radioactivity.
, and the fixed radiation monitoring system. Airborne radioactive materials (particulates, noble gases, halogens, tritium) will be sampled and analyzed using appropriate techniques. Since local sampling will provide better estimates of airborne contamination levels existing in a work area than wi ll a monitor reading, such special air sampling will be us ed in the radiation work per mit program to keep radiation exposure due to airborne radioactivity ALARA. Portable continuous air monitors and fixed airborne radioactivity monitors will be used to provide alarm indications and additional information which will be used wit h local sampling for the assessment of airborne radioactivity.
Control of airborne radioactivity levels wi ll be assured through the use of the plant's heating, ventilation, and air conditioning (HVAC) systems, portable air movers and filters. The HVAC systems provide contro lled air movement and filtra tion for those areas with a high potential for airborne radioactivity problems. Specia l control techniques can be used, such as plastic enclosu res which isolate and vent ai rborne radioactivity arising from special work projects. Respiratory pr otection equipment will be available for use in those situations where airbor ne radioactivity hazards exis t and other cont rol measures are inadequate at t he location and time.12.5.3.7Respiratory Protection ProgramThe respiratory protection program will be developed through the guidance of Regulatory Guide 8.15 and will satisfy the requirements of 10CFR20. As a minimum, the respiratory protection program will provide the following:a.Procedures to implement the selection, use, maintenance and storage of respiratory protection devices.b.Adequate facilities to support the storage, issue, cleaning and maintenance of respiratory protection devices.c.Use of only NIOSH certified or NRC approved equipment.
Control of airborne radioactivity levels wi ll be assured through the use of the plant's heating, ventilation, and air conditioning (HVAC) systems, portable air movers and filters. The HVAC systems provide contro lled air movement and filtra tion for those areas with a high potential for airborne radioactivity problems. Specia l control techniques can be used, such as plastic enclosu res which isolate and vent ai rborne radioactivity arising from special work projects. Respiratory pr otection equipment will be available for use in those situations where airbor ne radioactivity hazards exis t and other cont rol measures are inadequate at t he location and time.12.5.3.7Respiratory Protection ProgramThe respiratory protection program will be developed through the guidance of Regulatory Guide 8.15 and will satisfy the requirements of 10CFR20. As a minimum, the respiratory protection program will provide the following:a.Procedures to implement the selection, use, maintenance and storage of respiratory protection devices.b.Adequate facilities to support the storage, issue, cleaning and maintenance of respiratory protection devices.c.Use of only NIOSH certified or NRC approved equipment.
CALLAWAY - SA12.5-14Rev. OL-21 5/15d.Medical certification prog ram for respirator users.e.Periodic review of the overall effectiveness of the respiratory protection program via the inter nal dosimetry program.12.5.3.8Radioactive Materials Handling Methods and procedures will be developed to control, handle, and store by-product, source, and special nuclear material in accordance with regulatory requirements.Subjects to be covered by these procedures will incl ude the following:a.Storage of radioactive material in appropriately shielded, labeled and secured containers.b.Minimizing the distance that radi oactive samples are transported by personnel and the use of special ex tension and remote handling tools when applicable.c.Use of shielded sample transporters as appropriated.Periodic testing to verify the integrity of the sealed material.e.Emergency procedures which detail the proper actions to be taken in the event of leakage and spills.f.Accountability of any by-produ ct or special nuclear material.
CALLAWAY - SA12.5-14Rev. OL-21 5/15d.Medical certification prog ram for respirator users.e.Periodic review of the overall effectiveness of the respiratory protection program via the inter nal dosimetry program.12.5.3.8Radioactive Materials Handling Methods and procedures will be developed to control, handle, and store by-product, source, and special nuclear material in accordance with regulatory requirements.Subjects to be covered by these procedures will incl ude the following:a.Storage of radioactive material in appropriately shielded, labeled and secured containers.b.Minimizing the distance that radi oactive samples are transported by personnel and the use of special ex tension and remote handling tools when applicable.c.Use of shielded sample transporters as appropriated.Periodic testing to verify the integrity of the sealed material.e.Emergency procedures which detail the proper actions to be taken in the event of leakage and spills.f.Accountability of any by-produ ct or special nuclear material.
CALLAWAY - SA Rev. OL-18 12/10TABLE 12.5-1  RADIATION PROTEC TION LABORATORY EQUIPMENTINSTRUMENTRADIATIONDETECTEDDETECTORTYPICAL QUANTITYLOCATIONREMARKSGamma Spectroscopy SystemGammaHPGe w/graded shield4Counting Room and/or Maintenance Training Annex/
CALLAWAY - SA Rev. OL-18 12/10TABLE 12.5-1  RADIATION PROTEC TION LABORATORY EQUIPMENTINSTRUMENTRADIATIONDETECTEDDETECTOR TYPICAL QUANTITYLOCATIONREMARKSGamma Spectroscopy SystemGammaHPGe w/graded shield4Counting Room and/or Maintenance Training Annex/


Operations Support FacilityProvides Gamma Isotopic analysis capabilities Gas Proportional CounterAlpha, Beta, GammaGas Flow Proportional1Counting RoomUsed for counting smears, effluent and radio chemistry  
Operations Support FacilityProvides Gamma Isotopic analysis capabilities Gas Proportional CounterAlpha, Beta, Gamma Gas Flow Proportional1Counting RoomUsed for counting smears, effluent and radio chemistry  


samples.Liquid Scintillation CounterBeta---1Counting LabTritium determinationIn Vivo CounterGammaNaI (TI) and/or HPGe1In Vivo Count Room Central Processing FacilityPersonnel In Vivo Counting; MultidetectorsNote:Sensitivities of this analytical instrumentation are dependent upon counting parameters such as sample geometry, count time and background, however, equipm ent from various manufacturers will be evaluated to ensure that the equipment when purchased will be sufficiently sensitive to perform the tasks for which it was intended. Instruments will be selected to satisfy m easurement and reporting requirements in effect at the ti me the equipment is ordered.
samples.Liquid Scintillation CounterBeta---1Counting LabTritium determinationIn Vivo CounterGammaNaI (TI) and/or HPGe1In Vivo Count Room Central Processing FacilityPersonnel In Vivo Counting; MultidetectorsNote:Sensitivities of this analytical instrumentation are dependent upon counting parameters such as sample geometry, count time and background, however, equipm ent from various manufacturers will be evaluated to ensure that the equipment when purchased will be sufficiently sensitive to perform the tasks for which it was intended. Instruments will be selected to satisfy m easurement and reporting requirements in effect at the ti me the equipment is ordered.
CALLAWAY - SA Rev. OL-15 5/06TABLE 12.5-2  PORTABLE RADIATION PROTECTION EQUIPMENT TYPETYPICAL DETECTOR TYPE 1APPROXIMATE RANGEACCURACYTYPICAL QUANTITYLOCATIONPortable Count Rate Meter/
CALLAWAY - SA Rev. OL-15 5/06TABLE 12.5-2  PORTABLE RADIATION PROTECTION EQUIPMENT TYPE TYPICAL DETECTOR TYPE 1APPROXIMATE RANGE ACCURACY TYPICAL QUANTITYLOCATIONPortable Count Rate Meter/
FriskerGeiger-Mueller/ Scintillation 0-500,000 CPM 10% FS10See Note 3Low-Range Survey MeterGeiger-Mueller/
Frisker Geiger-Mueller/ Scintillation 0-500,000 CPM 10% FS10See Note 3Low-Range Survey MeterGeiger-Mueller/
IIonization Chamber 0-2  R/hr/
IIonization Chamber 0-2  R/hr/
0-5 R/hr 10% FS25 High-Range Extendable Probe Survey MeterGeiger-Mueller0-1000 R/hr 10% FS 4Mid-Range Survey MeterIGeiger-Mueller/
0-5 R/hr 10% FS25 High-Range Extendable Probe Survey MeterGeiger-Mueller0-1000 R/hr 10% FS 4Mid-Range Survey MeterIGeiger-Mueller/
IIonization Chamber 0-50 R/hr 0-100 R/hr 10% FS5High-Range Survey MeterIGeiger-Mueller/
IIonization Chamber 0-50 R/hr 0-100 R/hr 10% FS5High-Range Survey MeterIGeiger-Mueller/
IIonization Chamber 0-1000 R/hr IIndication 20% FS3Neutron Survey MeterBF  Tube Inside Polyethylene Sphere 0.2 mrem/hr 10rem/hr 10% FS4Ultra-High Range Survey Meter Geiger-Mueller/
IIonization Chamber 0-1000 R/hr IIndication 20% FS3Neutron Survey MeterBF  Tube Inside Polyethylene Sphere 0.2 mrem/hr 10rem/hr 10% FS4 Ultra-High Range Survey Meter Geiger-Mueller/
IIonization Chamber 0-10,000 R/hr 10% FS2Alpha Count Rate Meter/
IIonization Chamber 0-10,000 R/hr 10% FS2Alpha Count Rate Meter/
FriskerScintillation0-500,000 CPM 10% FS5Portal Monitor/Personnel  Contamination Monitor Gamma Scintillation/
FriskerScintillation0-500,000 CPM 10% FS5Portal Monitor/Personnel  Contamination Monitor Gamma Scintillation/
Gas Flow ProportionalVariable-----6Beta-Gamma Count Ratemeter/FriskerGeiger-Mueller0-500,000 CPM 10% FS20Electronic DosimeterSolid State0-100 R/hr/ 0-1000R 10% FS250 CALLAWAY - SATABLE 12.5-2 (Sheet 2)
Gas Flow ProportionalVariable-----6Beta-Gamma Count Ratemeter/FriskerGeiger-Mueller0-500,000 CPM 10% FS20Electronic DosimeterSolid State0-100 R/hr/ 0-1000R 10% FS250 CALLAWAY - SATABLE 12.5-2 (Sheet 2)
Rev. OL-15 5/06Electronic Dosimeter ReaderNot ApplicableNot ApplicableNot Applicable 8See Note 3Calibration Transfer InstrumentIIonization ChamberVarious 5%1Portable Area MonitorIGeiger-Mueller/
Rev. OL-15 5/06 Electronic Dosimeter ReaderNot ApplicableNot ApplicableNot Applicable 8See Note 3Calibration Transfer InstrumentIIonization ChamberVarious 5%1Portable Area MonitorIGeiger-Mueller/
Ionization Chamber/Solid State 0-2 R/hr 0-100 R/hr 25% Max.5Gamma Survey MeterNaI0-500k CPM 10%2High Volume Air SamplerNot Applicable> 15 CFM 20%  5Low Volume Air SamplerNot Applicable0-3 CFM 20%15Portable Continuous Air Monitor (3 channel type)NaI (I-131); Beta Scintillation (Particulate and  
Ionization Chamber/Solid State 0-2 R/hr 0-100 R/hr 25% Max.5Gamma Survey MeterNaI0-500k CPM 10%2High Volume Air SamplerNot Applicable> 15 CFM 20%  5Low Volume Air SamplerNot Applicable0-3 CFM 20%15Portable Continuous Air Monitor (3 channel type)NaI (I-131); Beta Scintillation (Particulate and  


Line 492: Line 428:


to 1E-1 (gaseous) 20% of Indication 4Note 1:The detector types listed are typical of the instrumentation used. As instrumentati on technology improves, an equivalent or better type of instrumentation may be substituted.Note 2:The quantities listed for each type of instrumentation ar e the typical numbers calibrated and ready for use.Note 3:Instrumentation is located in the RP Facilities on 1984' elevation of the Control Building, the Calibration Facilities, Dosimetry Processing Facilities or at specific work lo cations in the pl ant while in use.Note 4:The above information is referenc ed in the Radiological Em ergency Response Plan, Sect ion 6.7.3, and complies with Regulatory Guide 1.97.
to 1E-1 (gaseous) 20% of Indication 4Note 1:The detector types listed are typical of the instrumentation used. As instrumentati on technology improves, an equivalent or better type of instrumentation may be substituted.Note 2:The quantities listed for each type of instrumentation ar e the typical numbers calibrated and ready for use.Note 3:Instrumentation is located in the RP Facilities on 1984' elevation of the Control Building, the Calibration Facilities, Dosimetry Processing Facilities or at specific work lo cations in the pl ant while in use.Note 4:The above information is referenc ed in the Radiological Em ergency Response Plan, Sect ion 6.7.3, and complies with Regulatory Guide 1.97.
TYPETYPICAL DETECTOR TYPE 1APPROXIMATE RANGEACCURACYTYPICAL QUANTITYLOCATION}}
TYPE TYPICAL DETECTOR TYPE 1APPROXIMATE RANGE ACCURACY TYPICAL QUANTITYLOCATION}}

Revision as of 00:10, 8 July 2018

Callaway Plant, Unit 1 - Final Safety Analysis Report, Revision OL-22, Chapter 12, Radiation Protection
ML17061A157
Person / Time
Site: Callaway Ameren icon.png
Issue date: 04/05/2017
From: John Klos
Plant Licensing Branch IV
To:
Union Electric Co
Klos J
References
Download: ML17061A157 (143)
v • d • e


Text

CALLAWAY - SP12.0-iTABLE OF CONTENTSCHAPTER 12.0RADIATION PROTECTIONSectionPage12.1ENSURINGTHATOCCUPATIONALRADIATIONEXPOSURESARE ASLOWASREASONABLYACHIEVABLE.................................................12.1-112.1.1POLICY CONSIDERATIONS..................................................................12.1-112.1.2DESIGN CONSIDERATIONS..................................................................12.1-1 12.1.2.1PlantDesign.......................................................................................12.1-112.1.2.2Scale-ModelProgram.........................................................................12.1-2 12.1.2.3Second-LevelDesignReviewsby the Lead Architect-Engineer........12.1-312.1.2.4DesignReviewsbySNUPPS.............................................................12.1-312.1.2.5ExamplesofRadiationProtectionDesignReviews...........................12.1-4 12.1.2.6Decommissioning...............................................................................12.1-612.1.3OPERATIONAL CONSIDERATIONS......................................................12.1-6 12.

1.4REFERENCES

........................................................................................12.1-612.2RADIATIONSOURCES................................................................................12.2-112.2.1CONTAINED SOURCES.........................................................................12.2-1 12.2.1.1Containment.......................................................................................12.2-112.2.1.2AuxiliaryBuilding................................................................................12.2-2 12.2.1.3FuelBuilding.......................................................................................12.2-312.2.1.4TurbineBuilding.................................................................................12.2-412.2.1.5RadwasteBuilding..............................................................................12.2-5 12.2.1.6SourcesResultingfromDesignBasisAccidents...............................12.2-512.2.1.7StoredRadioactivity...........................................................................12.2-512.2.2AIRBORNE RADIOACTIVE MATERIAL SOURCES...............................12.2-612.2.2.1ModelForCalculatingAirborneConcentrations.................................12.2-7 12.

2.3REFERENCES

........................................................................................12.2-9 CALLAWAY - SPTABLE OF CONTENTS (Continued)SectionPage12.0-ii12.3RADIATIONPROTECTIONDESIGNFEATURES.......................................12.3-1 12.3.1FACILITY DESIGN FEATURES..............................................................12.3-1 12.3.1.1PlantDesignDescriptionforasLowasis ReasonablyAchievable(ALARA)...........................................................................12.3-112.3.1.2RadiationZoningandAccessControl................................................12.3-712.3.2SHIELDING.............................................................................................12.3-8 12.3.2.1DesignObjectives..............................................................................12.3-812.3.2.2GeneralShieldingDesign...................................................................12.3-9 12.3.2.3ShieldingCalculationalMethods......................................................12.3-1312.3.3VENTILATION.......................................................................................12.3-13 12.3.3.1DesignObjectives............................................................................12.3-1412.3.3.2DesignCriteria..................................................................................12.3-14 12.3.3.3DesignGuidelines............................................................................12.3-1412.3.3.4DesignDescription...........................................................................12.3-1612.3.3.5AirCleaningSystemDesign.............................................................12.3-1712.3.4AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION............................................................................12.3-1812.3.4.1AreaRadiationMonitoring................................................................12.3-1812.3.4.2AirborneRadioactivityMonitoring.....................................................12.3-2312.

3.5REFERENCES

......................................................................................12.3-3512.4DOSEASSESSMENT..................................................................................12.4-112.4.1EXPOSURES WITHIN THE PLANT........................................................12.4-1 12.4.1.1DirectRadiationDoseEstimates........................................................12.4-112.4.1.2AirborneRadioactivityDoseEstimates..............................................12.4-4 12.4.1.3IllustrativeExamplesofDoseAssessment........................................12.4-512.4.2EXPOSURES AT LOCATIONS OUTSIDE PLANT STRUCTURES........12.4-7 12.4.2.1DirectRadiationDoseEstimates........................................................12.4-712.4.2.2ExposuresDuetoAirborneRadioactivity...........................................12.4-8 CALLAWAY - SPTABLE OF CONTENTS (Continued)SectionPage12.0-iii12.

4.3REFERENCES

........................................................................................12.4-8 12.5RADIATION PROTECTION PROGRAM.....................................................12.5-1 CALLAWAY - SP12.0-ivRev. OL-1412/04LIST OF TABLESNumberTitle12.2-1Neutron Fluxes on Inside Surface of the Primary Shield Wall at the core Centerline (100% Power)12.2-2Gamma Fluxes on Inside Surface of the Primary Shield Wall at the Core Centerline (100% Power)12.2-3Pressurizer Shielding Source Terms12.2-4Spent Fuel Shutdown Sources (Full Core)12.2-5Radiation Sources Residual Heat Removal System 12.2-6Chemical and Volume Control System Sources12.2-7Fuel Storage Pool Water Activities12.2-8Secondary System Activities 12.2-9Conservative Basis Accumulated Radioactivity in the Gaseous Waste Processing System After Forty Years Operation12.2-10Deleted12.2-11Parameters and Assumptions for Calculating Airborne Radioactive Concentrations (Ref. 7)12.2-12Airborne Radioactivity Concentrations (mCi/cc)12.3-1List of Computer Codes used in Shielding Design Calculations 12.3-2Area Radiation Monitors12.3-3Inplant Airborne Radioactivity Monitors12.3-4Power Supplies for Area and In-Plant Airborne Monitors 12.4-1Illustrative Examples of Dose Assessment12.4-2Average Number of Personnel Per PWR Unit for the Period 1969-1977 (Ref. 1, 2, 3)

CALLAWAY - SPLIST OF TABLES (Continued)NumberTitle12.0-vRev. OL-1412/0412.4-3Average Occupational Radiation Exposure (Man-Rem Dose) Per PWR Unit for the Period 1969-1977 (Ref. 1, 2, 3)12.4-4Average Occupational Radiation Exposure (Man-Rem Dose) Based on PWR Plant Age (Ref 1, 2, 3)12.4-5Distribution of the Number of Personnel (>100 Milirem

ÚYr) According to Work Function12.4-6Distribution of Personnel (>100 Millirem/Yr) According to Employee Category12.4-7Percentages of Personnel Dose by Work Function (Ref3)12.4-8Annual Occupational Exposures for Various PWR Vendor's Units (REF. 1,2,3)12.4-9Average Annual Occupational Exposure and Power for Individual PWRS (REF 1, 2, 3)12.4-10Average Individual Exposure Based on PWR Plant Age (REF. 1,2,3) 12.4-11Cumulative Average of Annual Exposure by YearS of Operation - PWRS (REF. 1, 2, 3)12.4-12Estimates of Occupancy Times in Plant Radiation Areas and Gamma Doses to Plant Personnel12.4-13Distribution of Direct Radiation Man-rem Doses According to Work Functions12.4-14Annual Occupancy in Plant Areas Containing Airborne Radioactivity12.4-15Doses to plant Personnel Caused by Airborne Radioactivity CALLAWAY - SP12.0-viRev. OL-1412/04LIST OF FIGURESNumberTitle12.2-1Old Steam Generator Storage Facility General Layout12.3-1Typical Valve Compartment Arrangement12.3-2Radiation Zones for Normal Operation 12.3-3Control Room Isometric12.3-4Decontamination System12.4-1Cumulative Average of Annual Exposure by Years of Operation - PWRs (Table 12.4-11)

CALLAWAY - SP12.1-1Rev. OL-155/0612.1ENSURINGTHATOCCUPATIONALRADIATIONEXPOSURESARE ASLOWASREASONABLYACHIEVABLE12.1.1POLICY CONSIDERATIONSThe policy of Union Electric, through guidance from Regulatory Guides 1.8, 8.8, and 8.10 and 10 CFR Part 20, includes a commitment to ensure that occupational radiation exposures (ORE) are as low as reasonably achievable (ALARA). Union Electric will establish radiation protection practices for all applicable plant activities and will provide a qualified radiation protection organization to accomplish this goal. Utility management recognizes and will emphasize the importance of each individual's responsibilities to maintain ORE-ALARA. Section 12.1.1 of the Site Addendum identifies who has the ultimate responsibility for establishing an ALARA philosophy for operating, maintaining, refueling, and testing. The operating organizational structure and responsibilities are discussed in Section 13.1.2. Section 12.1.1 of the Site Addendum also indicates the responsibility of key personnel in implementation of ORE-ALARA policies. 12.1.2DESIGN CONSIDERATIONSA major objective in the design of the Callaway power block has been to limit the potential radiation exposures of operating personnel. Union Electric has emphasized this objective to its lead architect engineer (Bechtel) from the outset of the project, through its reviews of project design criteria and frequent design reviews as the detailed design of the power block has evolved. Specific design considerations and the guidelines employed in developing the Callaway design are presented in Sections 12.3.1.1 and 12.3.3.3. Estimated occupational radiation doses are given in Section 12.4. To ensure that ORE are ALARA, significant elements of the design program have been implemented as described in the following sections. 12.1.2.1PlantDesignThe design engineers and first level supervisors assigned by Bechtel to the Callaway project have, in most cases, performed similar design work on other nuclear power plants. Through this experience, they have developed sensitivity to and knowledge of radiation protection aspects of design, which have been applied to Callaway. Bechtel design engineers are also made aware of other operating experience through NRC Licensee Event Reports, NRC Operating Experience Bulletin Information Reports, and Bechtel-generated problem alert reports. The Bechtel designers have, where practical, followed the design guidelines of NRC Regulatory Guide 8.8. The designers have also followed recommendations of the NSSS supplier (Westinghouse). Many of these recommendations are available in documented form, in the Westinghouse information packages and in Reference1. Other CALLAWAY - SP12.1-2Rev. OL-155/06recommendations developed from discussions with Westinghouse. Westinghouse representatives met with Bechtel designers, as appropriate. Design of NSSS equipment, within Westinghouse's scope of supply, to ALARA objectives is described in Reference1. In the early stages of the plant design, calculations were performed to quantify potential concentrations of radioactivity throughout process systems and buildings of the power block. These calculations made use of measured data from operating plants, and also followed the methodology of NRC Regulatory Guide 1.112. Systems designs and equipment specifications have been influenced by these assessments (see Section12.1.2.5). Additionally, radiation dose rates have been estimated throughout the power block. Specifications for equipment and/or systems to be purchased reflect the objective to keep ORE-ALARA. For example, the specifications for the solid radwaste system state that "special consideration shall be given to eliminate points where radioactive materials may tend to accumulate." The specifications also require provisions for remote flushing and rinsing of portions of equipment that contain radioactive material and state that "special design consideration shall be given to minimizing operator exposure to radiation during the maintenance of equipment."12.1.2.2Scale-ModelProgramThe use of scale models is an innovative aspect of the Callaway project that has proven effective for design and design review. Three classes of scale models have been used with respect to radiation protection: (1) preliminary design model, (2) study models, and (3) final design model. The preliminary design model (3/8 inch = 1 foot) was, in effect, a three-dimensional layout tool. Model builders constructed the main elements of the buildings (walls, floors, columns, etc.) and set in place major equipment (tanks, pumps, motors, etc.). Plant design engineers then routed piping, electrical cable trays, and HVAC ducts, and located valves and valve actuators on the model. The three-dimensional aspects of the model, compared to conventional layout drawings, placed radiation protection, among other considerations, into sharp focus for both design engineers and reviewers and facilitated evaluation of design alternatives, such as choosing the best valve placement. Prior to completion of the preliminary design models, several reviews by utility and SNUPPS staff were held, as discussed further in Section 12.1.2.4. Upon completion of the preliminary design model, the approved layouts were committed to paper as design drawings. Study models were less formal in concept than the preliminary design and final design models and were constructed to evaluate specific design features or alternatives. For example, a study model was constructed of portions of the auxiliary building in order to evaluate the arrangement of radioactive demineralizers. The results of the review of this model are discussed in Section 12.1.2.5.

CALLAWAY - SP12.1-3Rev. OL-155/06The final design model (3/4 inch = 1 foot) was built by model makers from design drawings, but was constructed sufficiently early to permit design changes, e.g., to facilitate maintenance. The final design model comprises the reactor building, auxiliary building, control building, radwaste building, and turbine building. Additional design reviews by Bechtel and by SNUPPS were performed, using the final design model (see Sections 12.1.2.3 and 12.1.2.4). These reviews again focused on such factors as operability, maintainability, and radiation protection. 12.1.2.3Second-LevelDesignReviewsby the Lead Architect-EngineerThe term second-level review pertains to reviews beyond that of first-level supervision. On-project reviews are conducted by engineering and supervisory personnel from groups, other than the group that originated the design, and by higher level supervisory personnel. The reviews are generally interdisciplinary. For example, layout of shielding and valve operators in the radwaste building is performed by plant design personnel and is reviewed by mechanical/nuclear engineering personnel, as well as by the project engineers for plant design and systems. These reviews bring a broader base of experience to bear on all aspects of design, including radiation protection. Participants are professional engineers with from 5 to 20 years of nuclear power plant design experience. Off-project reviews are performed, when requested by the project organization, by members of the Bechtel technical staff. There have been no off-project reviews specifically addressed to radiation protection. However, shielding calculations have been the responsibility of the chief nuclear engineer and his staff and, through this involvement in the design, the staff personnel have contributed to the ALARA review. 12.1.2.4DesignReviewsbySNUPPSUtility reviews of all safety-related systems, structures, and equipment have been coordinated through the SNUPPS Technical Committee, which is composed of senior-level utility engineers, one from each SNUPPS utility. It is the responsibility of the Technical Committee to obtain comments from appropriate personnel within their companies and to bring those comments to meetings of the Committee, where decisions are made on the basis of discussion and eventually a vote. Throughout the design and construction phase of the SNUPPS project, the Technical Committee has met or had a telephone conference call on the average of once every week and a half, and total meeting days per year of the Technical Committee have averaged about 50. The SNUPPS technical director participates in Technical Committee meetings. Assisting the Technical Committee are various plant review groups, which are ad hoc groups of utility personnel selected to review and recommend action on specific aspects of design. The total meeting days per year have averaged about 15. A SNUPPS staff member participates in each meeting.

CALLAWAY - SP12.1-4Rev. OL-155/06Radiation protection has been an important aspect of the reviews by the Technical Committee and the plant review groups and has been the subject of numerous specific reviews, as discussed further in Section 12.1.2.5. At least six distinct reviews of the scale models have specifically included consideration of radiation protection during plant operation and maintenance. At these reviews, representatives of SNUPPS have included, in addition to Technical Committee members and SNUPPS staff, the health physics superintendant from Ginna Station, several licensed senior reactor operators, operations and maintenance supervisors from other nuclear power plants, and utility engineering personnel experienced in nuclear plant operation. Several design changes resulted from these reviews. Examples are described in Section12.1.2.5. During design and construction of the first SNUPPS unit, SNUPPS staff and qualified personnel from other SNUPPS utilities participated in a review of the effectiveness of design to the ALARA objectives. Periodically during construction, preoperational testing, and start-up of the SNUPPS units, qualified personnel from each SNUPPS utility will participate in ALARA reviews of their respective plants. 12.1.2.5ExamplesofRadiationProtectionDesignReviewsa.Radiation Zone DrawingsEvery location within the power block has been assigned a radiation zone classification. The method of establishing radiation zone classifications has been as follows. Shortly after initiation of the SNUPPS project, Bechtel prepared shielding design criteria which were reviewed by a SNUPPS plant review group. Participants in that review were utility engineering personnel, including one person with an SRO license from Ginna Station.

Subsequently, Bechtel prepared radiation zone drawings which define the zone classification of each location in the power block. These drawings were reviewed by the utilities and SNUPPS staff. Participants in those reviews included the Technical Committee, utility engineering personnel, utility operating and health physics personnel, and SNUPPS staff (technical director and civil/structural engineer). b.Reactor CavitySNUPPS learned in the Spring of 1975 that, unless neutron shielding was provided for the reactor cavity, neutron dose rates in the containment would be 10 to 100 times too high to permit operator access to the containment for reasonable periods of time during full-power operation. This conclusion was based on measured dose rates in the Calvert Cliffs plant, which is designed (as is SNUPPS) for access to the outside of the reactor vessel for performance of inservice inspection. SNUPPS had Bechtel undertake a study of possible neutron shield configurations, the effect of the shield on subcompartment pressure and pressure loadings on the reactor vessel, and obtainable dose reduction factors. SNUPPS has had numerous CALLAWAY - SP12.1-5Rev. OL-155/06design review meetings with Bechtel. Participants for SNUPPS have included: SNUPPS staff (executive director, technical director, and licensing manager), the Technical Committee, and utility engineering personnel. SNUPPS also contracted for an outside review by the NUS Corporation, which included independent estimates of neutron streaming and the effectiveness of neutron shielding materials. The design of the neutron shield in the containment is described in Sections 3.8.3.1.4 and 12.3.2.2.1. Subsequent to NRC approval of Callaway's revised GDC-4 submittal, a permanent cavity seal/neutron shield was installed to reduce neutron dose rates in containment.c.Radioactive DemineralizersAs a result of a review of a study model of the auxiliary building, the following design changes were made:1.A single wall between the corridor and the valve and piping compartments was designed to replace space-consuming overlapping (staggered for radiation protection) access walls. This allows more room in the corridor and more accessibility for maintenance within the compartments. 2.Vertical valve controls were designed to replace horizontal controls. This eliminates the need for 90-degree turns in the valve control fixtures and eliminates the access difficulty, which horizontal valve control rods pose as obstacles to maintenance. 3.A concrete shielding floor was provided above the valve and piping compartments to minimize exposure to the valve control operators. d.Airborne Radioactivity in ContainmentAirborne radioactivity (predominantly noble gases) has limited containment access in operating PWRs. Leakage of noble gases from the reactor coolant through the packing of the pressurizer spray valve has been determined to be a significant contributor to the gaseous activity in containment. To alleviate this situation, the following design provisions have been incorporated in the SNUPPS plants:1.A packless, low-leakage, ball-type pressurizer spray valve. 2.Provisions for continuous stripping of noble gases from the reactor coolant (see Section 9.3.4). 3.Addition of a mini-purge system to permit purging of the containment during power operation, prior to operator access (see Section 9.4.6).

CALLAWAY - SP12.1-6Rev. OL-155/06e.Steam Generator MaintenanceA final example of the results of scale-model reviews by SNUPPS is provision of permanent maintenance platforms below the steam generators to facilitate access and thereby reduce associated personnel radiation doses for maintenance operations, such as eddy-current inspection of tubes and sludge-lancing. 12.1.2.6DecommissioningThe following features of the plant design will assist decommissioning crews to maintain ORE-ALARA during the eventual decommissioning of Callaway. a.The building arrangements, compartmentation, corridors, doorways, and hatches provide the ability to remove most items of equipment intact or, alternatively, to isolate and entomb specific areas. b.The design features to maintain ORE-ALARA throughout the plant operating life are also applicable to the eventual decommissioning of the plant. These features include equipment design for ease of accessibility and maintenance, floor and wall coatings, provisions for remote flushing of equipment, ability to use remote handling equipment, decontamination techniques, and component design features to minimize crud buildup. c.Specifications and limitations on cobalt content in equipment components will serve to limit radiation doses from crud buildup during both operation and subsequent decommissioning. d.The amount of potentially radioactive buried pipe is limited. 12.1.3OPERATIONAL CONSIDERATIONSRefer to Section 12.1.1.3 of the Site Addendum for operational considerations. 12.

1.4REFERENCES

1."Design, Inspection, Operation, and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures As Low As Reasonably Achievable," WCAP - 8872, April 1977 CALLAWAY - SP12.2-1Rev. OL-2211/1612.2RADIATIONSOURCESThe sources of radiation that form the basis for shield design calculations and the sources of airborne radioactivity required for the design of personnel protective measures and for dose assessment are discussed and identified in this section.12.2.1CONTAINED SOURCESThe shielding design is based on full-power operation with 0.25 percent fuel cladding defects (Ref. 1, 2, 3, 4). The sources were obtained by multiplying the ANSI N237 fission product sources by two (Ref. 5). Sources in the primary coolant include fission products released from fuel clad defects and activation and corrosion products. The sources in the primary coolant are discussed in Section 11.1 and listed in Table 11.1-4. Throughout most of the primary coolant system, activation products, principally nitrogen-16 during reactor operation, are the primary radiation sources for shielding design. For all systems transporting radioactive materials, conservative allowance is made for transit decay, while, at the same time, providing for daughter product formation.In this section, the design sources are presented by building location and system. General location of the equipment discussed in this section is shown in the general arrangement drawings provided in Section1.2

.12.2.1.1Containment12.2.1.1.1Reactor Core The primary radiations within the containment during normal operation are neutrons and gamma rays emanating from the reactor core. Tables 12.2-1 and 12.2-2 list neutron and gamma multigroup fluxes at the core centerline location outside the reactor vessel. The tables are based on nuclear parameter values discussed in Chapter 4.0. Table 12.2-4 lists core gamma fluxes at the core centerline location outside the reactor vessel after shutdown, for shielding requirements during shutdown and inservice inspection. 12.2.1.1.2Reactor Coolant SystemSources of radiation in the reactor coolant system are fission products released from fuel and activation and corrosion products which are circulated in the reactor coolant. These sources are listed in Table 11.1-4, and their bases are discussed in Section11.1

.During operation, the activation product nitrogen-16 is the predominant activity in the reactor coolant pumps, steam generators, and reactor coolant piping. The contained source of radiation within the pressurizer is comprised of a liquid volume activity, a vapor volume activity, and a deposited activity. These activities are identified in Table12.2-3

.

CALLAWAY - SP12.2-2Rev. OL-2211/1612.2.1.1.3Secondary Coolant CycleUnder normal operating conditions, there is insignificant radioactive contamination present within the steam and power conversion system. It is possible to spread contamination to this cycle via steam generator tube leakage. Based on the primary-to-secondary leak rate given in Table 11.1A-1, the equilibrium secondary system activities are developed in Section ll.l and provided in Table 11.1-4. The condensate demineralizers and steam generator blowdown system further reduce the radioactivity level in the secondary cycle, as described in Section11.1. An evaluation of the secondary coolant activity in Tables 11.1-6 (Sheet 4) and 11.1-4 verifies that shielding is not required for the steam and power conversion system, with the exception of the components that could potentially concentrate the radioactivity. The condensate demineralizers and the steam generator blowdown demineralizers are the only components which could potentially concentrate the radioactivity. These are discussed in Sections 12.2.1.4 and 12.2.1.5, respectively. 12.2.1.1.4Auxiliary Systems Residual heat removal system-see Section 12.2.1.2.1

.Chemical and volume control system-see Section12.2.1.2.2. Steam generator blowdown system-see Section 12.2.1.5. 12.2.1.1.5Spent Fuel Handling and TransferThe spent fuel assemblies are the predominant source of radiation in the containment after plant shutdown for refueling. A reactor operating time necessary to establish fission product buildup near equilibrium for the reactor at rated power is used in determining the source strength. Shielding requirements for spent fuel transfer are based on the fission product activity present 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown to conservatively take credit for the time elapsed prior to the initiation of refueling operations. Source terms for spent fuel are listed in Table12.2-4. 12.2.1.2AuxiliaryBuilding12.2.1.2.1Residual Heat Removal SystemThe pumps, heat exchangers, and associated piping of the residual heat removal (RHR) system contain radioactive materials. For plant shutdown, the RHR pumps and heat exchanger sources result from the radioactive isotopes carried in the reactor coolant, discussed in Section 12.2.1.1.2, considering 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay following shutdown. The radiation source terms for the RHR system are listed in Table12.2-5.

CALLAWAY - SP12.2-3Rev. OL-2211/1612.2.1.2.2Chemical and Volume Control SystemThe CVCS source activity is the reactor coolant inventory which is provided in Table 11.1-4. More than 1minute of N-16 coolant activity decay is provided before the letdown line exits the containment, and, therefore, is not significant in determining shielding requirements for the CVCS equipment outside the containment.Major equipment items include the letdown and excess letdown heat exchangers, mixed bed and cation bed demineralizers, reactor coolant filter, volume control tank, and charging pumps. The seal water subsystem for the reactor coolant pumps includes the injection and return filters and the seal water heat exchanger. The design activities of the CVCS components are listed in Table 12.2-6. Heat exchanger and piping activities are derived from primary coolant activities. Radiation sources in the various pumps are assumed to be identical to the liquid sources in the tank from which the pump takes suction. 12.2.1.2.3Nuclear Sampling SystemThe major radiation sources in the nuclear sampling system originate from the RCS, RHR, and CVCS systems. The greatest radiation exposure would be to personnel taking the samples. To minimize this exposure, an integral shield has been incorporated into the sampling station design. Plant procedures provide contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere during the recovery phase following plant transient or accident.The sampling systems, which provide the capability to make required analyses are listed on Tables9.3-3 , 9.3-4 , 9.3-5, and 9.3-6. Figures 9.3-2 , 9.3-3, and 9.3-4 show the piping and instrumentation diagrams for the sampling systems. Process radioactivity monitors for the sampling systems are indicated in the tables and figures mentioned above. The area and airborne radioactivity monitors for worker protection are given in Section12.3.4 and are shown in Figure12.3-2

.12.2.1.3FuelBuilding12.2.1.3.1Spent Fuel Storage and Transfer The predominant radioactivity sources in the spent fuel storage and transfer areas in the fuel building are the spent fuel assemblies. Spent fuel assembly sources are discussed in Section12.2.1.1.5. For shielding design, the fuel storage pool assumptions are given in Section9.1.2. The major radionuclide concentrations in the water are provided in Table12.2-7

.The fuel transfer tube is completely shielded with permanent shielding to within radiation zone limits. No special access control, radiation monitoring, or posting is required. The CALLAWAY - SP12.2-4Rev. OL-2211/16expansion bellows for the fuel transfer tube are under water in the fuel transfer canal in the fuel building (see Figure3.8-48). There is no bellows inspection room or opening. The fuel transfer tube is completely surrounded by concrete or water, with the exception of the seismic gaps, so that no personnel access is possible. The fuel transfer tube in the seismic gap between the containment wall and the internal containment structure and in the seismic gap between the containment wall and the fuel building is shielded, using permanently installed lead loaded silicone foam rubber, to meet the radiation zone limits (see Figure12.3-2). Therefore, there is no unshielded portion of the fuel transfer tube.Shielding during dry fuel storage, transfer, and transport operations is provided by the HI-STORM UMAX Multi-Purpose Canister (MPC), the HI-TRAC VW transfer cask, the Vertical Ventilated Module (VVM), and the VVM closure lid. In addition to this shielding, the HI-STORM UMAX Certificate of Compliance (CoC) requires a minimum decay time of 3 years for fuel assemblies selected for dry fuel storage. 12.2.1.3.2Fuel Pool Cooling and Cleanup SystemSources in the fuel pool cooling and cleanup system (FPCCS) are the result of the transfer of radioactive isotopes from the reactor coolant into the fuel storage pool during refueling operations. The reactor coolant activities for fission, corrosion, and activation products (Table 11.1-4) are decayed for the amount of time required to remove the reactor vessel head following shutdown, are reduced by operation of the letdown system filters and demineralizers following shutdown, and are diluted by the total volumes of the water in the reactor vessel, refueling pool, and fuel storage pool. This activity then undergoes subsequent decay and accumulation on the FPCC filters and demineralizers.12.2.1.4TurbineBuilding12.2.1.4.1Main Steam Supply and Power Conversion SystemsPotential radioactivity in the main steam supply and power conversion systems is a result of steam generator tube leaks and fuel cladding defects, as discussed in Section12.2.1.1.3. This radioactivity is sufficiently low that no radiation shielding for equipment in the turbine building is required in order to meet the radiation zone requirements. The isotopic concentrations for a condensate demineralizer bed and other secondary system sources are listed in Table12.2-8.

CALLAWAY - SP12.2-5Rev. OL-2211/1612.2.1.5RadwasteBuilding12.2.1.5.1Boron Recycle and Liquid Radwaste SystemsThe system sources are radioisotopes, including fission and activation products, present in the reactor coolant. The components of the systems contain varying degrees of activity, depending on the detailed system and equipment design. The concentrations of radionuclides in the process fluids at various locations in the radwaste systems, such as pipes, tanks, filters and demineralizers are discussed in Section 11.1 and are listed in Tables 11.1-4 and 11.1-6. These nuclide concentrations for 0.25 percent failed fuel have been used in the final shielding design. Shielding for each component of the radwaste systems is based on maximum activity conditions, as given in Sections11.1 and 11.2. 12.2.1.5.2Gaseous Radwaste SystemRadiation sources for each component of the waste gas system are based on operation under the conditions given in Sections11.1 and 11.3. Tabulation of the activities is shown in Table12.2-9. 12.2.1.5.3Solid Radwaste SystemRadiation sources for each component providing influent to the solid radwaste system are based on operation under the conditions given in Sections 11.1 and 11.4. Tabulation of the activities is shown in Table11.4-3. 12.2.1.6SourcesResultingfromDesignBasisAccidentsThe radiation sources from design basis accidents include the design basis inventory of radioactive isotopes in the reactor coolant, plus postulated fission product releases from the fuel. Accident parameters and sources are discussed and evaluated in Chapters11.0 and 15.0. 12.2.1.7StoredRadioactivityThe principal sources of activity not stored inside the plant structures are the original steam generators, the original reactor vessel closure head, spent fuel assemblies stored within the Independent Spent Fuel Storage Installation (ISFSI), the Forced Helium Dehydration (FHD) skid stored within the ISFSI Support Building, the reactor makeup water storage tank (RMWST), the refueling water storage tank (RWST), the condensate storage tank (CST), and two radwaste discharge monitor tanks. The original steam generators and original reactor vessel closure head are stored in the Old Steam Generator Storage Facility (OSGSF). As shown in Figure 12.2-1, the OSGF contains the following 3 rooms or areas: the Old Steam Generator (OSG) bay, the Old CALLAWAY - SP12.2-6Rev. OL-2211/16Reactor Vessel Head (ORVH) room, and miscellaneous radioactive materials (RAM) storage room. Contaminated materials in the RAM storage room are limited to those not exceeding the threshold of Category 2 radioactive materials as defined in 10 CFR 37.The walls and ceiling of the OSGSF are constructed of reinforced concrete with a minimum thickness of 2 feet. Shielding calculations that conservatively assume 100% of the activity in the OSGSF is from high energy Co-60 have been performed to confirm the external dose rate will not exceed 2 mrem/hr for the OSGSF. The OSGSF is expected to yield an external surface dose rate of 0.6 mrem/hr or less. Personnel access to the OSGSF is controlled by two labyrinth entryways with locked steel doors. One entryway leads into the OSG bay, and one entryway leads into the RAM storage room. A locked access door is also provided between the OSG bay and ORVH room. Additionally, a lockable steel rollup door is also provided for exterior access to the RAM storage room. The OSG bay and ORVH room are treated as confined spaces due to emergency egress requirements associated with Building Officials and Code Administrators (BOCA) National Building Code, 1999. Personnel access to the facility is infrequent. The annual dose to an individual at the site boundary due to normal operation of the ISFSI is limited to the values specified in 10 CFR 72.104. The RMWST and CST are expected to contain concentrations of radionuclides which yield a surface dose rate of 0.5 mrem/hr or less. The RWST is expected to have a maximum contact dose of less than 10mrem/hr when the water is returned from the refueling pool. This will be rapidly reduced by processing through the fuel storage pool purification filter and demineralizer. The radwaste discharge monitor tanks will also have a dose rate of less than 10mrem/hr at contact. Tabulations of the activities within these tanks are provided in Table11.1-6 (Sheets1, 2, and 3). All spent fuel is stored in the fuel storage pool until it is placed in either a spent fuel storage canister for storage in the ISFSI, or a spent fuel shipping cask for transport offsite. Storage space is allocated in the radwaste building for the storage of spent filter cartridges and dewatered spent resins, and chemical/hazardous wastes. Radioactive wastes stored inside the plant structures are shielded so that there is Zone A access outside the structure (see Figure 12.3-2). If it becomes necessary to temporarily store radioactive wastes outside the plant structures, adequate radiation protection measures will be taken by the radiation protection staff. 12.2.2AIRBORNE RADIOACTIVE MATERIAL SOURCESThis section identifies the models, parameters, and sources required to evaluate airborne radionuclide concentrations during plant operations in various plant radiation areas where personnel occupancies are expected. An analysis of operating plant measurements of average airborne radionuclide concentrations (Ref. 6) and their respective MPCS for various stations in the auxiliary building (these stations include the waste handling area and the waste gas decay tank rooms, which are located in the radwaste building) indicates that the concentration to CALLAWAY - SP12.2-7Rev. OL-2211/16MPC ratios in all stations is well within the limits of Note5 of AppendixB to 10 CFR 20.1-20.601 to consider these airborne activities as insignificant. It is expected that the SNUPPS units would not have airborne radioactivity concentrations significantly greater than the operating plant data for corresponding locations. However, it is possible that, within these stations, there may be rooms where maximum airborne concentrations can occur due to localized leakage of radioactive fluids, but these rooms house equipment and components that handle highly radioactive fluids and consequently are D or E zones and, therefore, will be normally inaccessible. Also the ventilation systems in the auxiliary and radwaste buildings are designed in such a manner that airborne contamination from high radiation zones will not generally spread into low radiation zones, since the airflow is from regions of lower potential for contamination to those with higher potential for contamination. Consequently, negligible airborne radioactivity concentrations are expected in those areas of the auxiliary and radwaste buildings which are accessible. Airborne radioactivity concentrations in the turbine building are also expected to be negligible, since possible leaks into the turbine building are only from the secondary side, and, also, the turbine building ventilation exhaust is high (at least 90,000 cfm). For example, airborne concentrations are calculated to be 2.3 x 10

-3 µCi/cc and 6.9x10

-12 µCi/cc for I-131 and Xe-133, respectively in the turbine building. Higher airborne concentrations can, however, occur in the containment, both during power operation and refueling--the former due to coolant leakage and the latter primarily due to the evaporation of the refueling pool. Likewise, airborne concentrations can also occur in the fuel building both during power operation and refueling due to the evaporation of the spent fuel pool. During power operation, the airborne radioactivity in the fuel building will be almost all due to tritium, since the continuous operation of the fuel pool cleanup system will effectively remove other isotopes from the pool. The assumptions and parameters required to evaluate the airborne radionuclide concentrations in the containment and fuel building both during power operation and refueling are listed in Table12.2-11. The concentrations in these buildings are listed in Table12.2-12. Even though some of these airborne concentrations may be high, limited occupancies in these areas ensures that the doses from airborne radioactivity to an individual will be a small fraction of the 10 CFR 20 limits for occupation exposures. Airborne radioactivity is monitored inside the plant, as described in Section12.3.4, and in process equipment and effluents, as described in Section11.5. 12.2.2.1ModelForCalculatingAirborneConcentrationsFor those regions which are characterized by a constant leakrate of the radioactive source at constant source strength and a constant exhaust rate of the contaminant, the CALLAWAY - SP12.2-8Rev. OL-2211/16peak or equilibrium airborne concentration of the radioisotope in the regions can be calculated, using the following equation:whereandFrom the above equation, it is evident that the peak or equilibrium concentration, CEqi , of the i th radioisotope in the applicable region will be given by the following expression:With high exhaust rates, this peak concentration will be reached within a few hours. (1) (LR)i=Leak or evaporation rate of the i th radioisotope in gm/sec, in the applicable region A i=activity concentration of the i th leaking or evaporating radioisotope in

µCI/gm(PF)i=partition factor or the fraction of the leaking activity that is airborne for the i th radioisotope Ti=Total removal rate constant for the i th radioisotope in sec

-1 from the applicable region

=(di + e)di + e are the removal rate constants in sec

-1 due to radioactive decay for the i th radioisotope and the exhaust from the applicable regiont=time interval between the start of the leak and the time at which the concentration is evaluated in secondsV=free volume of the region in which the leak occurs in cc C i(t)=airborne concentration of the i th radioisotope at time t in mCi/cm3 in the applicable region CEqi = (LR)i A i (PF)i /VTi(2)Cit ()LR ()iAiPF ()i 1eTi t--()VTi--------------------------------------------------------------

=

CALLAWAY - SP12.2-9Rev. OL-2211/1612.

2.3REFERENCES

1.C. M. Lederer, et. al., TableofIsotopes, Lawrence Radiation Laboratory, University of California (March, 1968). 2.ReactorPhysicsConstants, Argonne National Laboratory, ANL-5800 (July, 1963).3.H. Soodak, ReactorHandbook, Vol. III, Part A, Physics, second edition (1962). 4.D. A. Klopp, NAP-Multigroup Time-dependent Neutron Activation Prediction Code, IITRI-A6088-21 (January, 1966), conditions as given in Sections 11.1 , 11.2 , and 11.4. 5.ANSI N 237, "Source Term Specification," Final Draft, 1977. 6.NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station - Unit 1, Prepared for USNRC by EG & Idaho, Inc., July 1978.7.NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous & Liquid Effluents from Pressurized Water Reactors, USNRC, April 1976.

CALLAWAY - SPRev. OL-135/03TABLE 12.2-1 NEUTRON FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)Energy GroupNeutron Flux(neutrons/cm 2-sec)1(E > 1.0 Mev)7.6 x 10 81(5.53 Kev < E 1.0 Mev)1.2 x 10 103(0.625 ev E 5.53 Kev)7.1 x 10 94(E < 0.625 ev)1.8 x 10 9 CALLAWAY - SPRev. OL-135/03TABLE 12.2-2 GAMMA FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)GroupFlux (Mev/cm 2-sec)GroupEnergy(Mev/)1 3.7 x 10 9 7.52 3.3 x 10 9 4.03 1.7 x 10 9 2.54 1.0 x 10 9 0.8 CALLAWAY - SPRev. OL-135/03TABLE 12.2-3 PRESSURIZER SHIELDING SOURCE TERMSLiquidVolumeIsotopeActivity

(µCi/gm)N-161.8 (max)Cr-5l1.68E-03Mn-543.10E-04Fe-551.61E-03Fe-599.30E-04Co-581.53E-02Co-602.02E-03Br-832.53E-04Br-843.08E-05Br-853.39E-07Rb-861.48E-04Rb-881.32E-03Sr-896.84E-04Sr-902.11E-05Sr-911.29E-04Y-89M6.16E-08Y-901.35E-05 Y-91M8.44E-05Y-911.33E-04Y-937.05E-06Zr-955.72E-05Nb-954.84E-05Nb-95M2.92E-05Mu-997.34E-02Tc-993.68E-09Ru-1034.14E-05Ru-1061.00E-05Te-125M2.75E-05Te-127M2.73E-04Te-1273.26E-04Te-129M1.27E-03Te-1298.23E-04Te-131M6.13E-04Te-1311.15E-04Te-1321.24E-02I-1298.55E-13I-1305.17E-04I-1313.83E-01I-1321.71E-02I-1331.46E-01I-1349.09E-04I-1352.68E-02Cs-1345.24E-02Cs-1351.04E-10Cs-1362.10E-02Cs-1373.79E-02Ba-137M3.58E-02Ba-1403.55E-04La-1403.43E-04Ce-1411.32E-04Ce-1436.86E-05Ce-1446.86E-05Pr-1438.67E-05 Pr-1446.86E-05SteamVolumeIsotopeActivity

(µCi/gm)Kr-83M9.74E-04Kr-85M1.14E-02Kr-851.13E+00Kr-871.92E-03Kr-881.36E-02Kr-896.89E-06Xe-131M1.31E-01Xe-133M1.35E-01Xe-1331.59E+01Xe-135M9.03E-05Xe-1356.79E-02Xe-1371.49E-05Xe-1382.69E-04LiquidVolumeIsotopeActivity

(µCi/gm)

CALLAWAY - SPRev. OL-135/03TABLE 12.2-4 SPENT FUEL SHUTDOWN SOURCES (FULL CORE)Photon EnergyTime After Shutdown(Mev)4 Hours12 Hours1 Day1 Week1 Month3 Months0.43.1 x 10 112.3 x 10 111.9 x 10 119.2 x 10 103.8 x 10 101.3 x 10 100.81.3 x 10 129.8 x 10 118.0 x 10 114.0 x 10 112.3 x 10 111.2 x 10 111.33.9 x 10 112.9 x 10 112.5 x 10 111.6 x 10 111.2 x 10 115.8 x 10 101.75.1 x 10 113.8 x 10 113.3 x 10 112.3 x 10 116.2 x 10 102.9 x 10 92.27.2 x 10 102.6 x 10 101.5 x 10 108.5 x 10 96.7 x 10 95.0 x 10 92.58.9 x 10 104.7 x 10 103.7 x 10 102.5 x 10 107.9 x 10 93.5 x 10 83.58.2 x 10 92.0 x 10 91.3 x 10 99.6 x 10 82.0 x 10 81.5 x 10 7Values given in Mev/cc-sec.

CALLAWAY - SPRev. OL-135/03TABLE 12.2-5 RADIATION SOURCES RESIDUAL HEAT REMOVAL SYSTEMIsotopeActivities(µCi/gm)Cr-511.51E-03Mn-542.48E-04Fe-551.28E-03 Fe-597.98E-04Co-581.28E-02Co-601.60E-03 Br-832.49E-03Br-842.28E-05Br-85negligible Kr-83M1.34E-02Kr-85M1.20E-01Kr-851.68E-02 Kr-871.48E-02Kr-881.57E-01Kr-89negligible Rb-861.39E-04 Rb-881.75E-01Rb-895.28E-08Sr-895.73E-04 Sr-901.64E-05Sr-918.01E-04Y-89M5.15E-08 Y-902.58E-06Y-91M5.30E-04Y-911.06E-04 Y-934.25E-05Zr-954.79E-05Nb-954.00E-05 Nb-95M1.45E-06Mo-991.32E-01Tc-992.00E-10 Ru-1033.59E-05Ru-1068.00E-06Te-125M2.32E-05Te-127M2.24E-04Te-1275.63E-04Te-129M1.12E-03 Te-1297.66E-04Te-131M1.82E-03Te-1313.34E-04 Te-1322.08E-02I-1292.63E-14I-1302.75E-03 I-1314.37E-01I-1326.36E-02I-1335.46E-01 I-1343.17E-03I-1352.06E-01Xe-131M3.94E-02 Xe-133M2.07E-01 Xe-1331.06E+01Xe-135M3.56E-02Xe-1355.42E-01 Xe-137negligibleXe-1388.11E-07Cs-1344.10E-02 Cs-1354.48E-13Cs-1362.11E-02Cs-1372.95E-02 Cs-1384.47E-04Ba-137M2.79E-02Ba-1403.58E-04 La-1402.54E-04Ce-1411.14E-04Ce-1436.03E-05 Ce-1445.41E-05Pr-1438.18E-05Pr-1445.41E-05IsotopeActivities(µCi/gm)

CALLAWAY - SPTABLE 12.2-5 (Sheet 2)Rev. OL-135/03I-1314.37E-01I-1326.36E-02 I-1335.46E-01I-1343.17E-03I-1352.06E-01 XE-131M3.94E-02XE-133M2.07E-01XE-1331.06E+01 XE-135M3.56E-02XE-1355.42E-01XE-1372.78E-21 XE-1388.11E-07CS-1344.10E-02CS-1354.48E-13 CS-1362.11E-02CS-1372.95E-02CS-1384.47E-04 BA-137M2.79E-02BA-1403.58E-04LA-1402.54E-04 CE-1411.14E-04CE-1436.03E-05CE-1445.41E-05 PR-1438.18E-05PR-1445.41E-05IsotopeActivities(µCi/gm)

CALLAWAY - SPRev. OL-135/03TABLE 12.2-6 CHEMICAL AND VOLUME CONTROL SYSTEM SOURCESLETDOWN MIXED BED DEMINERALIZERIsotopeActivity (µCi/cc)Cr-513.30E+01Mn-543.36E+01Fe-552.23E+02Fe-592.80E+01Co-586.94E+02Co-602.96E+02Br-836.26E-01Br-847.47E-02Br-858.17E-04Rb-861.15E+00Rb-881.78E+00Sr-892.30E+01Sr-903.25E+00Sr-913.42E-01Y-89M2.07E-03Y-903.21E+00 Y-91M2.25E-01Y-915.13E+00Y-931.88E-02Zr-952.40E+00Nb-953.44E+00Nb-95M2.40E+00Mo-993.02E+02Tc-999.73E-04Ru-1031.12E+00Ru-1061.14E+00Te-125M1.04E+00Te-127M1.72E+01Te-1271.74E+01Te-129M2.99E+01Te-1291.92E+01Te-131M1.95E+00Te-1313.64E-01Te-1325.49E+01I-1291.12E-06 CALLAWAY - SPTABLE 12.2-6 (Sheet 2)Rev. OL-135/03Reactor Coolant FilterNOTE:All other demineralizers and filters throughout the plant are shielded with these same source terms, since these are the most radioactive.I-1301.41E+00I-1312.83E+03I-1326.73E+01I-1334.30E+02I-1342.21E+00I-1356.92E+01Cs-1343.88E+03Cs-1351.73E-05Cs-1361.22E+02Cs-1373.26E+03Ba-137M3.08E+03Ba-1403.66E+00La-1403.98E+00Ce-1412.96E+00Ce-1437.16E-02Ce-1447.20E+00Pr-1439.56E-01P4-1447.20E+00Specific SourceGamma Energy Strength(Mev/) (Mev/cc-sec) 0.8 5.7 x 10 7 1.3 1.5 x 10 7IsotopeActivity (µCi/cc)

CALLAWAY - SPTABLE 12.2-6 (Sheet 3)Rev. OL-135/03RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities

(µCi/gm)IsotopeActivities

(µCi/cc)IsotopeActivity (µCi/cc)Cr-511.90E-03Cr-511.73E-04Kr-83M2.69E-01Mn-543.10E-04Mn-542.82E-05Kr-85M2.01E+00Fe-551.60E-03Fe-551.46E-04Kr-852.40E-01 Fe-591.00E-03Fe-599.10E-05Kr-876.11E-01Co-581.60E-02Co-581.46E-03Kr-883.11E+00Co-602.00E-03Co-601.82E-04Kr-893.19E-03Br-831.00E-02Br-831.00E-03Xe-131M5.65E-01Br-845.42E-03Br-845.42E-04Xe-133M2.96E+00Br-856.25E-04Br-856.25E-05Xe-1331.52E+02Kr-83M4.54E-02Rb-868.08E-05Xe-135M3.88E-02Kr-85M2.25E-01Rb-881.90E-01Xe-1357.16E+00Kr-851.67E-02Sr-896.65E-05Xe-1376.88E-03Kr-871.32E-01Sr-901.90E-06Xe-1381.17E-01Kr-884.23E-01Sr-911.24E-04 Kr-891.13E-02Y-902.28E-07Rb-861.77E-04Y-91M6.84E-05Rb-884.17E-01Y-911.22E-05Sr-897.29E-04Y-936.46E-06Sr-902.08E-05Zr-955.46E-06Sr-911.35E-03Nb-954.55E-06Y-902.50E-06Mo-991.60E-02Y-91M7.50E-04Ru-1034.10E-06Y-911.33E-04Ru-1069.10E-07 CALLAWAY - SPTABLE 12.2-6 (Sheet 4)Rev. OL-135/03Y-937.08E-05Te-125M2.64E-06----------Zr-956.00E-05Te-127M2.55E-05Nb-955.00E-05Te-1277.74E-05Mo-991.75E-01Te-129M1.27E-04Ru-1034.50E-05Te-1291.46E-04 Ru-1061.00E-05Te-131M2.28E-04Te-125M2.90E-05Te-1311.00E-04 Te-127M2.80E-04Te-1322.46E-03Te-1278.50E-04I-1304.37E-04Te-129M1.40E-03I-1315.62E-02Te-1291.60E-03I-1322.08E-02Te-131M2.50E-03I-1337.92E-02Te-1311.10E-03I-1349.79E-03 Te-1322.70E-02I-1353.96E-02I-1304.37E-03Cs-1342.38E-02 I-1315.62E-01Cs-1361.24E-02 I-1322.08E-01Cs-1371.71E-02 I-1337.92E-01Ba-137M3.04E-03 I-1349.79E-02Ba-1404.18E-05 I-1353.96E-01La-1402.85E-05Xe-131M3.98E-02Ce-1411.33E-05 Xe-133M2.17E-01Ce-1437.60E-06Xe-1331.08E+01Ce-1446.27E-06Xe-135M3.00E-02Pr-1439.50E-06Xe-1356.44E-01Pr-1446.27E-06 Xe-1372.04E-02RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities

(µCi/gm)IsotopeActivities

(µCi/cc)IsotopeActivity (µCi/cc)

CALLAWAY - SPTABLE 12.2-6 (Sheet 5)Rev. OL-135/03Xe-1389.89E-02--------------------Cs-1345.21E-02Cs-1362.71E-02NOTE :

Cs-1373.75E-02Ba-137M3.33E-02The moderating heat exchanger, chiller heat exchanger, and letdown reheat heat exchanger are the same as the combined liquid and gaseous sources for the volume control tank.Ba-1404.58E-04La-1403.12E-04Ce-1411.46E-04Ce-1438.33E-05Ce-1446.87E-05Pr-1431.04E-04 Pr-1446.87E-05RegenerativeHeat ExchangerShell Side andExcess Letdownand Letdown HeatExchanger Tube SideVolume Control Tank Liquid Volumeand Seal Water HeatExchangerVolume Control TankGaseous VolumeIsotopeActivities

(µCi/gm)IsotopeActivities

(µCi/cc)IsotopeActivity (µCi/cc)

CALLAWAY - SPTABLE 12.2-6 (Sheet 6)Rev. OL-135/03Boric Acid TanksIsotopeActivity (µCi/cc)Cr-514.16E-06Mn-548.56E-07Fe-554.49E-06Fe-592.41E-06Co-584.08E-05Co-605.63E-06Br-832.54E-08Br-841.92E-12Br-85negligibleRb-868.40E-06Rb-881.87E-09Rb-895.94E-14Sr-897.32E-06Sr-905.90E-08Sr-916.39E-08Y-89M6.58E-10Y-905.00E-08Y-91M4.27E-08Y-913.54E-07 Y-933.61E-09Zr-951.52E-07Nb-951.35E-07Nb-95M1.18E-07Mo-998.72E-05Tc-991.45E-11Ru-1031.06E-07Ru-1062.77E-08Te-125M7.23E-08Te-127M7.40E-07Te-1277.68E-07Te-129M3.21E-06Te-1292.05E-06Te-131M5.16E-07Te-1319.41E-08Te-1321.59E-05I-1295.08E-15I-1303.06E-07 CALLAWAY - SPTABLE 12.2-6 (Sheet 7)Rev. OL-135/03I-1317.46E-04I-1321.67E-05I-1331.09E-04I-1342.26E-09I-1351.06E-05Cs-1343.56E-03Cs-1353.71E-09Cs-1361.11E-03Cs-1372.59E-03Cs-1387.90E-08Ba-137M2.45E-03Ba-1407.64E-07La-1408.17E-07Ce-1413.31E-07Ce-1431.92E-08Ce-1441.90E-07Pr-1431.93E-07Pr-1441.90E-07Boric Acid TanksIsotopeActivity (µCi/cc)

CALLAWAY - SPRev. OL-135/03TABLE 12.2-7 FUEL STORAGE POOL WATER ACTIVITIESIsotope**Other isotopes will be present in much lower concentrations.Activities (µCi/cc)Co-58 1.3E-04 Co-60 4.0E-04 Cs-134 2.0E-05 Cs-137 2.0E-04 CALLAWAY - SPRev. OL-135/03TABLE 12.2-8 SECONDARY SYSTEM ACTIVITIESCondensate DemineralizerHigh TDS Regenerant Collector TankSecondary Liquid Waste Drain Collector TankSecondary Liquid Waste Monitor TankIsotopeActivity (µCi/cc)IsotopeActivity (µCi/cc)IsotopeActivity (µCi/cc)IsotopeActivity (µCi/cc)Cr-511.85E-06Cr-512.04E-07Cr-513.98E-11Cr-512.38E-12Mn-544.96E-07Mn-546.43E-08Mn-548.99E-12Mn-546.43E-13Fe-552.01E-06Fe-552.61E-07Fe-553.60E-11Fe-552.61E-12 Fe-591.33E-06Fe-591.73E-07Fe-592.67E-11Fe-591.72E-12Co-581.86E-05Co-582.41E-06Co-583.57E-10Co-582.41E-11Co-602.27E-06Co-602.94E-07Co-604.05E-11Co-602.94E-12 Br-834.96E-07Br-836.43E-08Br-838.83E-11Br-835.83E-13Br-842.94E-08Br-843.81E-09Br-845.23E-12Br-847.03E-18Br-854.27E-11Br-855.54E-12Br-857.62E-15Br-856.31E-66 Rb-861.52E-07Rb-861.96E-08Rb-863.95E-12Rb-861.94E-13Rb-887.73E-08Rb-881.00E-08Rb-881.53E-11Rb-883.08E-22Sr-899.40E-07Sr-891.22E-07Sr-891.86E-11Sr-891.21E-12 Sr-902.10E-08Sr-902.73E-09Sr-903.75E-13Sr-902.73E-14Sr-914.21E-08Sr-915.46E-09Sr-917.13E-12Sr-913.01E-14Y-89M8.46E-11Y-89M1.10E-11Y-89M1.67E-15Y-89M1.09E-16 Y-901.73E-08Y-902.25E-09Y-901.36E-13Y-902.29E-14Y-91M2.86E-08Y-91M3.70E-09Y-91M4.84E-12Y-91M2.01E-14Y-911.50E-07Y-911.95E-08Y-912.90E-12Y-911.94E-13 Y-932.19E-09Y-932.84E-10Y-933.68E-13Y-931.61E-15Zr-959.26E-08Zr-951.20E-08Zr-951.79E-12Zr-951.20E-13Nb-959.43E-08Nb-951.22E-08Nb-951.78E-12Nb-951.23E-13 Nb-95M6.62E-08Nb-95M8.59E-09Nb-95M2.61E-13Nb-95M8.80E-14Nb-994.89E-05Nb-996.35E-06Nb-993.14E-09Nb-995.82E-11Tc-995.99E-12Tc-997.77E-13Tc-992.59E-17Tc-997.96E-18 Ru-1034.37E-08Ru-1035.66E-09Ru-1038.89E-13Ru-1035.63E-14Ru-1069.93E-09Ru-1061.29E-09Ru-1061.79E-13Ru-1061.29E-14Te-125M2.29E-08Te-125M2.96E-09Te-125M4.47E-13Te-125M2.95E-14 Te-127M2.39E-07Te-127M3.10E-08Te-127M4.48E-12Te-127M3.10E-13Te-1272.61E-07Te-1273.39E-08Te-1278.06E-12Te-1273.25E-13 CALLAWAY - SPTABLE 12.2-8 (Sheet 2)Rev. OL-135/03Te-129M1.28E-06Te-129M1.66E-07Te-129M2.66E-11Te-129M1.65E-12Te-1298.26E-07Te-1291.07E-07Te-1291.83E-11Te-1291.06E-12Te-131M2.78E-07Te-131M3.61E-08Te-131M3.08E-11Te-131M2.98E-13Te-1315.17E-08Te-1316.70E-09Te-1315.77E-12Te-1315.44E-14 Te-1326.79E-06Te-1328.81E-07Te-1323.86E-10Te-1328.18E-12I-1291.44E-15I-1291.87E-16I-1302.34E-10I-1291.88E-20I-1301.45E-06I-1301.88E-07I-1317.85E-08I-1301.18E-11 I-1312.45E-03I-1313.18E-04I-1322.68E-09I-1313.09E-08I-1321.98E-05I-1322.57E-06I-1336.17E-08I-1322.82E-11I-1334.60E-04I-1335.97E-05I-1342.03E-10I-1334.52E-09 I-1341.14E-06I-1341.48E-07I-1351.19E-08I-1341.91E-14I-1356.77E-05I-1358.77E-06Cs-1341.20E-09I-1353.71E-10Cs-1346.00E-05Cs-1347.78E-06Cs-1357.31E-18Cs-1347.77E-11 Cs-1357.70E-13Cs-1359.99E-14Cs-1365.94E-10Cs-1356.54E-19Cs-1362.04E-05Cs-1362.65E-06Cs-1378.66E-10Cs-1362.60E-11Cs-1374.37E-05Cs-1375.67E-06Ba-137M8.21E-10Cs-1375.67E-11 Ba-137M4.14E-05Ba-137M5.36E-06Ba-1409.02E-12Ba-137M5.36E-11Ba-1403.42E-07Ba-1404.44E-08La-1407.56E-12Ba-1404.36E-13La-1403.63E-07La-1404.71E-08Ce-1413.70E-12La-1404.67E-13 Ce-1411.77E-07Ce-1412.29E-08Ce-1436.63E-13Ce-1412.27E-13Ce-1436.34E-09Ce-1438.23E-10Ce-1441.88E-12Ce-1436.91E-15Ce-1441.03E-07Ce-1441.34E-08Pr-1431.84E-12Ce-1441.34E-13 Pr-1437.36E-08Pr-1439.55E-09Pr-1441.90E-12Pr-1439.40E-14Pr-1441.03E-07Pr-1441.34E-08Pr-1441.34E-13Condensate DemineralizerHigh TDS Regenerant Collector TankSecondary Liquid Waste Drain Collector TankSecondary Liquid Waste Monitor TankIsotopeActivity (µCi/cc)IsotopeActivity (µCi/cc)IsotopeActivity (µCi/cc)IsotopeActivity (µCi/cc)

CALLAWAY - SPRev. OL-135/03TABLE 12.2-9 CONSERVATIVE BASIS ACCUMULATED RADIOACTIVITY IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATIONIsotopeActivity atPlant Shutdown(curies)Kr-8515317.5Kr-85m10.9 Kr-871.05 Kr-8810.6 Xe-131m128.5 Xe-13316536 Xe-133m145 Xe-13582 Xe-135m0.035 Xe-1380.04 I-1310.207 I-1320.00095 I-1330.04 I-1340.000187 I-1350.0065This table is based on 40 years' continuous operation with 0.25 percent fuel defect. Power is assumed to be 3,565 MWt. The data are based on a volume control tank purge rate of 0.7 scfm and 100 percent stripping efficiency.

CALLAWAY - SPRev. OL-135/03TABLE 12.2-10 DELETEDTable 12.2-10, Sheets 1 through 3, Deleted.

CALLAWAY - SPRev. OL-135/03TABLE 12.2-11 PARAMETERS AND ASSUMPTIONS FOR CALCULATING AIRBORNE RADIOACTIVE CONCENTRATIONS (REF. 7)a.Leak RatesPounds/DayEquivalent reactor coolant leak into containment during power for noble gases5,300Equivalent reactor coolant leak into containment for halogens5.3Equivalent reactor coolant leak into containment for other isotopes240b.Evaporation RatesPounds/HrFrom refueling pool into containment atmosphere (based on pool temperature of 120°F, and building air temperature of 70°F and 50 percent relative humidity and pool surface area of 1,500 square feet and 30 ft/minute flow parallel to the pool surface)499.5From fuel storage pool transfer canal and connecting slots into fuel building atmosphere during refueling (based on pool temperature of 137

°F and building air temperature of 110

°F and 95 percent relative humidity and with pool surface area of 2111.5ft 2 1024From fuel storage pool transfer canal and connecting slots into fuel building atmosphere during power (based on pool temperature of 95

°F and building air temperature of 70

°F and 50 percent RH and pool surface area of 2111.5ft 2 409c.Partition FactorsHalogensParticulatesTritiumContainment during power1.0010.35Fuel storage and refueling pool surfaces1Negligible1 CALLAWAY - SPTABLE 12.2-11 (Sheet 2)Rev. OL-135/03d.Ventilation RatesCFMContainment during power4,000Containment during refueling20,000 Fuel building during power20,000 Fuel building during refueling20,000e.Volumes of the RegionsCFContainment2.5 x 10 6Fuel building8.2 x 10 5f.Maximum Annual Individual OccupancyHrs/yrContainment during power 5hr/wk for50 wks/year250Containment during fuel handling 10 hr/day for ~ 6.25 days when the refueling pool is full of water62.5Fuel building during power 5hr/week for 50 weeks/year250Fuel building during refueling 10 hrs/day for ~ 10 days and 8hrs/day for ~ 3 days125g.Miscellaneous Information1.Failed fuel percentage for fission products0.122.Reactor coolant specific activitiesTable 11.1-13.Refueling and fuel storage pool concentration (µCi/gm). (These are the maximum concentrations during refueling)I-131I-1333.21 x 10-52.32 x 10-64.Fuel storage pool cleanup rate during power (gpm) (for conservatism, no cleanup of fuel storage pool or refueling pool is assumed during refueling)300 CALLAWAY - SPTABLE 12.2-11 (Sheet 3)Rev. OL-135/035.Decay of isotopes in the pools during refuelingNot included for conservatism6.Duration of refueling pool evaporation (hrs)1507.Duration of fuel storage pool evaporation during refueling (hrs)3208.Duration of fuel storage pool evaporation during power (hours/year)84409.Tritium release to environment via containment ventilation exhaust during refueling (Ci) (based on a total tritium release of 1000 Ci via gaseous effluents, durations of evaporation quoted in items 6, 7 and 8, and evaporation rates from pools given in Item b)2410.Tritium release to environment via fuel building ventilation exhaust during refueling (Ci) (same bases as given for Item9)10411.Tritium release to environment via fuel building ventilation exhaust during power (Ci) (same bases as given for Item9)1094 CALLAWAY - SPRev. OL-135/03TABLE 12.2-12 AIRBORNE RADIOACTIVITY CONCENTRATIONS (µCi/cc)NuclideContainment(power)Containment(refueling)Fuel Building(power)Fuel Building(refueling)H-32.34E-74.7E-63.81E-069.55E-06Cr-511.25E-12negligiblenegligiblenegligibleMn-542.07E-13negligiblenegligiblenegligibleFe-551.07E-12negligiblenegligiblenegligible Fe-596.63E-13negligiblenegligiblenegligibleCo-581.06E-11negligiblenegligiblenegligibleCo-601.33E-12negligiblenegligiblenegligible Br-831.76E-11negligiblenegligiblenegligibleBr-842.61E-12negligiblenegligiblenegligibleKr-83m6.60E-8negligiblenegligiblenegligible Kr-85m6.02E-7negligiblenegligiblenegligibleKr-851.18E-7negligiblenegligiblenegligibleKr-871.39E-7negligiblenegligiblenegligible Kr-888.34E-7negligiblenegligiblenegligibleKr-895.83E-10negligiblenegligiblenegligibleRb-888.01E-7negligiblenegligiblenegligible Rb-895.63E-10negligiblenegligiblenegligibleSr-893.54E-12negligiblenegligiblenegligibleSr-912.49E-13negligiblenegligiblenegligible Mo-995.06E-11negligiblenegligiblenegligibleTe-127m1.86E-13negligiblenegligiblenegligibleTe-1274.01E-13negligiblenegligiblenegligible Te-129m9.26E-13negligiblenegligiblenegligibleTe-1296.58E-13negligiblenegligiblenegligibleTe-131m1.34E-12negligiblenegligiblenegligible Te-1312.69E-13negligiblenegligiblenegligibleTe-1321.65E-11negligiblenegligiblenegligibleI-1301.95E-11negligiblenegligiblenegligible I-1313.82E-92.14E-10negligible7.6E-10I-1323.66E-10negligiblenegligiblenegligibleI-1334.14E-91.55E-11negligible5.5E-11 I-1347.41E-11negligiblenegligiblenegligibleI-1351.34E-9negligiblenegligiblenegligibleXe-131m2.75E-7negligiblenegligiblenegligible Xe-133m1.35E-6negligiblenegligiblenegligibleXe-1337.24E-5negligiblenegligiblenegligibleXe-135m7.64E-9negligiblenegligiblenegligible CALLAWAY - SPTABLE 12.2-12 (Sheet 2)Rev. OL-135/03Xe-1352.55E-6negligiblenegligiblenegligibleXe-1371.26E-9negligiblenegligiblenegligibleXe-1382.22E-8negligiblenegligiblenegligible Cs-1341.67E-11negligiblenegligiblenegligibleCs-1368.48E-12negligiblenegligiblenegligibleCs-1371.20E-11negligiblenegligiblenegligible Cs-1382.07E-8negligiblenegligiblenegligibleBa-137m1.14E-11negligiblenegligiblenegligibleBa-1401.43E-13negligiblenegligiblenegligible La-1401.07E-13negligiblenegligiblenegligibleNOTES:1.Iodine airborne concentrations during refueling are calculated very conservatively, assuming no purification by fuel pool cleanup system and no decay in the pool and a partition factor of 1 at the water-air interface at pool surface.2.Continuous pool cleanup, decay in the pool, and lower evaporation rates are expected to reduce iodine air concentrations to negligible levels during power operation in the fuel building.3.Xe-133 and I-131 air concentrations in the containment ventilation exhaust duct are expected to be 2 x 10-2 and 1.6x10

-6 µCi/cc, respectively during reactor head venting. However, the containment airborne concentrations for these isotopes are expected to be significantly less during head venting since the radioactivity is directly piped to the exhaust duct. The maximum value for Xe-133 is based on operating plant measurements normalized for a reactor system which has continuous stripping of the noble gases in the volume control tank. The maximum value for I-131 is based on operating plant measurements.NuclideContainment(power)Containment(refueling)Fuel Building(power)Fuel Building(refueling)

CALLAWAY - SP12.3-1Rev. OL-2211/1612.3RADIATIONPROTECTIONDESIGNFEATURES12.3.1FACILITY DESIGN FEATURESIn this section, specific design features for maintaining personnel exposures ALARA, commensurate with the guidance given in Regulatory Guide 8.8, are discussed.12.3.1.1PlantDesignDescriptionforasLowasis ReasonablyAchievable(ALARA)The equipment and plant design features employed to maintain radiation exposures as low as is reasonably achievable are based upon the design considerations of Section 12.1.2 and are outlined in this section for several general classes of equipment (Section 12.3.1.1.1) and several typical plant layout situations (Section 12.3.1.1.2

). 12.3.1.1.1Common Equipment and Component Designs for ALARAThis section describes the design features utilized for several general classes of equipment and components. These classes of equipment are common to many of the plant systems. Thus, the features employed for each system to maintain minimum exposures are similar and are discussed by equipment class in the following paragraphs.FILTERS - Liquid systems containing radioactive filters are provided with a semi-remote filter handling system for the removal of spent radioactive filter cartridges from their housings and for their transfer to the drumming station for packaging and shipment from the site for burial. The process is accomplished in such a manner that exposure to personnel and the possibility of inadvertent radioactive release to the environment is minimized. Each filter vessel is located within individual shielded compartments, provided with a ventilation supply and return, external remotely operated (via reach rods) vent and drain valving, and compartment drainage capabilities. Ventilation return ducts are equipped with a removable access panel through which portable radiation monitoring instruments can be inserted to check radiation levels within the compartment. The filter handling system with semi-remote tools has also been designed to be simple with a minimum of components susceptible to malfunction. Care has been taken to ensure that adequate space is provided to allow removing, cask loading, and transporting the cartridge to the solid radwaste area. DEMINERALIZERS - Demineralizers for radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to spent resin storage tanks prior to solidification and the fresh resin can be loaded into the demineralizer remotely. Underdrains and downstream strainers are designed for full system pressure drop. The demineralizers and piping are designed with provisions for being flushed with demineralized water. Strainers are installed in the vent lines to prevent the entry of spent resin into the exhaust duct. Each demineralizer compartment is equipped with a removable shield plug in the ceiling, offset from the centerline of the demineralizer CALLAWAY - SP12.3-2Rev. OL-2211/16vessel. Removal of the shield plug provides access for insertion of portable radiation monitoring instruments to check radiation levels within the compartment.EVAPORATORS - Evaporators are provided with chemical addition connections to allow chemicals to be used for descaling operations. Space is provided to allow the removal of heating tube bundles. The more radioactive components are separated by shield walls from those that are less radioactive. All instruments and controls are located on the accessible side of the shield wall. Valves in nonradioactive lines are located remote from the radioactive components. PUMPS - Wherever practicable, pumps have mechanical seals to reduce seal servicing time. Pumps and associated piping are arranged to provide adequate space for access to the pumps for servicing. Small pumps are installed in a manner which allows easy removal, if necessary. All pumps in the radioactive waste systems are provided with flanged connections for ease in removal. Piping or pump casing drain connections are provided for draining the pump for maintenance. TANKS - Whenever practicable, tanks are provided with sloped, dished heads or conical bottoms and bottom outlet connections. Overflow lines are directed to the waste collection system to control any contamination within the plant structures. HEAT EXCHANGERS - Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials with tube-to-tube sheet joints welded to minimize leakage. Impact baffles are provided, and tube side and shell side velocities are limited to minimize erosive effects. Flushing connections are provided. INSTRUMENTS - Instrument devices are located in low radiation zones and away from radiation sources, whenever practicable. Primary instrument devices which, for functional reasons, are located in high radiation zones are designed for easy removal to a lower radiation zone for calibration. Transmitters and readout devices are located in low radiation zones, such as corridors and the control room, for servicing. Some instruments (such as thermocouples) in high radiation zones are provided in duplicate to reduce the required access and service time, since immediate repair will not be necessary due to the backup instrument. Instrument and sensing line connections are located in such a way as to avoid corrosion product and radioactive gas buildup. VALVES - To minimize personnel exposures from valve operations, motor-operated, diaphragm, or other remotely actuated valves are used to the maximum extent practicable. Valves are located in valve galleries so that they are shielded separately from the major components. Long runs of exposed piping are minimized in valve galleries. In areas where manual valves are used on frequently operated process lines, either valve stem CALLAWAY - SP12.3-3Rev. OL-2211/16extenders or shielding are provided so that personnel need not enter the radiation area for valve operation. For infrequently operated equipment in Zone E, manual valves associated with safe operation, shutdown, and draining of the equipment may be provided with remote-manual operators or reach rods. For a definition of the radiation zones, see Figure 12.3-2. All other valve operations are performed with equipment in the shutdown mode. To the maximum practicable extent, simple straight reach rods have been used to allow the operators to retain the feel of whether the valves are tightly closed or not.

Valves with reach rods are installed either with their stems horizontal, with the reach rods also horizontal but above the heads of personnel, to allow ready access or with both the stem and reach rod orientation vertical, again to permit easy access. For valves located in the radiation areas, provisions are made to drain the adjacent radioactive components when maintenance is required. Wherever practicable, valves for clean, nonradioactive systems are separated from radioactive sources and are located in readily accessible areas. Manually-operated valves in the filter and demineralizer valve compartments required for normal operation and shutdown may be equipped with reach rods extending through or over the valve gallery wall. The valve gallery shield walls are designed for maximum expected filter activities. For most larger valves (2 1/2 inch and larger) in lines carrying radioactive fluids, either a double set of packing with a lantern ring and leakoff connection or a five ring set of packing with a carbon spacer are provided. Full ported valves are used in systems expected to contain radioactive solids.Valve designs with minimum internal crevices are used where crud trapping could become a problem, especially for piping carrying spent resin or evaporator bottoms. PIPING - The piping in pipe chases is designed for the lifetime of the unit. There are no valves or instrumentation in the pipe chase. Wherever radioactive piping is routed through areas where routine maintenance is required, pipe chases are provided to reduce the radiation contribution from these pipes to levels appropriate for the inspection requirements. Wherever practicable, piping containing radioactive material is routed to minimize the radiation exposure to the unit personnel. FLOOR DRAIN - Floor drains and properly sloped floors are provided for each room or cubicle having serviceable components containing radioactive liquids.Local gas traps or porous seals are not used on radwaste floor drains. Gas traps are provided at the common sump or tank.LIGHTING - Multiple electric lights are provided for each cell or room containing highly radioactive components so that the burnout of a single lamp will not require entry and CALLAWAY - SP12.3-4Rev. OL-2211/16immediate replacement of the defective lamp, since sufficient illumination will still be available. Normally, incandescent lights are provided which require less time for servicing, and hence the personnel exposure is reduced. The fluorescent lights which are used in some areas do not require frequent service, due to the increased life of the tubes. However, when the system in that room is secured and flushed out, the burned out lamps in the room can be replaced rapidly so as to minimize the exposure of the personnel.HVAC - The HVAC system design provides for the rapid replacement of the filter elements and housings.HYDROGEN RECOMBINERS - The more radioactive components are separated by a shield wall from those that are less radioactive. All instruments and controls are located on the accessible side of the shield wall. All valves in the radioactive lines are also located on the accessible side of the shield wall. Valves in the nonradioactive lines are located outside of the room.SAMPLE STATIONS - Sample stations for routine sampling of process fluids are located in the accessible areas. Shielding is provided at the local sample stations, as required, to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling. The counting room and laboratory facilities are described in Section12.5

.CLEAN SERVICES - Whenever practicable, clean services and equipment such as compressed air piping, clean water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways.12.3.1.1.2Common Facility and Layout Designs for ALARA This section describes the design features utilized for standard type plant processes and layout situations. These features are employed in conjunction with the general equipment designs described in Section 12.3.1.1.1 and include the features discussed in the following paragraphs.VALVE GALLERIES - Valve galleries are provided with shielded entrances for personnel protection. The valve galleries are divided into subcompartments which service only two or three components and are further subdivided by stud walls so that the personnel are only exposed to the valves and piping associated with one component at any given location. Threshold berms and floor drains are provided to control radioactive leakage. To facilitate decontamination in the valve galleries, concrete floors are covered with a smooth surfaced coating which will permit easy decontamination.PIPING - Pipes carrying radioactive materials are routed through controlled-access areas properly zoned for that level of activity. Radioactive piping runs are analyzed to determine the potential radioactivity level and area dose rate. Where it is necessary that radioactive piping be routed through corridors or other low radiation zone areas, shielded pipeways are provided. Valves and instruments are not placed in the radioactive CALLAWAY - SP12.3-5Rev. OL-2211/16pipeways. Whenever practicable, equipment compartments are used as pipeways only for those pipes associated with equipment in that compartment.When possible and practical, radioactive and nonradioactive piping are separated to minimize personnel exposure. Should maintenance be required, provision is made to isolate and drain the radioactive piping and associated equipment.Potentially radioactive piping is always located in appropriately zoned and restricted areas. Process piping is monitored to ensure that access is controlled to limit exposure

(Section12.5

).Piping is designed to minimize low points and dead legs. Drains are provided on piping where low points and dead legs cannot be eliminated. Thermal expansion loops are raised rather than dropped, where possible. In radioactive systems, the use of nonremovable backing rings in the piping joints is prohibited to eliminate a potential crud trap for radioactive materials. Butt welds are used in lieu of socket welds for all resin slurry and evaporator bottoms piping. Piping carrying resin slurries or evaporator bottoms is run vertically as much as possible, and large radius bends are utilized instead of elbows.Whenever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of the main pipe.PENETRATIONS - To minimize radiation streaming through penetrations, as many penetrations as practicable are located with an offset between the source and the accessible areas. If offsets are not practicable, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel. If these two methods are not used alternate means are employed, such as baffle shield walls, grouting the area around the penetration, or slanting the penetration through the wall.CONTAMINATION CONTROL - Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. Equipment vents and drains from highly radioactive systems are piped directly to the collection system instead of allowing any contaminated fluid to flow across to the floor drain. All welded piping systems are employed on contaminated systems, to the maximum extent practicable, to reduce system leakage and crud buildup at joints. The valves in radioactive systems 2-1/2 inch and larger are provided with leak-off connections piped directly to the collection system.Decontamination of potentially contaminated areas within the plant is facilitated by the application of suitable smooth surfaced coatings to the concrete floors.Floor drains with properly sloping floors are provided in all potentially contaminated areas of the plant. In addition, radioactive and potentially radioactive drains are separated from nonradioactive drains.

CALLAWAY - SP12.3-6Rev. OL-2211/16In controlled access areas where contamination is expected, radiation monitoring equipment is provided (Section12.3.4). Those systems which become highly radioactive, such as the radwaste slurry transport system, are provided with flush and drain connections. Certain systems have provisions for chemical and mechanical cleaning prior to maintenance.A decontamination facility, used for decontamination of removable components, is located adjacent to the auxiliary building. The decontamination system, shown in Figure12.3-4, consists of wash tanks, pumps, filters, spray booth, ultrasonic generator, turbulator, and associated piping. EQUIPMENT LAYOUT - In those systems where process equipment is a major radiation source (such as fuel pool cleanup, radwaste, etc.), pumps, valves, and instruments are separated from the process component. This allows servicing and maintenance of these items in reduced radiation zones. Control panels are located in low radiation zones (ZonesA orB).Major components (such as tanks, demineralizers, and filters) in the radioactive systems are isolated in individual shielded compartments, insofar as practicable.Provision is made on some major plant components for the removal of these components to lower radiation zones for maintenance.Labyrinth entranceway shields or shielding doors are provided for each compartment from which radiation could stream to access areas and exceed the radiation zone dose limits for those areas. For potentially high radiation components (such as filters and demineralizers), completely enclosed shielded compartments with hatch openings or removable wall sections are used.Equipment in the nonradioactive systems which requires maintenance is located outside the radiation areas.Figure 12.3-1 (Sheets 1-5) provides typical layout arrangements for demineralizers, filters, spent resin storage tanks, waste gas compressors, hydrogen recombiners, and sample racks and their associated valve compartments or galleries.Exposure from routine in-plant inspection is controlled by locating, whenever possible, inspection points in properly shielded low background radiation areas. Radioactive and nonradioactive systems are separated as far as practicable to limit radiation exposure from routine inspection of nonradioactive systems. For radioactive systems, emphasis is placed on adequate space and ease of motion in a properly shielded inspection area. Where longer times for routine inspection are required and permanent shielding is not feasible, sufficient space for portable shielding is provided. In high radiation areas where routine inspection is required, remote viewing devices are provided, as needed. When this is not practicable, written procedures which reduce radiation exposure by reducing CALLAWAY - SP12.3-7Rev. OL-2211/16the total time exposed to the radiation field are used, and access to the high radiation areas is under the direct supervision of the plants radiation protection personnel.FIELD RUN PIPING - All radioactive process piping, large and small, has been run and shielded by the architect/engineer. Scale models have been used to ensure that any possible interferences are taken into account.12.3.1.2RadiationZoningandAccessControlAccess to areas inside the plant structures and plant yards is regulated and controlled by radiation zoning and access control (Section 12.5.2). Each radiation zone defines the radiation level range to which the aggregate of all contributing sources must be attenuated by shielding. All plant areas are categorized into radiation zones according to expected radiation levels and anticipated personnel occupancy with consideration given toward maintaining personnel exposures ALARA and within the standards of 10 CFR 20.

Each room, corridor, and pipeway of every plant building is evaluated for potential radiation sources during normal operation, shutdown, and emergency operations; for maintenance occupancy requirements; for general access requirements; and for material exposure limits to determine appropriate zoning. Radiation zone categories employed and their descriptions are given in Figure 12.3-2 (Sheet 1), and the specific zoning for each plant area is shown in Figure12.3-2 (Sheets 1-6). All frequently accessed areas, i.e., corridors, are shielded for ZoneA or ZoneB access.The control of ingress or egress of plant operating personnel to controlled access areas and procedures employed to ensure that radiation levels and allowable working time are within the limits prescribed by 10 CFR 20 are discussed in Section 12.5

.Any area having a radiation level which could result in a dose equivalent in excess of 5 mrem in one hour at 30 centimeters from the source or from any surface that the radiation penetrates, is posted with signs bearing the radiation symbol and the words, "CAUTION, RADIATION AREA." Access alert barriers (e.g. signs, chain, rope, door, etc.) are provided for all radiation areas. Locations of these barriers are shown on Figure 12.3-2. Any areas having a radiation level which could result in a dose equivalent in any one hour in excess of 100 mrem at 30 centimeters from the source or from any surface that the radiation penetrates is posted with the radiation symbol and the words, "CAUTION (or DANGER), HIGH RADIATION AREA." High radiation areas are kept locked or barricaded if greater than 1,000 mrem/hr except during periods when access to the area is required in which case positive control is exercised over each individual entry. Any area in which radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in one hour at 1 meter from a radiation source or from any surface that the radiation penetrates is posted with the radiation symbol and the words, "GRAVE DANGER, VERY HIGH RADIATION AREA."Radiological surveys shall be posted or available at the entrance to the RCA and may be posted at the entrance to a radiological posted area requiring a RWP for entry. Hot spot CALLAWAY - SP12.3-8Rev. OL-2211/16tags are used to identify the highest source of radiation within a designated radiation, high radiation or very high radiation area. The locations of all radioactive equipment and the shield wall thicknesses are given on the general arrangement drawings in Section1.2

.The normal radiation level in the counting room will be the natural background level. Fly ash was specifically excluded from all concrete used for the counting room.The facilities and equipment used for handling and shielding radioactive equipment calibration sources are described in Section12.5

.12.3.2SHIELDINGThe bases for the nuclear radiation shielding and the shielding configurations are discussed in this section.12.3.2.1DesignObjectivesThe basic objective of the plant radiation shielding is to reduce personnel and population exposures, in conjunction with a program of controlled personnel access to and occupancy of radiation areas, to levels that are ALARA within the dose regulations of 10 CFR 20. Shielding and equipment layout and design are considered in ensuring that exposures are kept ALARA during all anticipated personnel activities in all areas of the plant containing radioactive materials, in accordance with Regulatory Guide 8.8.Two basic plant conditions are considered in the nuclear radiation shielding design: normal, full-power operation, and plant shutdown.The shielding design objectives for the plant during normal operation, including anticipated operational occurrences, and for shutdown operations are:a.To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10 CFR 20.b.To assure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspection, and safety-related operations required for each plant equipment and instrumentation area.c.To reduce potential equipment neutron activation and mitigate the possibility of radiation damage to materials.d.The control room will be sufficiently shielded such that the direct dose plus the inhalation dose (calculated in Chapter 15.0) will not exceed the limits of GDC-19 (see Section 6.4 for a more detailed discussion).

CALLAWAY - SP12.3-9Rev. OL-2211/1612.3.2.2GeneralShieldingDesignShielding is provided, as necessary, to attenuate direct radiation and scattered radiation through walls and penetrations to less than the upper limit of the radiation zone for each area shown in Figure 12.3-2. The minimum shielding requirements for all plant areas are given on scaled layout drawings in Section 1.2. General locations of the plant areas and equipment discussed in this section are also shown in the general arrangement drawings of Section 1.2. Design criteria for penetrations comply with the intent of Regulatory Guide 8.8 and are discussed in Section12.3.1.1.2

.The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 147 lb/ft

3. Whenever poured-in-place concrete has been replaced by concrete blocks or other material, design assures protection on an equivalent shielding basis, as determined by the characteristics of the material selected (Ref. 1). Compliance of concrete radiation shield design with Regulatory Guide 1.69 is discussed in Appendix 3A. Water is used as the primary shield material for areas above the spent fuel transfer and storage areas.For design basis accidents, the reactor building reduces the plant radiation intensities from fission products inside the containment to acceptable emergency levels, as defined by GDC-19, for the control room (see Sections 12.3.2.2.6 , 6.4, and 15.6.5).12.3.2.2.1Reactor Building Interior Shielding DesignDuring reactor operation, most areas inside the reactor building are Zone E and normally inaccessible.The main sources of radiation are the reactor vessel and the primary loop components, consisting of the steam generators, pressurizer, reactor coolant pumps, and associated piping. The reactor vessel is shielded by the concrete primary shield, reactor cavity shield, and the concrete secondary shield, which also surrounds all the other primary loop components. Air cooling is provided to prevent overheating, dehydration, and degradation of the shielding and structural properties of the primary shield.The reactor cavity seal/neutron shield, at the RPV seal ledge, attenuates neutron streaming. This seal/shield has been designed to provide equivalent neutron attentuation to 12 inches of water as was previously provided by water bag shielding. MORSE analyses have been performed which predict that limited personnel access may be allowed at the operating floor during power operation. See the response to NRC Question331.1 for a discussion of the Morse analyses. Components of the letdown system are located in shielded areas which are normally Zone E restricted access areas. Shielding is provided for N-16 delay piping, the excess letdown heat exchanger, and the regenerative heat exchanger.

CALLAWAY - SP12.3-10Rev. OL-2211/16After shutdown, most of the containment is accessible for limited periods of time, and all access is controlled. Areas are surveyed to establish allowable working periods. Dose rates are expected to range from 0.5 to 1,000 mrem/hr, depending on the location inside the containment (excluding reactor cavity). These dose rates result from residual fission products, neutron-activated materials, and corrosion products in the reactor coolant system.Spent fuel is the primary source of radiation during refueling due to its high buildup of fission product activity. However, radiation levels are limited in areas outside the refueling pool by the shielding effects of the thick structural walls of the refueling pool. The operators involved in the refueling operations are shielded from the spent fuel by the depth of water maintained above the fuel assemblies.12.3.2.2.2Auxiliary Building Shielding DesignDuring normal operation, the major components in the auxiliary building containing potentially high radioactivity are those in the chemical and volume control system. These include the letdown lines, the volume control tank, purification filters and demineralizers, and the charging pumps.Shielding is provided for each piece of equipment consistent with the access and zoning requirements of adjacent areas (Figure12.3-2

).Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to ZoneC access.Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.Following reactor shutdown, the residual heat removal (RHR) system pumps and heat exchangers are in operation to remove heat from the reactor coolant system. The radiation levels in the vicinity of this equipment will temporarily reach Zone E levels due to corrosion and fission products in the reactor water. Shielding is provided to attenuate radiation from RHR equipment during shutdown cooling operations to levels consistent with the radiation zoning requirements of the adjacent areas.12.3.2.2.3Fuel Building Shielding DesignSpent fuel is the primary source of radiation in the fuel building. Because of the extremely high activity of the fission products contained in the spent fuel elements and the proximity of Zone B and C areas, extensive shielding has been provided for areas surrounding the fuel storage pool, cask loading pit and the fuel transfer canal to ensure CALLAWAY - SP12.3-11Rev. OL-2211/16that radiation levels remain below zone levels specified for adjacent areas. Water provides the shielding above the spent fuel assemblies during fuel handling operations.During spent fuel loading operations, the HI-TRAC VW transfer cask provides shielding through a design employing water jackets, steel barriers and lead shielding which minimizes personnel exposure. For a detailed discussion of shielding during ISFSI operations, refer to UMAX FSAR Section 5.The fuel pool cooling and cleanup system (FPCCS) (Section9.1.3) shielding is based on the maximum activity discussed in Section12.2.1 and the access and zoning requirements of adjacent areas. Equipment in the FPCC system to be shielded includes the FPCC heat exchangers, pumps, piping, filters, and demineralizers.12.3.2.2.4Radwaste Building Shielding DesignShielding is provided, as necessary, around the following equipment in the radwaste building to ensure that the radiation zone and access requirements are met for surrounding areas.a.Liquid waste collection tanks and pumpsb.Liquid waste monitor tanks and pumpsc.Chemical drain tank and pump d.Liquid waste boron recycle and secondary liquid waste evaporatorse.Solid radwaste solidification equipmentf.Solid radwaste drumming and storage areas g.Evaporator bottoms tanksh.Liquid radwaste pipingi.Radwaste filters and demineralizers j.Spent resin storage tank and pumpk.Gaseous radwaste surge tank, recombiners, and compressorsl.Gas decay tanksShielding is based upon operation with maximum activity conditions, as discussed in Sections 11.1 , 11.2 , 11.3, and11.4.

CALLAWAY - SP12.3-12Rev. OL-2211/16Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to ZoneC access.Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.12.3.2.2.5Turbine Building Shielding DesignRadiation shielding is not required for any process equipment located in the turbine building. All areas in the turbine building are classified ZoneA.12.3.2.2.6Control Room Shielding Design The design basis LOCA dictates the shielding requirements for the control room. Shielding is provided to permit access and occupancy of the control room under LOCA conditions with radiation doses limited to 5 rem whole body from all contributing modes of exposure for the duration of the accident, in accordance with GDC-19. A complete discussion of control room habitability during a LOCA is provided in Section 6.4. Figure 12.3-3 provides an isometric view of the control room shielding.12.3.2.2.7Diesel Generator Building Shielding Design There are no radiation sources in the diesel generator building. Therefore, no shielding is required within the building.12.3.2.2.8Miscellaneous Plant Areas and Plant Yard AreasSufficient shielding is provided for all plant buildings containing radiation sources so that radiation levels at the accessible outside surfaces of the buildings are maintained below Zone A levels. Plant yard areas which are frequently occupied by plant personnel are fully accessible during normal operation and shutdown. These areas are surrounded by a security fence and closed off from areas accessible to the general public. Access to outside storage tanks which have a contact dose rate greater than 0.5 mrem/hr is restricted by a fence located at a distance at which the dose rate is less than 0.5 mrem/hr.12.3.2.2.9Independent Spent Fuel Storage Installation (ISFSI) Shielding DesignThe ISFSI is designed for interim dry storage of spent nuclear fuel. The shielding is sufficient to comply with the requirements of 10 CFR 72.104 and 10 CFR 72.106. The dose rate on the VVM closure lid is calculated to be 0.25 mrem/hr neutron and 0.69 mrem/hr gamma for a total dose rate of 0.94 mrem/hr. The dose rate on the outlet duct screen is calculated to be 0.80 mrem/hr neutron and 1.32 mrem/hr gamma for a total CALLAWAY - SP12.3-13Rev. OL-2211/16dose rate of 2.12 mrem/hr. The dose rate on the side of the HI-TRAC VW transfer cask is calculated to be 237 mrem/hr neutron and 1,611 mrem/hr gamma, for a total dose rate of 1,848 mrem/hr. The regulatory dose limit at the owner-controlled area boundary is 25 mrem annual whole body dose for Normal and Off-Normal conditions and 5 rem whole body for accident conditions.For a detailed discussion of the VVM and HI-TRAC VW transfer cask shielding properties, refer to the HI-STORM UMAX FSAR Chapter 5.12.3.2.3ShieldingCalculationalMethodsThe shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in Section12.2.1. The thickness of each shield wall surrounding the radioactive equipment is determined by approximating, as closely as possible, the actual geometry and physical condition of the source or sources. The isotopic concentrations are converted to energy group sources, using data from the Table of Isotopes (Ref.2).The geometric model (Ref.3-11), assumed for the shielding evaluation of pipe tanks, heat exchangers, filters, demineralizers, evaporators, and the containment is a finite cylindrical volume source. In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source (Ref.3-11). Typical computer codes that are used for shielding analysis are listed in Table12.3-1 (Ref.12-21).The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area. Shielding requirements are evaluated at the point of maximum radiation dose through any wall. Therefore, the actual anticipated radiation levels in the greater region of each plant areas are less than this maximum dose rate and, therefore, less than the radiation zone upper limit.Where shielded entryways to compartments containing high radiation sources are necessary, labyrinths or mazes are used. The mazes are constructed so that the scattered dose rate plus the transmitted dose rate through the shield wall from all contributing sources are below the upper limit of the radiation zone specified for each plant area.12.3.3VENTILATIONThe plant heating, ventilating, and air-conditioning (HVAC) systems are designed to provide a suitable environment for personnel and equipment during normal operation and anticipated operational occurrences. Parts of the plant HVAC systems perform safety-related functions.

CALLAWAY - SP12.3-14Rev. OL-2211/1612.3.3.1DesignObjectivesThe plant HVAC systems for normal plant operation and anticipated operational occurrences are designed to meet the requirements of 10 CFR 20, "Standards for Protection Against Radiation," and 10 CFR 50, "Licensing of Production and Utilization Facilities."12.3.3.2DesignCriteriaDesign criteria for the plant HVAC systems include the following:a.During normal operation and anticipated operational occurrences, the average and maximum airborne radioactivity levels to which plant personnel are exposed in the restricted areas of the plant are ALARA and within the limits specified in 10CFR20.b.During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary will be ALARA and within the limits specified in 10 CFR 20 and 10 CFR 50.c.The plant siting dose guidelines of 10 CFR 100 will be satisfied, following those hypothetical accidents described in Chapter15.d.The dose to control room personnel shall not exceed the limits specified in GDC-19, following those hypothetical accidents described in Chapter 15.0 and Section6.4

.12.3.3.3DesignGuidelinesIn order to accomplish the design objectives, the following guidelines are followed, wherever practicable.12.3.3.3.1Guidelines to Minimize Airborne Radioactivitya.Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.b.Equipment vents and drains are piped directly to a collection device connected to the collection system, instead of allowing any contaminated fluid to flow across the floor to the floor drain.c.All-welded piping systems are employed on contaminated systems, to the maximum extent practicable, to reduce system leakage. If welded piping systems are not employed, drip trays are provided at the points of potential leakage. Drains from drip trays are piped directly to the collection system.

CALLAWAY - SP12.3-15Rev. OL-2211/16d.Suitable coatings are applied to the concrete floors of potentially contaminated areas to facilitate decontamination.e.To minimize the amount of airborne radioactivity as a result of valve leakage, all valves 2-1/2 inches and larger in the radioactive systems are provided with double sets of packing with lantern rings and with leak-off connections that are piped to drain headers. Diaphragm or bellows seal valves are used on those systems where essentially no leakage can be tolerated.f.Contaminated equipment has design features that minimize the potential for airborne contamination during maintenance operations. These features include flush connections for draining and flushing the pump prior to maintenance and flush connections on piping systems that could become highly radioactive.12.3.3.3.2Guidelines to Control Airborne Radioactivitya.The airflow is directed from areas with lesser potential for contamination to areas with greater potential for contamination.b.In building compartments with a potential for contamination, a greater volumetric flow is exhausted from the area that is supplied to the area to minimize the amount of uncontrolled exfiltration from the area.c.Consideration is given to the possible disruption of normal airflow patterns by maintenance operations, and provisions are made in the design to prevent adverse air flow direction.d.The air cleaning system's design, maintenance, and testing criteria are discussed in detail in the response to Regulatory Guides 1.52 and 1.140 found in Section 9.4. An illustrative example of the air cleaning system design is given in Section12.3.3.5

.e.Air being discharged from potentially contaminated areas is passed through HEPA filters and charcoal absorbers to remove particulates and halogens or means are provided to isolate these areas upon indication of contamination to prevent the discharge of contaminants to the environment.f.Suitable containment isolation valves are installed in accordance with GDC-54 and 56, including valve controls, to assure that the containment integrity is maintained.g.Redundant seismic Category I systems and/or components are provided for portions of the ventilation system that serve areas required for safe CALLAWAY - SP12.3-16Rev. OL-2211/16shutdown of the reactor plant. Included herein are the plant control room and selected engineered safety feature equipment rooms.h.Atmospheric tanks which contain radioactive materials are vented to the respective building ventilation system.12.3.3.3.3Guidelines to Minimize Personnel Exposure from HVAC Equipmenta.Ventilation fans and filters are provided with adequate access space to permit servicing with minimum personnel radiation exposure. The HVAC system is designed to allow rapid replacement of components.b.Ventilation ducts are designed to minimize the buildup of radioactive contamination within the ducts, to the extent practicable.c.Ventilating air is recirculated in the clean areas only.d.Access and service of ventilation systems in potentially radioactive areas is provided by component location to minimize operator exposure during maintenance, inspection, and testing as follows:1.The outside air supply units and building exhaust system components are enclosed in ventilation equipment rooms. These equipment rooms are located in radiation Zone B and are accessible to the operators. Work space will be provided around each unit for anticipated maintenance, testing, and inspection.2.Local cooling equipment servicing the normal building requirements will be located in areas of low contamination potential (radiation Zones A or B) (refer to Figure 12.3-2

).12.3.3.4DesignDescriptionThe ventilation systems serving the following structures are considered to be potentially radioactive and are discussed in detail in Section9.4

a.Containment building (see Section 9.4.6

)b.Auxiliary building (see Section 9.4.3

)c.Fuel building (see Section 9.4.2

)d.Radwaste building (see Section 9.4.5

)e.Turbine building (see Section 9.4.4

)

CALLAWAY - SP12.3-17Rev. OL-2211/16f.Portions of the access control area (see Section 9.4.1.2

).Although the control room is considered to be a nonradioactive area, radiation protection is provided to assure habitability (see Section6.4

).Other structures (e.g., pump intake structures, administrative building, etc.) contain no potential source of radioactivity and are not addressed in this chapter.12.3.3.5AirCleaningSystemDesignThe guidance and recommendations of Regulatory Guides 1.52 and 1.140 concerning maintenance and in-place testing provisions for atmospheric cleanup systems, air filtration, and adsorption units have been used as a reference in the design of the various ventilation systems. The extent to which Regulatory Guide 1.52 and 1.140 have been followed is discussed in Section9.4.Provisions specifically included to minimize personnel exposures and to facilitate maintenance or in-place testing operations are as follows:a.The loading of the filters and adsorbers with radioactive material during normal plant operation is a slow process. Therefore, in addition to monitoring for pressure drop, the filters are checked for radioactivity on a scheduled maintenance basis with portable equipment. The filter elements are replaced before the radioactivity level is of sufficient magnitude to create a personnel hazard. Filters whose radioactivity level (due to a postulated accident) is such that a change of filter elements would constitute a personnel hazard can be removed intact. No shielding is provided since it is not required for the level of radioactivity developed during normal operation. In case of excessive radioactivity caused by a postulated accident, the whole filter is replaced before normal personnel access is resumed. It will not be necessary for workers to handle filter units immediately after a design basis accident, so exposures can be minimized by allowing the short-lived isotopes to decay before changing the filter.b.Active elements of the atmospheric cleanup systems are designed to permit ready removal.c.Access to active elements is direct from working platforms to simplify element handling. Ample space is provided on the platforms for accommodating safe personnel movement during replacement of components, including the use of necessary material handling facilities and during any in-place testing operation.d.No filter bank is more than three filter units high, where each filter unit is 2 feet by 2 feet. The access to the level or platform at which the filter is serviced is by stairs or elevators.

CALLAWAY - SP12.3-18Rev. OL-2211/16e.The clear space for doors is a minimum of 3feet by 7feet.f.The filters are designed with replaceable 2feet by 2feet units that are clamped in place against compression seals. The filter housing is designed, tested, and proven to be airtight with bulkhead type doors that are closed against compression seals.12.3.4AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION12.3.4.1AreaRadiationMonitoringThe area radiation monitoring system (ARMS) is provided to supplement the personnel and area radiation survey provisions of the plant radiation protection program described in Section12.5 and to ensure compliance with the personnel radiation protection guidelines of 10 CFR 20, 10 CRF 50, 10 CFR 70, and Regulatory Guides8.2, 8.8, and 8.12. 12.3.4.1.1Design Bases The principal objectives and criteria of the ARMS are provided below.

12.3.4.1.1.1Safety Design BasesThe area radiation monitoring system has no function related to the safe shutdown of the plant or the capability to mitigate the consequences of accidents that could result in offsite exposures comparable to the guideline exposure of 10 CFR 100 and, therefore, has no safety design bases. See Appendix7A for a discussion of Regulatory Guide1.97. 12.3.4.1.1.2Power Generation Design BasesPOWER GENERATION DESIGN BASIS ONE - The ARMS functions continuously to immediately alert plant personnel entering or working in nonradiation or low-radiation areas of increasing or abnormally high radiation levels which, if unnoticed, could possibly result in inadvertent overexposures. POWER GENERATION DESIGN BASIS TWO - The ARMS serves to inform the control room operator of the occurrence and approximate location of an abnormal radiation increase in nonradiation or low-radiation areas.POWER GENERATION DESIGN BASIS THREE - The ARMS complies with the requirements of 10 CFR 50, Appendix A, General Design Criterion 63 for monitoring fuel and waste storage and handling areas.

CALLAWAY - SP12.3-19Rev. OL-2211/16POWER GENERATION DESIGN BASIS FOUR - Certain monitors located near the spent fuel pool and new fuel storage vault act as criticality alarm monitors and conform to the requirements of 10 CFR 70, Regulatory Guides 8.5 and 8.12 and Standards ANSI/

ANS-8.3-1979 and USASN2.3-1967.12.3.4.1.1.3Codes and Standards Codes and standards applicable to the area radiation monitors are indicated in Table3.2-1. 12.3.4.1.2System Description12.3.4.1.2.1General DescriptionThe ARMS consists of five-decade range GM tube detectors located throughout the plant to warn personnel of abnormal gamma radiation levels. The detector signals are transmitted to the control room over individual cables. The displays, both local and in the control room, are five-decade logarithmic ratemeters. The alarms are both audible and visual, and are located in the control room and near the local detectors. 12.3.4.1.2.2Criteria for Area Monitor Selection The following design criteria are applicable to the area radiation monitoring system. RANGE - The ARMS has a five-decade range from 10

-1 to 10+4 mrem/hr, except for the reactor coolant sample room radiation monitor which has a range from 10 to 10 5 mrem/hr. The ranges are made sufficiently wide to measure the radiation levels expected in the areas concerned. The system continues to read upscale if exposed to radiation levels above the maximum range. SENSITIVITY - Gamma sensitive to photon energies of 100 keV and above.

RESPONSE - In any range, the readout indicates at least 90 percent of its end point reading within 5 seconds after a step change in radiation level at the detector. ENERGY DEPENDENCE - The dose rate (mrem/hr) readout is within +/-20 percent of the actual dose rate in each detected area from photon energies between 100 keV and 2.5meV. DIRECTIONAL DEPENDENCE - The dose rate readout does not vary more than 10 percent when exposed to a single point radiation source of approximately 1.0MeV from any point in the frontal hemisphere of a horizontal plane.

CALLAWAY - SP12.3-20Rev. OL-2211/16ENVIRONMENTAL DEPENDENCE - The system meets the above requirements for all variations of temperature, pressure, and relative humidity within each area monitored, as listed below: For instruments located outside the containment: For instruments located inside the containment:

EXPOSURE LIFE - Each monitor located inside the containment maintains its characteristics up to an integrated dose of 10 7rads. Each monitor located outside the containment maintains its characteristics up to an integrated dose of 10 6rads. 12.3.4.1.2.3AlarmsEach monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two-alarm setpoints provide sequential alarms on increasing radiation levels. Loss of power will cause an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected. 12.3.4.1.2.4Check Sources Each monitor is provided with a check source, operated from the control room, which simulates a radiation level in the area for operational and gross calibration checks. 12.3.4.1.2.5Power Supplies The power supplies for all of the monitors are given in Table12.3-4. 12.3.4.1.2.6Calibration and MaintenanceThe area radiation monitors are calibrated by the manufacturer, using a Cs-137 source. The manufacturer's calibration standards are traceable to The National Bureau of Standards primary calibration standard sources and are accurate to at least 5 percent. Temperature,°F60 to 120Humidity, %RH5 to 95PressureAtmosphericTemperature,°F50 to 150Humidity, %RH5 to 100Pressure+/- 2 psig CALLAWAY - SP12.3-21Rev. OL-2211/16The source-detector geometry during this primary calibration is identical to the source-detector geometry in field calibrations. A secondary standard counted in reproducible geometry during the primary calibration is supplied with each monitoring system. Channel checks and source checks are performed at regular intervals to ensure proper monitor function. The monitors are recalibrated at regular intervals, and following repairs or modifications, using the secondary radionuclide standard.12.3.4.1.2.7SensitivitiesEach area radiation monitor is able to detect radiation levels as low as 0.1 mrem/hr, except for post-accident sample room radiation monitor which detects radiation levels as low as 10mrem/hr. 12.3.4.1.2.8Criteria for Location of Area Monitors Generally, area radiation monitors are provided in areas to which personnel normally have access and for which there is a potential for personnel unknowingly to receive high radiation doses (e.g., in excess of 10 CFR 20 limits) in a short period of time because of system failure or improper personnel action. Any plant area which meets one or more of the following criteria is monitored: a.ZoneA areas which, during normal plant operations, including refueling, could exceed the radiation limit of 0.5 mrem/hr upon system failure or personnel error or which would be continuously occupied following an accident requiring plant shutdown. b.Zone B areas where personnel could otherwise unknowingly receive high levels of radiation exposure due to system failure or personnel error. c.Areas in which quantities of special nuclear material in excess of 10 CFR70.24 quantities are received, stored, or handled, e.g. spent fuel pool and new fuel storage vault. In the case of new fuel, two detectors are provided at the operating floor of the fuel building to serve as criticality monitors. In the case of spent fuel, two detectors are provided on opposite walls above the spent fuel pool to serve as criticality monitors. In both cases, the monitors will provide a distinct audible and visual personnel alarm at no more than 15mrem/hr to signal the need for personnel evacuation. These monitors are provided in accordance with 10 CFR70.24 requirements and will detect the criticality described in 10CFR70.24(a)(1).d.For areas in which fuel is stored or radioactive waste systems and handling equipment are located, area radiation monitors are provided to detect CALLAWAY - SP12.3-22Rev. OL-2211/16conditions that might result in loss of residual heat removal capability and excessive radiation levels and to alert the operators to initiate appropriate safety action in accordance with GDC-63 of 10 CFR 50, AppendixA. The location of each area radiation detector is indicated on the radiation zoning and access control drawings, Figure 12.3-2, and are listed in Table 12.3-2. Consistent with the above criteria, the following general areas are monitored: a.Main control room b.Radwaste building corridorsc.Auxiliary building corridorsd.Fuel storage and handling area e.Radwaste pipe tunnelf.Railway access (fuel building)g.Hot machine and hot instrument shops h.Containmenti.Radwaste solidification areaj.Sampling rooms and laboratory12.3.4.1.2.9Setpoints The bases for the area radiation monitor setpoints are determined from the need to alert operators to abnormal radiation levels in the area. The setpoints are sufficiently above the normal radiation levels in the measured areas to avoid spurious alarms. The normal setpoints for the individual area radiation monitors are provided in Table12.3-2. To support temporary plant conditions or evolutions, the normal setpoint levels may be increased. At these increased levels, access to the area will be controlled in accordance with the Radiation Protection Program based on actual radiation levels. Area radiation monitor setpoints are administratively controlled when not at the normal setpoints.12.3.4.1.2.10Safety Evaluation The ARMS is designed to operate unattended for extended periods of time, detecting and measuring ambient gamma radiation. Ambient radiation dose rate at the detector is indicated locally at the detector and remotely in the main control room. These monitors CALLAWAY - SP12.3-23Rev. OL-2211/16cause an audible and visual alarm at the detector and in the main control room if the radiation levels exceed preset limits. All components are solid state, and the system is designed for high reliability. The system is not essential for safe shutdown of the plant, and it serves no active emergency function during operation. The system serves to warn plant personnel of high radiation levels in various plant areas. All monitors are independent, and failure of one unit has no effect on any other. 12.3.4.2AirborneRadioactivityMonitoringMonitoring for the presence of airborne radioactivity inside the plant is necessary for the protection of plant personnel, in compliance with Regulatory Guide8.2 and within the limits established by 10CFR 20.1-20.601. The airborne radioactivity monitors provide information necessary to ensure that gaseous, particulate, and iodine radioactivity do not exceed 10 MPC hours in areas occupied by the station personnel. The systems consist of permanently installed, continuous monitoring devices together with a program and provisions for specific sample collections and laboratory analyses. 12.3.4.2.1Design Bases The principal objectives and criteria of the airborne radiological monitoring systems (AiRMS) are provided below. 12.3.4.2.1.1Safety Design BasesSAFETY DESIGN BASES - There are no safety design bases for the monitoring of airborne radioactivity for inplant personnel protection. The control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are required to automatically initiate operation of engineered safety features systems in the event that airborne radioactivity in excess of the allowable limits exists. Additional design bases are given in the following sections: a.Containment purge isolation system, Sections6.2.4 , 7.3.2 , 9.4, and11.5. b.Fuel building ventilation isolation, Sections7.3.3 , 9.4.2, and 11.5. c.Control room intake isolation, Sections6.4.1 , 7.3.4 , 9.4.1 , and 11.5. These radioactivity monitors are protection system elements and are designed in accordance with IEEE Standard 279 due to safety design bases of 6.2.4 , 6.4.1 , 7.3 , 9.4 , and11.5.

CALLAWAY - SP12.3-24Rev. OL-2211/16The safety evaluation of these systems is discussed in Section7.3. These monitors also serve for inplant worker protection, and this function is discussed at length in this section. 12.3.4.2.1.2Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The airborne radioactivity monitors operate continuously to detect airborne particulates, iodine, and/or noble gases in the air upstream of all filters in the containment, auxiliary building, fuel building, radwaste building, waste gas decay tank rooms, access control area, and control room for the protection of the workers. POWER GENERATION DESIGN BASIS TWO - The airborne radioactivity monitors are designed to detect 10 MPC-hours or better in any compartment or room served by the monitoring system. POWER GENERATION DESIGN BASIS THREE - The containment atmosphere monitors are designed to detect leakage of radioactivity from the reactor coolant system into the containment atmosphere. This function is described in greater detail in Section5.2.5. The containment purge monitors will serve as a backup to the containment atmosphere monitors while the purge is in operation. 12.3.4.2.1.3Codes and Standards Codes and standards applicable to the airborne radioactivity monitors are indicated in Table3.2-1. The monitors listed in Section12.3.4.2.1.1 have additional codes and standards applied due to their safety-related functions discussed in other sections, as noted above. 12.3.4.2.2System Description 12.3.4.2.2.1General Description 12.3.4.2.2.1.1Data Collection The AiRMS consist of particulate, iodine, and noble gas monitors with the attendant controls, alarms, pumps, valves, and indicators required to meet the design objectives in Section 12.3.4.2.1. Each monitor consists of the detector assembly and a local microprocessor. The local microprocessor processes the detector assembly signal in digital form, computes average radioactivity levels, stores data, performs alarm or control functions, and transmits the digital signal to the control room microprocessor. Signal transmission is accomplished via two two-wire daisy-chain loops. Each loop allows data transmission in either direction, ensuring that a single fault in the loop will not prevent the control room microprocessor from receiving the data.

CALLAWAY - SP12.3-25Rev. OL-2211/16The local microprocessors for monitors which perform safety functions (control room ventilation, fuel building ventilation, containment atmosphere, and containment purge monitors, refer to Section11.5) are wired directly to individual indicators located on the seismic CategoryI AiRMS cabinets in the control room. The input from the safety-related channels to the daisy-chain loop is an isolated signal to ensure that the safety-related signals will not be affected by signals or conditions existing in the nonsafety portion of the system. The control room microprocessor provides controls and indication for the AiRMS. Indication is via a Visual Display located in the control room. The signals from each monitor may also be recorded on a system printer. The safety-related monitors are also recorded on analog strip chart recorders. 12.3.4.2.2.1.2Selection Criteria for Airborne Monitors12.3.4.2.2.1.2.1IntroductionThe type of fixed instrumentation used for monitoring airborne radioactivity is offline. The offline system extracts a sample from the process stream and transports that sample to the radiation monitoring system which contains the specified equipment to detect particulates, halogens, and/or noble gases. 12.3.4.2.2.1.2.2Sampling Criteria The sampling system for the particulate/halogen/noble gas monitors is designed and installed in accordance with ANSI N13.1-1969 guide to sampling of airborne radioactive materials. Systems whose sensitivity is dependent upon sample flow employ isokinetic nozzles and suitable control of the flow rate. 12.3.4.2.2.1.2.3Detection Criteria Since both radioactive particulates and radioactive noble gases are beta emitters, beta-sensitive scintillation detectors are used to sense radioactivity to minimize the effects due to background radiation and, consequently, obtain a lower minimum detectable concentration. Where spectrometric analysis is required (such as in iodine monitoring) an NaI (Tl) gamma scintillation detector assembly is employed. 12.3.4.2.2.1.2.4Instrumentation Criteria Instrumentation necessary to indicate, alarm, and perform control functions is provided to complete the monitoring system. Since radioactive concentrations may vary substantially, wide-range instruments are utilized. All airborne radiation monitors include provisions for obtaining a gas sample for laboratory isotopic analysis. The particulate and charcoal filters can readily be removed for laboratory analyses.

CALLAWAY - SP12.3-26Rev. OL-2211/16The airborne particulate monitors each consist of a fixed filter upon which radioactive particulate matter is deposited by means of a positive displacement pump that draws a continuous sample, using an isokinetic nozzle from the ventilation exhaust duct for the particular area. The fixed filter is located in front of a beta scintillation detector coupled to a photomultiplier tube which responds to the scintillations emitted from the crystal as a result of incident radiation giving up its energy within the crystal. Each airborne iodine monitor consists of a charcoal cartridge upon which iodine is adsorbed. The air sample is prefiltered to remove particulates. The charcoal cartridge is located in front of a gamma scintillation detector coupled to a photomultiplier tube. Each airborne noble gas monitor consists of a fixed volume sample chamber through which prefiltered sample air is passed. A beta scintillation detector is located within the sample chamber to detect the activity level of the air sample. All of the detectors and sample chambers are enclosed in heavily shielded lead pigs. Two motor-operated valves, operated locally, are provided to permit air purging of the sample chamber to facilitate background activity checks. The sensitivities and alarm setpoints are given in Table12.3-3. The high-alarm points are based on the most restrictive isotopes which are expected to be present. 12.3.4.2.2.1.3Criteria for Airborne Radioactivity Monitor LocationsThe criteria for locating airborne radioactivity monitors are dependent upon the point of leakage, the ability to identify the source of radioactivity so that corrective action may be performed, and whether personnel may be exposed to the airborne radioactivity. a.Airborne radioactivity monitors sample normally accessible personnel operating areas for which there is a potential for airborne radioactivity. b.Areas not normally accessible are monitored, prior to personnel entry, with portable monitors or samplers, depending upon the potential for airborne radioactivity and the work to be performed in the area. c.Exhaust ducts servicing an area containing processes which, in the event of major leakage, could result in concentrations within the plant approaching the limits established by 10CFR20.1-20.601 for plant workers are monitored. d.Dilution from other exhaust ducts is considered when locating monitors in exhaust systems to ensure maximum coverage and still be able to detect 10 CFR 20.1-20.601 airborne radioactivity limits in the area with the lowest ventilation flow.

CALLAWAY - SP12.3-27Rev. OL-2211/16e.Outside air intake ducts for the control building are monitored to measure possible introduction of radioactive materials into the control room to ensure the habitability of those areas requiring personnel occupancy for safe shutdown. f.Airborne radioactivity monitors are located so that the actual sample chamber and detector location are in an area where the background radiation is low. Detailed physical locations are provided on the radiation zone drawings, Figure12.3-2. 12.3.4.2.2.1.4AlarmsEach monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two alarm setpoints provide sequential alarms on increasing radioactivity levels. Loss of power will cause an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected. Alarms from the AiRMS are provided in the control room on the plant annunciator (audible and visual), the balance-of-plant computer alarm display (visual), and the AiRMS Visual Display. In addition, the safety-related channels have individual alarm lights on the safety-related indicators on the AiRMS control panel. The balance-of-plant and AiRMS computers also provide printouts of each alarm. The pumping systems are controlled from the control room and are provided with a low-flow alarm to alert the operator of pump failure or any other condition which causes a loss of flow through the sample system, and a flow control valve and flow controller to automatically compensate for filter loading. 12.3.4.2.2.1.5Check Sources Each monitor is provided with a check source, operated from the control room, which simulates a radioactive sample in the detector assembly for operational and gross calibration checks. 12.3.4.2.2.1.6Power Supplies All Class 1E inplant AiRMS are powered from Class 1E motor control centers. The power supplies for all of the monitors are given in Table12.3-4. 12.3.4.2.2.1.7Calibration and MaintenanceThe airborne radioactivity monitors are calibrated by the manufacturer for the principal radionuclides listed in Table 12.3-3. The manufacturer's calibration standards are traceable to National Bureau of Standards primary calibration standard sources and are accurate to at least 5 percent. The source detector geometry during this primary CALLAWAY - SP12.3-28Rev. OL-2211/16calibration is identical to the sample detector geometry in actual use. Secondary standards counted in reproducible geometry during the primary calibration are supplied with each continuous monitor. Channel checks and source checks are performed at regular intervals to ensure proper monitor function. The monitors are recalibrated at regular intervals, and following repairs or modifications, using the secondary radionuclide standard.12.3.4.2.2.1.8Sensitivities The AiRMS is capable of detecting 10 MPC-hours of airborne radioactivity. The most restrictive isotope for each type of monitor is that isotope with the lowest worker maximum permissible concentration (WMPC), as defined in TableI, ColumnI, of Appendix B to 10 CFR 20.1-20.601.For the containment atmosphere and containment purge system monitors, the High and Alert alarm set points are based on the Offsite Dose Calculation Manual. The sensitivities and alarm setpoints are given in Table 12.3-3. The High alarm points are based on the most restrictive isotopes which are expected to be present. The concentration levels are as defined in TableI, ColumnI, of AppendixB to 10 CFR 20.1-20.601 or Technical Specification limits, considering dilution. The sensitivity of the airborne radioactivity monitors is based on a 95 percent confidence level for 0.5MeV beta or gamma radiation in a 1mr/hr gamma radiation background at standard pressure and ambient temperature. The fixed volume noble gas detector assemblies have a minimum detectable concentration of 2x10

-7 µCi/cc, using Kr-85/Xe-133 as the limiting isotope. The fixed filter particulate detector assemblies have a minimum detectable concentration of 1x10-11 µCi/cc, using Cs-137 as the limiting isotope. The filter assembly has a collection efficiency of 99percent for particles of 0.3micron or larger. The charcoal filter halogen detector assemblies have a minimum detectable concentration of 1 x 10-11 µCi/cc, using I-131 as the limiting isotope. The charcoal filter assembly has a collection efficiency of at least 95percent for iodine. 12.3.4.2.2.1.9Ranges and SetpointsThe ranges of the various airborne radioactivity monitors were chosen based on the detection of radioactivity in concentrations ranging from 10 MPC-hours or lower in compartments served up to those from postulated accidents.

CALLAWAY - SP12.3-29Rev. OL-2211/16The fixed volume noble gas detector assemblies have a range of 10

-7 to10-2 µCi/cc. The fixed filter particulate detector assemblies have a range of 10-12 to 10-7 µCi/cc. The charcoal filter halogen detector assemblies have a range of 10-11 to 10-6 µCi/cc. The setpoints are chosen to alert the operators to airborne radioactivity that might be present so that 10 CFR 20.1-20.601 limits on worker exposure or Technical Specification limits will not be exceeded. The setpoints for control of ventilation are discussed in Section7.3. The setpoints on the monitors used for reactor coolant pressure boundary leakage detection are discussed in Section5.2.5. The ranges and setpoints for the airborne radioactivity monitors are provided in Table12.3-3. 12.3.4.2.2.1.10Expected System Parameters The expected ranges of system parameters, such as flow rate, composition, and concentrations, are summarized in Table12.3-3. Detailed information on the individual HVAC systems can be found in Section9.4. 12.3.4.2.2.2Monitoring SystemsThe systems discussed in the following sections are summarized in Table12.3-3. 12.3.4.2.2.2.1Access Control Area Ventilation Exhaust Radioactivity MonitorThe access control area ventilation exhaust radioactivity monitor, 0-GK-RE-41, continuously monitors for particulate radioactivity in the access control area ventilation exhaust upstream of the HVAC filters. The sample is extracted from the duct through an isokinetic nozzle, in accordance with ANSI Standard N13.1-1969, to ensure that a representative sample of the system is obtained. After passing through the fixed filter detector assembly and the pumping system, the sample is discharged back to the duct. The high and high-high alarms function to alert the operator to airborne particulate radioactivity in the access control area. Noble gases are monitored by the unit vent monitor described in Section11.5. The airborne iodine radioactivity in the access control area can be inferred from the particulate and noble gas radioactivity levels. Since no radioactive equipment is in the area, there is no source for release of iodine or noble gas. Particulates may be present due to dust on the protective clothing becoming airborne.

CALLAWAY - SP12.3-30Rev. OL-2211/16Further determination of the source and the corrective action to be taken will be based on monitoring with portable detection and sampling equipment. Indication for this system is provided on the AiRMS Visual Display in the control room. 12.3.4.2.2.2.2Deleted 12.3.4.2.2.2.3Waste Gas Decay Tank Area Ventilation Exhaust Radioactivity MonitorThe waste gas decay tank area ventilation radioactivity monitor, 0-GH-RE-23, continuously monitors for gaseous radioactivity in the discharge duct from the waste gas decay tank area upstream of the radwaste building exhaust filter adsorber. The sample point provides rapid detection of a leak in the waste gas processing system and, in conjunction with the radwaste building exhaust radioactivity monitor and the radwaste building effluent monitor, helps localize the affected area in the event of an alarm on either monitor.The sample is extracted from the exhaust duct and passed through the fixed volume noble gas detector assembly and the pumping system. Then the sample is discharged back to the duct. The high alarm provides indication of a leak in the decay tanks, compressors, piping, or valves. The high-high alarm indicates that concentrations in the decay tank room are at or near 10 MPC for the most restrictive isotope expected to be present (Kr-85 or Xe-133).Back up for this monitor is provided by the radwaste building exhaust and effluent monitors.Indication of this monitor is provided on the AiRMS Visual Display in the control room.12.3.4.2.2.2.4Auxiliary Building Ventilation Exhaust Radioactivity Monitor The auxiliary building ventilation exhaust radioactivity monitor, 0-GL-RE-60, continuously monitors for particulate radioactivity in the auxiliary building ventilation system upstream of the filter-adsorber units. The sample point is located to monitor between the last point of possible radioactivity entry to the ventilation system from the areas served and the filter adsorber unit. The sample is extracted through an isokinetic nozzle, in accordance with ANSI Standard N13.1-1969, to ensure that a representative sample is provided to the fixed filter particulate detector assembly. Then the sample is discharged through the pumping system back to the duct. The cartridge filter will be removed for laboratory isotopic analysis upon activation of high activity alarm. The high alarm alerts the operator to high airborne particulate radioactivity levels in the auxiliary building. Indication of this monitor is provided on the AiRMS Visual Display in the control room.

CALLAWAY - SP12.3-31Rev. OL-2211/16If required, the portable monitor described in Section12.3.4.2.2.2.9 and/or grab samples will be utilized to determine airborne radioactivity levels and iodine concentrations in specific areas to aid in the determination of the source of the release. 12.3.4.2.2.2.5Containment Atmosphere Radioactivity MonitorsThe containment atmosphere radioactivity monitors, 0-GT-RE-31 and 0-GT-RE-32, continuously monitor the containment atmosphere for particulate, iodine, and gaseous radioactivity which could result in personnel exposure during periods of containment access. Other functions of these monitors are covered in Sections5.2.5 , 7.3 , 9.4 , and 11.5.Samples are extracted from the operating deck level (El. 2047'-6") of the containment through the monitoring system sample lines.The monitors are located as close as possible to the containment penetrations to minimize the length of the sample tubing and the effects of sample plateout. The sample points are located in areas which ensure that representative samples are obtained. Each sample passes through the penetration and then through the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detector assemblies. After passing through the pumping system, the sample is discharged back to the containment through a separate penetration.Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and through isolated signals, on the AiRMS Visual Display in the control room. The containment atmosphere radioactivity monitors are seismic CategoryI systems and completely redundant. The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere. 12.3.4.2.2.2.6Containment Purge System Radioactivity MonitorsThe containment purge system radiation monitors, 0-GT-RE-22 and 0-GT-RE-33, continuously monitor the containment purge exhaust duct during normal purge operations for particulate, iodine, and gaseous radioactivity for worker protection as backup monitors for the containment atmosphere monitors. Other functions for the containment purge monitors are given in Sections5.2.5 , 7.3 , 9.4, and 11.5. The purge monitors are seismic CategoryI and completely redundant.

The sample points are located outside the containment between the containment isolation dampers and the containment purge filter adsorber unit.

CALLAWAY - SP12.3-32Rev. OL-2211/16Each monitor is provided with two isokinetic nozzles to ensure that representative samples are obtained from both normal purge and minipurge. Isokinetic nozzle selection is accomplished by sample selector valves which automatically align the correct nozzle to the monitor, based on operation of the minipurge and normal purge exhaust fans. The sample is extracted through the selected nozzle and then passes through the selector valve, the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detectors. The sample then passes through the pumping system and is discharged back to the duct. Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS Visual Display in the control room. The containment purge radiation monitors provide backup for the containment atmosphere radiation monitors. The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere. For plant conditions during CORE ALTERATIONS and during movement of irradiated fuel within containment, the function of the monitors is to alarm only and the trip signals for automatic actuation of CPIS may be bypassed. One instrumentation channel at a minimum is required for the alarm only function during plant refueling activities.12.3.4.2.2.2.7Control Room Ventilation Radioactivity MonitorsThe control room ventilation radioactivity monitors, 0-GK-RE-04 and 0-GK-RE-05, continuously monitor the supply air of the normal heating, ventilation, and air-conditioning system for particulate, iodine, and gaseous radioactivity to provide protection for the control room operators in the event of high airborne radioactivity in the control room HVAC supply duct. This seismic CategoryI system is completely redundant. Samples are extracted through individual isokinetic nozzles, in accordance with ANSI Standard N13.1-1969, and flow through the fixed filter (particulate), charcoal filter (iodine), and fixed volume (gaseous) detector assemblies prior to passing through the pumping system for discharge to the auxiliary building atmosphere. The high and high-high alarms alert operators to high airborne radioactivity in the control room supply duct. The safety control functions are described in Sections6.4 , 7.3, and 11.5. Indication for these monitors is provided on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS Visual Display in the control room.

CALLAWAY - SP12.3-33Rev. OL-2211/1612.3.4.2.2.2.8Fuel Building Ventilation Exhaust Radioactivity MonitorsThe fuel building ventilation exhaust radioactivity monitors, 0-GG-RE-27 and 0-GG-RE-28, continuously monitor for particulate, iodine, and gaseous radioactivity in the fuel building ventilation exhaust system for the protection of the workers in the fuel building. The other functions for these monitors are described in Sections9.4 and 11.5. During normal operation, each of the monitors extracts a sample from the normal exhaust duct through individual isokinetic nozzles and sample selector valves. This normal sample point is upstream of the fuel building normal exhaust filter adsorber unit. The high and high-high alarms alert operators to high airborne particulate, iodine, or gaseous radioactivity in the fuel building. The ventilation control functions are described in Sections7.3 and 11.5. Indication is provided by individual indicators on the AiRMS control panel and through isolated signals by the AiRMS Visual Display in the control room. 12.3.4.2.2.2.9Portable MonitorPortable air radioactivity samplers are provided to allow periodic localized monitoring of specific air volumes of interest independent of the fixed monitor systems. The samplers are used to verify that airborne radioactivity levels within the plant operating spaces are within allowable limits and also to verify the proper operation of fixed monitor systems. A portable particulate, iodine, and gaseous continuous air monitor or a particulate only continuous air monitor will be used to monitor the local areas where there is a possibility of airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally, using the portable monitor. 12.3.4.2.2.3Safety EvaluationDue to their safety-related functions, discussed in other sections, the control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are redundant, independent, seismic CategoryI with Class 1E power supplies. These monitors all have safety-related control functions which are described in Section7.3. The following monitors are located upstream of filters and therefore, are effective for inplant personnel protection: a.Containment atmosphereb.Containment purge CALLAWAY - SP12.3-34Rev. OL-2211/16c.Control room supplyd.Fuel building exhaust e.Auxiliary building exhaustf.Access control area exhaustg.Waste gas decay tank room exhaust h.Portable monitorAll the inplant areas where the potential for airborne radioactivity exists are, therefore, monitored. The process and effluent radioactivity monitors are discussed in Section11.5. When the dilution factors of the ventilation system are large enough to preclude direct detection of halogens and noble gases at MPC levels, the release of these isotopes will be accounted for because the particulate activity, present along with halogens and noble gases, will be detected at much lower concentrations. Corresponding ratios among the various species of the airborne radioactivity are periodically determined by laboratory analysis so that the leakage of halogen and gaseous radioactivity can be accounted for. The waste gas decay tank room exhaust radioactivity monitor is designed to detect noble gas radioactivity, because this is the predominant type of activity present. Particulates and the iodine may be present in very small amounts, but will not be significant in this case since they will be removed by the waste gas processing system. A portable particulate, iodine, and gaseous monitor or a particulate only continuous air monitor is used to monitor the local areas where there is a possibility of airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally, using the portable monitor. On the basis of the above discussion, the AiRMS is adequate and sufficient to ensure personnel protection from airborne radioactivity. The system will provide indication to the operator that airborne radioactivity exists. The location of the airborne radioactivity can then be further identified by using portable air samplers that can be manually connected to various selected sample points in the building. Use of these sample points will direct radiation protection personnel to the particular area of concern. The exact location of the airborne radioactivity can then be found by sampling the particular subcompartments in that area. The combination of the AiRMS, in conjunction with administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns CALLAWAY - SP12.3-35Rev. OL-2211/16throughout the plant, plant equipment layout, and restricted radiation Zone E areas, is sufficient to ensure that airborne radioactivity levels are safe in terms of the required duration of personnel access throughout all areas of the plant. A general review of these concepts follows: a.Equipment location is such that hot piping and equipment are located in Radiation Zone D and E areas, which are restricted, and entry is limited by administrative control. Radiation Zone B and C areas do not contain piping and components that would result in significant airborne radioactivity sources. This reduces the possibility of airborne radioactivity exposure to occupants of Radiation Zone B and C areas where general entry is permitted.b.Air flow patterns are consistent with the basic ventilation design criteria of the plant. Clean filtered outside air is supplied to Zone B areas (corridors, clean areas); these areas are exhausted by drawing air into the rooms and areas of successively higher potential for airborne contamination. Air flow is such that air flow reversal or exfiltration from potentially contaminated areas is precluded. This ventilation arrangement eliminates the possibility of personnel exposure to airborne radioactivity in continuous occupancy areas, such as Radiation Zone B areas. c.High radiation areas (Radiation Zone E areas) where dose levels exceed 100 mrem/hr are restricted and conspicuously posted in accordance with 10 CFR 20.100-20.2401. These areas are not normally entered. d.Radiation protection programs are discussed in Section12.5. 12.

3.5REFERENCES

1.R. L. Walker and M. Grotenhuis, ASummaryofShielding ConstantsforConcrete , ANL-6443 (November 1961).2.C. M. Lederer, et. al., TableofIsotopes, Lawrence Radiation Laboratory, University of California (March 1968).3.R. G. Jaeger, et. al., EngineeringCompendiumon RadiationShielding, Shielding Fundamentals and Methods, I (1968).4.G. W. Goldstein, X-RayAttenuationCoefficientsfrom 10keVto100MeV , National Bureau of Standards Circular 583 (issued April 10, 1957).5.ReactorPhysicsConstants, Argonne National Laboratory, ANL-5800 (July 1963).

CALLAWAY - SP12.3-36Rev. OL-2211/166.J. F. Kircher and R. E. Bowman, EffectsofRadiation on Materials and Components (March 1964). 7.T. Rockwell, ReactorShieldingDesignManual, D. Van Nostrand Co., New York (1956). 8.C. R. Tipton, Jr., ReactorHandbook, Vol. I, Materials, second edition (1962). 9.H. Soodak, ReactorHandbook, Vol. III, Part A, Physics, second edition (1962). 10.E. P. Blizzard and L. S. Abbott, ReactorHandbook Vol.III, Part B, Shielding, second edition (1962).11.N. M. Schaeffer, ReactorShieldingforNuclearEngineers, TID-25951 (1967). 12.D. S. Duncan and A. B. Spear, GraceI-AnIBM704-709 Program Design for Computing GammaRayAttenuationandHeatinginReactorShields, Atomics International (June 1959). 13.D. S. Duncan and A. B. Spear, GraceII-AnIBM709 Programfor Computing Gamma RayAttenuationandHeatinginCylindricalandSphericalGeometries , Atomics International (November 1959). 14.W. W. Engle, Jr., AUser'sManualforANISN:AOne Dimensional Discrete Ordinates TransportCodewith AnisotropicScattering, Union Carbide Corporation, Report No. K-1693 (1967). 15.E. D. Arnold and B. F. Maskewitz, SDC,AShielding DesignCodefor Fuel Handling Facilities, ORNL-3041 (March 1966). 16.Richard E. Malenfant, "QAD,ASeriesofPoint-Kernel General-Purpose Shielding Programs," Los Alamos Scientific Laboratory, LA 3573 (October 1966).17.D. A. Klopp, NAP-MultigroupTime-DependentNeutron Activation Prediction Code, IITRI-A6088-21 (January 1966). 18.E. A. Straker, P. N. Stevens, D. C. Irving, and V. R. Crain, MORSE-AMultigroup Neutron andGamma-Ray MonteCarloTransportCode, ORNL-4585 (September 1970). 19.W. A. Rhoades and F. R. Mynatt, TheDOTIII Two-DimensionalDiscrete Ordinates TransportCode, ORNL-TM-4280 (1973). 20.R. E. Malefant, "G 3:AGeneralPurposeGamma-Ray ScatteringProgram," Los Alamos Scientific Laboratory, LA 5176 (June 1973).

CALLAWAY - SP12.3-37Rev. OL-2211/1621.M. J. Bell, "ORIGEN-TheORNLGenerationand DepletionCode," Oak Ridge National Laboratory, ORNL-4628 (May 1973).

CALLAWAY - SPRev. OL-135/03TABLE 12.3-1 LIST OF COMPUTER CODES USED IN SHIELDING DESIGN CALCULATIONSGRACE IMultigroup, multiregion, gamma-ray attenuation code used to compute gamma heating and gamma dose rates in slab geometry (Ref. 12). GRACE IIMultigroup, multiregion, gamma-ray attenuation code used to compute the dose rate or heat generation rate for a spherical or a cylindrical source with slab or concentric shields (Ref. 13). ANISNMultigroup, multiregion code solving the Boltzman transport equation for neutrons or gamma-rays in one dimensional slab, cylindrical, or spherical geometry (Ref. 14). SDCMultigroup, multiregion, Kernal integration gamma-ray, shield design code which calculates dose rates for 13 geometry options (Ref. 15). QADMultigroup, multiregion, three-dimensional, point Kernal code which calculates fast neutron and gamma-ray dose and heat generation rates (Ref. 16). NAPDetermines activation emission source strengths as a function of neutron exposure and decay time (Ref. 17). MORSE-CGThree-dimensional Monte Carlo neutron and gamma ray general transport code (Ref. 18). DOT IIITwo-dimensional neutron, gamma ray, discrete ordinate, transport code (Ref. 19). ORIGENIsotope generation and depletion code which solves equations of radioactive growth and decay for isotopes of arbitrary coupling (Ref. 21). G 3A general purpose gamma-ray scattering code (Ref. 20).

CALLAWAY - SPRev. OL-1610/07TABLE 12.3-2 AREA RADIATION MONITORSInstrumentNumberLocationRadiation ZoneRange (mrem/hr)Hi Alarm (mrem/hr)Hi-Hi Alarm (mrem/hr)0-SD-RE-1Radwaste Building Corridor, BasementB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-2Radwaste Building Corridor, BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-3Radwaste Building Corridor, BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-4Radwaste Building Corridor, Ground FloorB0.1-10 42.5 (1)(8)15 (2)(8)0 SD-RE-5Radwaste Building Corridor, Ground FloorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-6Solid Radwaste AreaB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-7Truck SpaceB0.1 - 10 415 (8)100 (8)0-SD-RE-8Sample LaboratoryB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-9RW Bldg Valve Room CorridorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-10RW Bldg Valve Room CorridorC0.1-10 415 (1)(8)100 (2)(8)0-SD-RE-11RW Bldg HVAC Filter UnitB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE 12Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-13Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-14Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-15Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-16Aux Bldg Corridor BasementB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-17Pipe Tunnel & Personnel AccessB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-18Aux Bldg Ground Floor CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-19Aux Bldg Ground Floor CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-20Aux Bldg Valve Room Corridor Ground FloorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-21Aux Bldg Valve Room Corridor Ground FloorC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-22Aux Bldg Corridor Ground FloorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-23Aux Bldg Corridor Ground FloorB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-24RC Sample RoomC10 - 10 515 (1)(8)100 (1)(8)0-SD-RE-25Filter Unit Aux BldgA0.1 - 10 40.5 (1)(8)2.5 (2)(8)0-SD-RE-26RHR Heat Exchanger OutsideB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-27Ctmt Purge Filter UnitC0.1 - 10 415 (1)(8)100 (2)(8)0-SD-RE-28Personnel HatchB0.1-10 42.5 (1)(8)15 (2)(8)0-SD-RE-29Laundry Decontamination FacilityB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-30Laundry Decontamination FacilityB0.1-10 42.5 (1)(8)15 (2)(8)

CALLAWAY - SPTABLE 12.3-2 (Sheet 2)Rev. OL-1610/070-SD-RE-31Hot LaboratoryB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-32Control Bldg CorridorB0.1 - 10 42.5 (1)(8)15 (2)(8)0-SD-RE-33Control RoomA0.1 - 10 40.5 (1)(8)2.5 (2)(8)0-SD-RE-34Cask Handling Area B0.1 - 10 42.5 (1)15 (2)0-SD-RE-35New Fuel Storage Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-36New Fuel Storage Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-37Fuel Storage Pool Area (3)B0.1 - 10 42.5 (1)15 (2)0-SD-RE-38Fuel Storage Pool Area (3)B0.1 -10 42.5 (1)15 (2)0-SD-RE-39Seal Table AreaE0.1 - 10 41000 (4)(8)10000 (5)(8)0-SD-RE-40Personnel Access Hatch AreaE0.1 - 10 41000 (4)(8)10000 (5)(8)0-SD-RE-41Manipulator Bridge Crane (6)E0.1 -10 430 (7)100 (7)0-SD-RE-42Containment BuildingE0.1 -10 41000 (4)(8)10000 (5)(8)0-SD-RE-47Boron Concentration Monitoring System RoomE0.1 - 10 415 (1)(8)100 (2)(8)(1)High alarm set for maximum radiation level for the radiation zone for that area.

(2)High-high alarm set for the maximum radiation level for the radiation zone above the one for that area.

(3)These monitors serve as criticality monitors.

(4)High alarm based on highest expected radiation level for the area during full power operation.

(5)High-high alarm set one order of magnitude above high alarm to indicate extremely high radiation level.

(6)This monitor will be operational on the bridge crane at these setpoints only during plant shutdown when fuel handling activities are occurring in mode 6. During modes 1, 2, 3, 4, and 5 the high alarm and high-high alarm setpoints will be reset to 5,000 mrem/hr and 10,000 mrem/hr respectively.(7)High and high-high alarms are established for the protection of operators on the bridge crane during mode 6.

(8)High alarm and high-high alarm setpoints listed are the normal alarm setpoints. Alarm setpoints above these levels may be used to support plant conditions and will be administratively controlled.InstrumentNumberLocationRadiation ZoneRange (mrem/hr)Hi Alarm (mrem/hr)Hi-Hi Alarm (mrem/hr)

CALLAWAY - SPRev. OL-174/09TABLE 12.3-3 INPLANT AIRBORNE RADIOACTIVITY MONITORSMonitorType (continuous)Range

µCi/ccMDC(1)

µCi/ccControlling IsotopeAlert (15) Alarm

µCi/ccOGTRE31Particulate (3) 10-12 to 10-71 x 10-11Cs1371.0 x 10-9 (16) OGTRE32Iodine (4) 10-11 to 10-61 x 10-10I1311.0 x 10-8Containment atmosphere monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-1331.0 x 10-4OGTRE22Particulate (3) 10-12 to 10-71 x 10-11Cs137 5 x 10-8OGTRE33Iodine (4) 10-11 to 10-61 x 10-10I131 5 x 10-8Containment purge system monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-133(10)OGGRE27Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-8OGGRE28Iodine (4) 10-11 to 10-61 x 10-10I131 9 x 10-9Fuel building exhaust monitors (2)Gaseous (3) 10-7 to 10-22 x 10-7Xe-1331.6 x 10-3OGKRE04Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-8OGKRE05Iodine (4) 10-11 to 10-61 x 10-10I131 9 x 10-9Control room air supply monitorsGaseous (3) 10-7 to 10-22 x 10-7Xe-1331.1 x 10-3OGLRE60Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-8Auxiliary Building ventilation exhaust monitorOGKRE41Particulate (3) 10-12 to 10-71 x 10-11Cs137 1 x 10-9(8)Access control area ventilation exhaust monitorOGHRE23Gaseous (3) 10-7 to 10-22 x 10-7Kr85Xe-133 1 x 10-5(8)Waste gas decay tank area ventilation exhaust monitorPortable monitorParticulate (3) 10-12 to 10-71 x 10-11Cs137NAIodine (4) 10-11 to 10-61 x 10-10I131NAGaseous (3) 10-7 to 10-22 x 10-7Kr85NA CALLAWAY - SPTABLE 12.3-3 (Sheet 2)Rev. OL-174/09 IHigh (15)AlarmµCi/ccFlowVentilationFlow (cfm)SubcompartmentFlow Rate(cfm)DilutionFactorMinimum RequiredSensitivity

(µCi/cc)Monitor Control Function1 x 10-7420,000NANA1 x 10-7(6)(6)NA9 x 10-7420,000NANA9 x 10-8(6)(6)NA6.0 x 10-4420,000NANA1 x 10-4(6)(6)NA1 x 10-720,000/4,000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and control functions.9 x 10-820,000/4,000NANA9 x 10-8(6)(6)(11) (14)20,000/4,000NANA1 x 10-4(6)(6)1 x 10-720,000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and control functions.9 x 10-820,000NANA9 x 10-8(6)(6)3.2 x 10-3(13)20,000NANA1 x 10-4(6)(6)1 x 10-72000NANA1 x 10-7(6)(6)See Table 11.5-3 for process and control functions.9 x 10-72000NANA9 x 10-8(6)(6)2.2 x 10-3(12)2000NANA1 x 10-4(6)(6)1 x 10-712,000100 8 x 10-3(5)8 x 10-10(6),(9)Alarms1 x 10-8(7)6,0001001.67 x 10-2(5)1.67 x 10-9(6),(9)Alarms1 x 10-4(7)5002500.5(5)5 x 10-5(6),(9)AlarmsNANANANAAlarmsNANANANANANANANASample Flow for each channel is 3 cfm.

(1)MDC = minimum detectable concentration.(2)When fuel is in the building.(3)Beta scintillation detector.

(4)Gamma scintillation detector.(5)Dilution factor = (6)Minimum required sensitivity of monitor in

µCi/cc at 10 MPChrs for the controlling isotope=dilution factor x 10 MPC.(7)10 MPC x dilution factor.Subcompartmental flow in cfmTotal flow in cfm----------------------------------------------------------------------

-

CALLAWAY - SPTABLE 12.3-3 (Sheet 3)Rev. OL-174/09(8)MPC x dilution factor.(9)Grab samples will be analyzed in the laboratory, and iodine concentrations will be calculated, using previously established ratios.(10)Alert alarm is administratively established at a point sufficiently below the High alarm so as to provide additional assurance that Offsite Dose Calculation Manual (ODCM) limits arenot exceeded. (11)High alarm is set to ensure that ODCM limits are not exceeded.(12)Submersion dose rate does not exceed 2 mr/hr in the control room.(13)Submersion dose rate does not exceed 4 mr/hr in the fuel building.

(14)High alarm setpoint is established to ensure that ODCM limits are not exceeded.(15)Alert and High alarm setpoint values do not include instrument loop uncertainty estimates.(16)Alert alarm value is set to meet the criteria of Note 10 and to meet RCS leakage detection requirements described in FSAR Section 5.2.5.2.3

.

CALLAWAY - SPRev. OL-195/12TABLE 12.3-4 POWER SUPPLIES FOR AREA AND IN-PLANT AIRBORNE MONITORSArea Radiation MonitorsMonitorNumberNormalPowerSupplyRestored AfterLoss of OffsitePowerRemarks0-SD-RE-1 0-SD-RE-20-SD-RE-30-SD-RE-4 0-SD-RE-50-SD-RE-60-SD-RE-7 0-SD-RE-80-SD-RE-90-SD-RE-10 0-SD-RE-110-SD-RE-120-SD-RE-13 0-SD-RE-140-SD-RE-150-SD-RE-16 0-SD-RE-170-SD-RE-180-SD-RE-19 0-SD-RE-200-SD-RE-210-SD-RE-22 0-SD-RE-230-SD-RE-240-SD-RE-25 0-SD-RE-260-SD-RE-270-SD-RE-28 0-SD-RE-290-SD-RE-300-SD-RE-31 0-SD-RE-320-SD-RE-330-SD-RE-34 0-SD-RE-350-SD-RE-36Supplied by regulated instrumentation ac power which is supplied from the diesel generators on loss of offsite power.Yes CALLAWAY - SPTABLE 12.3-4 (Sheet 2)Rev. OL-195/120-SD-RE-370-SD-RE-38 0-SD-RE-390-SD-RE-400-SD-RE-41 0-SD-RE-420-SD-RE-47Containmentatmosphere0-GT-RE-31 0-GT-RE-32Class 1E MCCsYesContainmentpurge system0-GT-RE-220-GT-RE-33Class 1E MCCsYesFuel buildingexhaust 0-GG-RE-270-GG-RE-28Class 1E MCCsYes Control room air supply0-GK-RE-04 0-GK-RE-05Class 1E MCCsYes In-Plant Airborne Radioactivity Monitors (Non-1E)Auxiliary building ventilation exhaust 0-GL-RE-60Non-1E MCCsNoPower is lost to system also, so monitor reading is not meaningful.Access control area ventilation exhaust0-GK-RE-41Non-1E MCCSNoPower is lost to system also, so monitor reading is not meaningful.Waste gas decay tank area ventilation exhaust 0-GH-RE-23Non-1E MCCsNoPower is lost to system also, so monitor reading is not meaningful.MonitorNumberNormalPowerSupplyRestored AfterLoss of OffsitePowerRemarks CALLAWAY - SP12.4-1Rev. OL-2211/1612.4DOSEASSESSMENTRadiation exposures in the plant are primarily due to direct radiation from components and equipment containing radioactive fluids. In addition, in some plant radiation areas there can be radiation exposure to personnel due to the presence of airborne radionuclides. In-plant radiation exposures during normal operation and anticipated operational occurrences are discussed in Section 12.4.1. Radiation exposures due to direct radiation at locations outside the plant structures, such as the boundary of the restricted area, are a function of the plant layout, equipment selection, and detailed system and shielding designs and are expected to be negligible. Radiation exposures due to the airborne radioactive effluent plume at these locations are expected to be insignificant. The radiation exposures at these locations are discussed in Section 12.4.2.2.Radiation exposures to operating personnel will be within 10CFR 20 limits. Radiation protection design features described in Section 12.3 and the radiation protection program outlined in Section 12.5 will assure that the occupational radiation exposures (ORE) to operating personnel during operation and anticipated operational concurrences will be as low as is reasonably achievable (ALARA).12.4.1EXPOSURES WITHIN THE PLANT 12.4.1.1DirectRadiationDoseEstimatesAnnual man-rem doses from direct radiation during the performance of routine functions, such as operation and surveillance, normal maintenance, radwaste handling, refueling, and inservice inspection, have been estimated, using the following bases:a.Radiation exposure data from operating PWRs are given in Tables 12.4-1 through 12.4-11 (Ref. 1, 2, 3) and Figure 12.4-1. b.Expected average dose rates in plant radiation areas are discussed in this section.c.Expected occupancy times for various work function personnel in the different plant radiation areas are listed in Table 12.4-12

.d.Anticipated manpower requirements for each unit are given in Table 12.4-12.e.Table 12.4-12 provides an estimate of the distribution of the annual man-rem according to work function.

CALLAWAY - SP12.4-2Rev. OL-2211/16The maximum and expected average dose rates in the plant radiation areas are given below:The maximum dose rates are determined by shielding calculations based on conservative assumptions regarding sources, self-shielding, locations, etc. The expected dose rates are estimated by assuming a failed fuel percentage of 0.12 and that stringent water chemistry control and improved design will minimize crud buildup and hence the expected dose rates in various radiation zones and recognition that real maximum doses in a given zone are localized effects. The expected average doses given above are used in computing the doses for personnel involved in all operations, except inservice inspection and special maintenance. For personnel involved in the performance of inservice inspection (ISI) and special maintenance tasks, an expected average dose rate of 200 mR/hr in the E Zone is used since these personnel will generally be working on reactor coolant system components.Direct radiation exposures to plant personnel can result from the performance of special maintenance functions. In view of the radiation protection design features described in Section12.3 and the radiation protection program outlined in Section 12.5, it is expected that exposures due to special maintenance will be minimized. However, an annual exposure of 150 man-rem/unit is realistically estimated, based on experience at operating PWRs.Exposure to plant personnel from direct radiation during the performance of routine functions is estimated to be approximately 220 man-rem/yr/unit. Details of the man-rem estimates are given in Table 12.4-12. A breakdown of the exposures (including the special maintenance category) by work functions is provided in Table 12.4-12. Table 12.4-13 provides the percentage of the annual total man-rem associated with each work function.Total occupancy in various radiation zones: Total occupancy time for personnel involved in different work functions in various radiation zones is as follows:ZoneMaximum Dose Rate Expected Average Dose Rate (mrem/hr)(mrem/hr)A 0.5 0.1 B 2.5 0.5C 15 2.5 D 100 15E>100100 CALLAWAY - SP12.4-3Rev. OL-2211/16a.Routine operation and surveillance - The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 2,080 hrs/yr. The major portion of the occupancy is expected to be in radiation Zones A and B. Combined occupancy in radiation Zones C, D, and E is expected to be approximately 4 percent of the total annual occupancy. The average annual dose for an individual involved in this work category is expected to be about 1 rem. The distribution of occupany times in various radiation zones is listed in Table 12.4-12.b.Routine maintenance: The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 2,080 hrs/yr.

Personnel involved in this work category are expected to spend more time in high radiation areas. However, by following maintenance procedures, such as flushing equipment in high radiation areas before performing maintenance and also by removing smaller equipment from high radiation areas to lower radiation areas for maintenance, the dose rate for personnel can be minimized. Consequently, the annual average dose rate for an individual involved in this work function is expected to be about 3.7 rem. The distribution of occupancy time in various radiation zones is listed in Table 12.4-12

.c.Inservice inspection: The total occupancy time for personnel involved in this work function in various radiation zones is expected to be 240 hrs/yr

(Table 12.4-12) at the rate of 40 hrs/wk for 8 wks/yr. The annual average dose for an individual involved in this work function is approximately 0.5 rem.d.Refueling: The refueling work is performed by personnel involved in routine operation and surveillance, radiation protection and rad chemistry, and routine maintenance. The expected occupancy time for personnel involved in the refueling operation is about 160hrs/yr. The average dose per individual in this category is estimated to be 0.5 rem over the occupancy period. The personnel involved in the refueling operation are expected to spend more time in the high radiation areas. The initial crud burst in the refueling pool and fuel storage pool regions will initially result in occupancies in radiation Zone C. However, the operation of the cleanup systems will result in significant reduction of the dose rates in the regions where occupancy is expected. In view of the above considerations, a combined occupancy of 70 percent in radiation Zones B, C, D, and E is considered reasonable. A realistic distribution of personnel occupancies in various radiation zones is provided in Table 12.4-12

.e.Special maintenance: The total occupancy time for personnel involved in special maintenance problems, generally unanticipated, in various radiation zones is expected to be 320 hrs/yr, and the dose rate per CALLAWAY - SP12.4-4Rev. OL-2211/16individual involved in this work function is estimated to be about 0.7 rem/yr. The distribution of occupancy time in various radiation zones is provided in Table 12.4-12

.f.Radwaste processing: The radwaste operations will be performed by personnel involved in work functions such as routine operation and surveillance, radiation protection, rad chemistry, and routine maintenance. The total times expended by personnel in various radiation levels for the category are given in Table 12.4-12. The expected occupancy time for personnel involved in radwaste processing is about 2080 hrs/yr. The annual average dose per individual in this category is estimated to be about 500 mrem.12.4.1.2AirborneRadioactivityDoseEstimates12.4.1.2.1Exposures Due to Airborne Radioactivity As already discussed in Section 12.2.2, negligible airborne concentrations and consequently negligible airborne radio-activity exposures are expected in those areas of the auxiliary, radwaste, and turbine buildings which are accessible (Ref.4). Exposures due to airborne radioactivity are possible in the containment and fuel building both during power operation and refueling. However, the design of the plant operating procedures described in Sections 12.3.3 , 12.3.4, and 12.5.3, and expected limited occupancies in these buildings will minimize exposures to airborne activity and ensure that the doses to an individual from airborne radioactivity are small fractions of the 10 CFR 20 limits for occupational workers and that annual man-rem exposures comply with the ALARA criteria within the plant. The annual man-rem exposures from airborne radioactivity will be a small fraction of the annual man-rem exposures due to direct radiation. Annual occupancy (man-hours), dose rates, and man-rem due to airborne radioactivity in these areas are given in Tables 12.4-14 and 12.4-15, respectively.12.4.1.2.2Model for Calculating Exposures Due to Airborne Radioactive SourcesThyroid and inhalation doses via inhalation pathway are calculated using the following equation:where C i=Airborne concentration of the i th nuclide in pci/cm 3 D oi C i BR ()tDF oi=

CALLAWAY - SP12.4-5Rev. OL-2211/16andTotal body submersion doses are calculated, using a finite cloud model.Annual man-rem exposures due to airborne radioactivity are calculated using the following equation:where and 12.4.1.3IllustrativeExamplesofDoseAssessmentDose assessments for various operations were based on actual operating plant data. A number of typical examples are provided in Table 12.4-1. Note that the maximum/minimum values for number of personnel do not correspond to the maximum/minimum number of days required or dose rates so that the man-rem totals are not simple multiplications of the maximum/minimum factors. The dose assessments in Table 12.4-1 were derived from the average number of personnel, average length of time, and average dose rate to perform each particular operation.a.Operation and SurveillanceThe dose rates in the corridors and other normally occupied areas are expected to be much lower than the maximum radiation levels for each BR =Breathing rate for occupational worker in cm 3/sec=347t=Time duration of inhalation of radioactivity contaminated air in seconds DF oi =Dose factor for adult for organ O (thyroid or lung) via inhalation in mrem/pCi inhaled for the i th isotope (These dose factors are taken from Regulatory Guide 1.109, Rev. 1)

D o=Dose in millirems to organ O due to inhalation(DR)o=Dose rate for organ in mrem/hrh=Annual occupancy in man-hours/yr D o=Annual exposure in man-rem/yr D o DR ()o 10 3-h=

CALLAWAY - SP12.4-6Rev. OL-2211/16zone shown in the radiation zone drawings (Figure 12.3-2). The expected radiation levels are provided in Section 12.4.1.1. Based on these expected dose rates and typical time periods for operation and surveillance, exposures were calculated. Typical examples from operating plants are given in Table 12.4-1. The total exposures for this category range from 13-30 man-rem.b.Routine MaintenanceA number of examples of normal man-rem associated with maintenance are provided in Table 12.4-1. The total number of annual man-rems of exposure associated with this category depend upon many variables, such as equipment run times and breakdowns, number of skilled personnel available, schedule, and crud trapping. Typically, routine maintenance can account for a large percentage of the annual man-rem.c.Radwaste ProcessingAnnual exposures for radwaste processing were determined as a result of the system evaluations described in Section 12.1.2.4. Average expected dose rates and personnel stay times were used. Typical data from operating plants for various tasks are presented in Table 12.4-1

.d.RefuelingBased on operating plant data, refueling operations have required 15 to 30 days, using 8 to 61 personnel who receive from 32 to 347 mrem/day.

Man-rem totals have ranged from 13 to 66. Typical man-rem exposures for individual tasks are provided in Table 12.4-1. Much of the exposure is associated with removal and replacement of the reactor vessel head and its appurtenances.e.Inservice InspectionInservice inspections at operating PWRs have taken much of the refueling outage, requiring from 3 to 38men depending on the inspections scheduled, at up to 2500 mrem/day. Total exposures have ranged from 10 to 24 man-rem. Steam generator eddy current testing, as required by the Technical Specifications, is perhaps the largest single cause of exposure for ISI. Table 12.4-1 provides a listing of various ISI functions and their associated man-rem from operating plant data.f.Special Maintenance CALLAWAY - SP12.4-7Rev. OL-2211/16Special maintenance is generally of a nonrecurring nature and not readily predictable. It includes implementation of design changes and unexpected repair or replacement of equipment and components.Designs are continually being improved so that the newer plants should not experience all of the problems that have occurred on operating plants.

Some special maintenance has created greater than 150 man-rems of exposure, but the frequency of occurrence is irregular. An estimated annual average exposure from special maintenance work is 150 man-rems. Some examples of special maintenance are provided in Table 12.4-1

.Steam generator tube plugging is perhaps the operation causing the largest exposures in the category of special maintenance. The exposures associated with tube plugging are dependent on the number of tubes to be plugged. Remotely operated equipment and explosive plugging will be used to minimize personnel exposures for this task. Improvements in secondary system water chemistry and tube support plates have reduced the likelihood of the need to plug tubes.The volatile chemical treatment to be employed for the secondary system will greatly alleviate the need for sludge lancing of the secondary side.Special materials are being utilized for the radwaste evaporator tubes to maximize corrosion resistance and thereby minimize the maintenance that might be needed.Other special maintenance, such as valve operator or pump impeller replacement, might occur several times in the life of the plant, but exposures would be much less than 150 man-rem.12.4.2EXPOSURES AT LOCATIONS OUTSIDE PLANT STRUCTURES 12.4.2.1DirectRadiationDoseEstimatesDirect radiation to plant personnel outside of plant structures from the containment, auxiliary, radwaste and turbine buildings is negligible. Sources of radioactivity not stored in plant structures are the Old Steam Generator Storage Facility (OSGSF), the Independent Spent Fuel Storage Installation (ISFSI), the ISFSI Support Building, reactor makeup water storage tank, refueling water storage tank, condensate storage tank, and two discharge monitor tanks. A fence is provided around the ISFSI and refueling water storage tank to restrict access due to potential dose rates. Administrative controls will be used to limit access to the ISFSI Support Building if radioactive material is stored there.

CALLAWAY - SP12.4-8Rev. OL-2211/16Dose rate at the nearest site boundary from the original steam generators, the original reactor vessel closure head, the ISFSI with 48 casks loaded, the ISFSI support building, the reactor makeup water storage tank, refueling water storage tank, condensate storage tank, and two discharge monitor tanks has been calculated to be approximately 10

-5 mrem/yr. Using the year 2020 projected populations, a population exposure of less than 10-3 man-rem will be received out to 50 miles in the exposed sectors.12.4.2.2ExposuresDuetoAirborneRadioactivityEstimates of doses at the site boundary due to released activity are given in Section 11.3.3.12.

4.3REFERENCES

1.NUREG-0109, Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1975.2.NUREG-75/032, Occupational Radiation Exposure At Light Water Cooled Power Reactors, 1969-1974.3.NUREG-0463, Occupational Radiation Exposure, Tenth Annual Report, 1977.4.NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station Unit 1.

CALLAWAY - SPRev. OL-215/15TABLE 12.4-1 ILLUSTRATIVE EXAMPLES OF DOSE ASSESSMENTOperation and Surveillance:Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-remFuel bldg.Refueling2910-1136 Aux. bldg.Entire yr5-99-1862Aux. bldg. equip. test1837-650.2Ctmt. (initial survey after SD)1-35-170.1-0.5 Total for oper. & surv.13-30Routine MaintenanceMisc. instrument calibration for pressurizer7319-920.2Aux. bldg. valves19-224-1613-702Aux. bldg. general maintenance8-458-133-2420.3-2 Aux. bldg instrument calibration41115-480.5Ctmt. decon.Refueling3321-1707Fuel bldg. decon.4652-200.3 Radwaste bldg. general maintenance11928-1370.8Radwaste bldg. decon.21-2439-1434Aux. bldg. decon.1143-190.1Ctmt. valves8-7410-2733-1850.9-6Ctmt. instrument calibration8-654-309-750.3-5 Replace oil in reactor coolant pumps5512-450.3Remove charcoal filters for ctmt. cleanup4-57-360.1-0.5Replace charcoal filters for ctmt. cleanup6-452-200.3-0.8Repair dampers & duct84-50.2-0.3General decon. & relamping29-6833-717-11Decon. Refueling canal2-184-230.1-6 CALLAWAY - SPTABLE 12.4-1 (Sheet 2)Rev. OL-215/15Overhaul reactor coolant pump seals21-3018-34Reactor coolant pump motor work3518-306Uncouple/couple reactor coolant pumps2215-190.8Install/remove reactor coolant pump scaffolding127-100.3Repair ctmt. sump pumps & level Indicators9-759-300.4-4Repair pressurizer relief valves5-314-90.2-0.3Replace excore detectors23-4215-301-8Polar crane maintenance9-789-220.3-3Ctmt. pressure test & valve repair6026-322Remove & clean SI system strainers314-250.6Check & repair snubbers3-436-210.1-2Remove equip. hatch5-102-90.002-0.1Replace equip. hatch2-33-70.03-0.04Radwaste ProcessingWaste drum loadingRefueling762-2806Loading spent resin casksRefueling1078-7604Dry waste drum handlingRefueling3-590.2-4 Spent fuel pool filter replacement240.6Transferring resin14-50.1RefuelingInstall and remove reactor cavity seal ring7-167-300.4-0.9Remove transfer tube flange12-172-100.3-0.4Install & remove upper guide structure lifting rig3-316-250.2-3Rig underwater lights2-184-130.02-0.3Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPTABLE 12.4-1 (Sheet 3)Rev. OL-215/15Flood refueling canal45-90.3Fuel handling (ctmt.)14-1649-923-7Fuel handling (fuel bldg.)5-2825-770.5-1Incore instrumentation removal3-3613-242-6

Remove & replace cable to reactor head11-139-270.9-3Remove tool access flanges & uncouple control rods10-139-390.8-9Install bullet noses1-96-150.7-2Remove bullet noses4-52-40.1-0.4Remove & replace head insulation7-104-50.3-0.4RPV studs - remove, replace & clean10-1926-403-13Install & remove alignment pins2-113-120.1-1Connect Polar Crane to Integrated Head Assembly (IHA) tripod & remove RPV head and IHA2-1011-370.6-3Replace RPV head and IHA1-36-180.4-7Decon. PRV head & vessel flanges & replace "O" rings7-1410-243-5Remove & replace holddown ring4-614-313-8Clean stud plugs & holes and install & remove plugs4-1112-261-6Clean incore instrument flanges, studs, & nuts9-2812-322-6RPV head gasket replacement6-3219-351-4Drain, fill, and vent reactor coolant system2-95-100.1-0.7Install handrails around refueling canal146-80.1Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPTABLE 12.4-1 (Sheet 4)Rev. OL-215/15Total for refueling15-308-6132-34713-66Inservice InspectionSteam generator tubes7-5113-3829-25004-22 Spent fuel sipping201714-2423 Spent fuel inspection5722-301-2Remove & replace primary manway covers4-1412-373-7Remove & replace secondary manway covers7-84-160.4-0.8Steam generator secondary side2-411-170.2-2Install tent around primary manway660.9Install primary loop dams231Clean steam generator manway studs64-120.6Total for ISI10-24Special MaintenanceReplace upper guide structure on control rods697817-78513Sludge lancing of steam generator secondary side51619-1052Steam generator tube plugging7-305-1020-200NOTES:(1)The number of days indicates the period over which the particular task was performed.It does not mean that personnel were working full time for that period.(2)The number of personnel indicates the total number of personnel who worked on aparticular task. It does not mean that all of the personnel were working simultaneouslyor full time on that task.Description of TaskNo. of Days (1)No. ofPersonnel (2)Mrem/man-dayNo. ofman-rem CALLAWAY - SPRev. OL-135/03TABLE 12.4-2 AVERAGE NUMBER OF PERSONNEL PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1, 2, 3)YearNumberofUnitsAverageNumberofPersonnel/Unit196921311970249319713250 1972440119735774197410602 197514548197625593197728642 1969-1977Overall average580NOTES:1.Only PWRs at power levels 450 MWe have been considered, with the exception ofSan Onofre 1 (430 MWe)2.Only PWRs that have been in commercial operation for at least 18 months at the endof the year have been considered. Multiple units have been considered for any yearduring which all units have been in commercial operation for at least 18 months priorto the end of the year. 3.Units considered for both tables are the same, with one exception. For the year 1972,Point Beach Unit 1 was used in Table 12.4-3, but not in Table 12.4-2 since no dataon the number of personnel were available for 1972.

CALLAWAY - SPRev. OL-135/03TABLE 12.4-3 AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE) PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1, 2, 3)YearNumberofUnitsAverage Man-Rem Dose/Unit19692741970242219713274 1972548219735610197410439 197514460197625461197728422 1969-1977Overall average441NOTES:1.Only PWRs at power levels 450 MWe have been considered, with the exception ofSan Onofre 1 (430 MWe)2.Only PWRs that have been in commercial operation for at least 18 months at the endof the year have been considered. Multiple units have been considered for any yearduring which all units have been in commercial operation for at least 18 months priorto the end of the year. 3.Units considered for both tables are the same, with one exception. For the year 1972,Point Beach Unit 1 was used in Table 12.4-3, but not in Table 12.4-2 since no dataon the number of personnel were available for 1972.

CALLAWAY - SPRev. OL-135/03TABLE 12.4-4 AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE) BASED ON PWR PLANT AGE (REF 1, 2, 3)Initial Operation DateUnits ConsideredMan-RemperUnit123330 664 571 775 481 341 465 665 745No. of Units 14 15 13 9

8 5

4 3

2 2Year ofOperation(Range of months)(9-18)(21-30)

(33-42)

(45-54)

(57-66)

(69-78)

(81-90)

(93-102)

(105-114)

(117-126)1 2 3

4 5

6 7

8 9

101/68Conn. Yankee-XXXXXXXXX1/68San Onofre 1-XXXXXXXXX3/70GinnaXXXXXXXX 3/71Robinson 2XXXXXXX12/71PalisadesXXXXXX12/72Maine YankeeXXXXX 12/72Surry 1XXXXX 5/73Surry 2XXXXX9/73Fort Calhoun 1XXXX6/74KewauneeXXX 9/74Three Mile Isl. 1XXX12/74Arkansas 1XXX4/75Rancho Seco-XX 8/75CookXX12/75Millstone Point 2XX5/76TrojanX 12/76St. LucieXNOTES:1.Only PWRs operating at power levels > 450 MWe have been considered, with the exception of San Onofre 1 (430 MWe).2.Multiunit plants have been excluded for the most part unless commercial operation dates are close together (example, Surry 1 & 2), or unless the additional units are not scheduled for commercial operation for approximately 2 years or more. (Example, Arkansas 1 & 2, Three Mile Island 1 & 2, St. Lucie, Davis-Besse 1, 2, & 3, Farley 1 & 2, and Salem 1 & 2).

CALLAWAY - SPTABLE 12.4-4 (Sheet 2)Rev. OL-135/033.Exposures reported to be in excess of 1,000 man-rem/year-unit are:ReactorYears of Operationat Year of RecordMan-Rem DosePalisades21,109Ginna31,032 Surry 1 & 243,165Ginna51,224Robinson 251,142 Surry 1 & 252,307 CALLAWAY - SPRev. OL-135/03TABLE 12.4-5 DISTRIBUTION OF THE NUMBER OF PERSONNEL (>100 MILIREMÚYR) ACCORDING TO WORK FUNCTION WorkFunctionPercentage as Per NUREG-75/032 (for the year 1974) (Ref.2)Percentage as Per NUREG-0109 (for the year 1975) (Ref.1)1.Reactor operations19.29.12.Routine maintenance34.559.5 3.Inservice inspection1.44.1 4.Special maintenance28.716.4 5.Waste processing2.17.6 6.Refueling14.13.3 See notes at end of Table 12.4-7

.

CALLAWAY - SPRev. OL-135/03TABLE 12.4-6 DISTRIBUTION OF PERSONNEL (>100 MILLIREM/YR) ACCORDING TO EMPLOYEE CATEGORY CategoryPercentage as Per NUREG-75/032 (for the year 1974)(Ref2)Percentage as Per NUREG-0109 (forthe year 1975) (Ref1)1.Station employees47.434.72.Utility employees18.16.1 3.Contract workers34.559.2 See notes at end of Table 12.4-7

.

CALLAWAY - SPRev. OL-135/03TABLE 12.4-7 PERCENTAGES OF PERSONNEL DOSE BY WORK FUNCTION (REF3) WorkFunctionPercentofDose1974197519761977Reactor operations and surveillance14.0%10.8%10.2%10.6%Routine maintenance45.4%52.6%31.0%28.9%In-service inspection2.7%3.0%6.0%6.6%

Special maintenance20.4%19.0%40.0%41.4%

Waste processing3.5%6.9%5.0%5.9%

Refueling14.0%7.7%7.9%6.6%

Notes on Tables 12.4-5 , 6 and 7:1.PWRs and BWRs operating at all power levels were considered in compiling thesetables.2.Percentages for 1974 and 1975 are based on approximately 39 percent and 50percent of the total exposures reported in the appropriate year for light water reactors which had been in commercial operation for at least one full year, as of 12/31/74 and 12/31/75, respectively.3.Percentages for 1976 and 1977 are on only those facilities which have been incommercial operation for at least one full year, as of 12/31/76 and 12/31/77, respectively.4.Distributions of personnel receiving an annual exposure of greater than 100millirems, according to either work function or employee category, are not availablefor 1976 or 1977.

CALLAWAY - SPRev. OL-135/03TABLE 12.4-8 ANNUAL OCCUPATIONAL EXPOSURES FOR VARIOUS PWR VENDOR'S UNITS (REF. 1,2,3)Notes - On Table 12.4-8:1.Only PWRs operating at power levels >450 MWe have been considered with the exception of San Onofre 1 (430 MWe).2.The power plants for this table are the same as those appearing on Table 12.4-4

.WestinghouseC-EB&WApprox Yr. ofOperationNo. of ReactorsMan-Rem Yr-UnitNo. of ReactorsMan-Rem Yr-UnitNo. of Man-Rem Reactors Yr-UnitComments171495117247---------------2824345223 211CE: Palisades - 1,1313760834193 335W: Ginna - 1,032 CE: Palisades produces very little power.467413229------W: Surry 1 & 2 - 3,165 568762471------W: Ginna 1 - 1,224Robinson 2 - 1,142 Surry 1 & 2 - 2,307 645531100---------------------74341---------------------------83443---------------------------92665------------W: San Onofre - 880 102745manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------manrem-yr-----------------------------

CALLAWAY - SPRev. OL-135/03TABLE 12.4-9 AVERAGE ANNUAL OCCUPATIONAL EXPOSURE AND POWER FOR INDIVIDUAL PWRS (REF 1, 2, 3)Plant NameTotal Yrs ofOperation(approx.)Full PowerMWe(net)Annual AveragePercentageof Full PowerAnnual Ave.Man-RemCommentsConn. Yankee1057579462First year data not avail.1973 - only 50 percent of full power was producedSan Onofre11043077325First year data not avail.Highest annual exposure: 1976 - 880 man-rem 1977 - 847 man-remGinna849064587Percentage of full power (about52%, 51%) was low in 1974 and 1976. Highest annual exposures: 1972 - 1,032 man-rem 1974 - 1,225 man-remRobinson 27665756071971 - Only 53 percent of fullpower was produced. Highest annual exposure 1975 - 1,142 man-remPalisades679843.4*490*Average per 5 yrs 1974 - Only produced 1 percent full power. 1972 - operated at restricted levels - 27 percent full power produced. Highest annual exposure 1973 - 1,133 man-rem.Maine Yankee579068237Percentages of full power producedwere low in 1973 (52 percent) and 1974 (55 percent)Fort Calhoun 14457632441975- Only 55 percent of full power was produced.

CALLAWAY - SPTABLE 12.4-9 (Sheet 2)Rev. OL-135/03Notes on Table 12.4-9:1.Only single unit PWRs have been considered.2.Full power data based on information given in "Commercial Nuclear Power Plants" by NUS Corp., 1978, Ed. 10.

3.Robinson full power is based on a summertime value.4.Only plants which have been in commercial operation for equal to or greater than 4 years were considered.

CALLAWAY - SPRev. OL-135/03TABLE 12.4-10 AVERAGE INDIVIDUAL EXPOSURE BASED ON PWR PLANT AGE (REF. 1,2,3)Notes on Table 12.4-10:1.Units considered are the same as those shown on Table 6, except that Palisades has been omitted from the first year average since no data is available for the number of personnel for that year.2.The high exposure average for years 3-5 include the following:Year of Operation (approx)No. ofUnitsAve. No. of Personnel/UnitAve. Exposure man-rem yr-unitAve. Exposure per Individual (rem-yr)1133621230.342155323300.62 3136616641.00 496235710.92 586917751.12 656274810.77 745403410.63 835834650.80 929876650.671029407450.79PlantYearGinna Average ExposureSurry 1 & 2 Average ExposureRobinson Average Exposure31.52------4---1.15---

51.391.241.35Rem/yrRem/yrRem/yr CALLAWAY - SPRev. OL-135/03TABLE 12.4-11 CUMULATIVE AVERAGE OF ANNUAL EXPOSURE BY YEARS OF OPERATION - PWRS (REF. 1, 2, 3)Note: 1.Plants considered are the same as those shown on Table 12.4-4

.Years of OperationPlant12345678910Conn. Yankee-106398379366432394438439462San Onofre 1-429982126171155174226327Ginna207319556478627613616589 Robinson 2364290425484618634608Palisades78606613536568490Maine Yankee117269285235237 Surry 1 & 21525188954631631Fort Calhoun 171183226244Kewaunee28149146 Three Mile Island 173180240Arkansas 121155189Rancho Seco-58225 Cook116208Millstone Pt 2168205Trojan174 St. Lucie152Total17213288429739044173234017731201701789No. of Units1415139854322 Average123219331434522468433400351395 CALLAWAY - SPRev. OL-135/03TABLE 12.4-12 ESTIMATES OF OCCUPANCY TIMES IN PLANT RADIATION AREAS AND GAMMA DOSES TO PLANT PERSONNELOperationZonePercentage ofOccupancyHrs/yrHourly DoseRate(Rems/Hr)Yearly DoseRate(Rems/Yr) No. of men Annual Exposures(Man-Rem/yr-Unit)Operation & SurveillanceA751,560 1 x 10-40.15385.7B21437 5 x 10-40.21388.3C3622.5 x 10-30.15385.9D0.9191.5 x 10-20.283810.8E0.12 1 x 10-10.20387.6Total1002,0800.9938.3Routine MaintenanceA751,560 1 x 10-40.15284.4B11229 5 x 10-40.11283.2C10.52182.5 x 10-30.542815.3D2.5521.5 x 10-20.782821.8E121 1 x 10-12.12858.8Total1002,0803.68104Radwaste ProcessingA701,456 1 x 10-40.1481.16B20416 5 x 10-40.280.68C81662.5 x 10-30.4183.28D1211.5 x 10-20.3182.52E121 1 x 10-12.1816.8Total1002,0803.1624.4RefuelingA3048 1 x 10-40.00460.22B4064 5 x 10-40.0361.47C24392.5 x 10-30.0964.48 CALLAWAY - SPTABLE 12.4-12 (Sheet 2)Rev. OL-135/03OperationZonePercentage ofOccupancyHrs/yrHourly DoseRate(Rems/Hr)Yearly DoseRate(Rems/Yr) No. of men Annual Exposures (Man-Rem/yr-Unit)D461.5 x 10-20.0964.14E23 1 x 10-10.30613.8Total1001600.5124.1Inservice Inspection (ISI)A55132 1 x 10-40.01460.6B3686 5 x 10-40.04461.96C7.0172.5 x 10-30.04461.96D1.02.51.5 x 10-20.03461.72E12.5 2 x 10-10.54623Total1002400.6229.2Special MaintenanceA75240 1 x 10-40.0242004.8B2064 5 x 10-40.032006.4C3102.5 x 10-30.0252005D131.5 x 10-20.042009E13 2 x 10-10.6200120Total1003200.71145 CALLAWAY - SPRev. OL-135/03TABLE 12.4-13 DISTRIBUTION OF DIRECT RADIATION MAN-REM DOSES ACCORDING TO WORK FUNCTIONSOperationAnnual Exposures (Man-rem/year Unit)PercentageOperation and surveillance38.310.5Routine maintenance10428.5 Radwaste processing24.46.7 Refueling24.16.6 Inservice inspection29.28 Special maintenance14539.7 Total365100 CALLAWAY - SPRev. OL-135/03TABLE 12.4-14 ANNUAL OCCUPANCY IN PLANT AREAS CONTAINING AIRBORNE RADIOACTIVITYBuilding hr/yrper manNo. menman-hours yrContainment-power2503750Containment-refueling62.5322,000Fuel building-power2502500Fuel building-refueling125141,750 CALLAWAY - SPRev. OL-135/03TABLE 12.4-15 DOSES TO PLANT PERSONNEL CAUSED BY AIRBORNE RADIOACTIVITYLocationDose Rate mrem/hrAnnual Occupancy (man-hr/yr)Annual Expose (man-rem/yr)ThyroidLungWhole BodyOperation& SurveillanceMaintenanceRefuelingTotalThyroidLungWhole BodyContainment - power8.640.050.48100650---7506.480.040.36Containment - refueling1.340.93N---1000100020002.681.86NFuel building - power0.620.62N400100---5000.310.31NFuel building - refueling4.713.28N------175017508.255.74NTotal1880.4N = Negligible CALLAWAY - SP12.5-1Rev. OL-155/0612.5RADIATION PROTECTION PROGRAM For descriptions of the radiation protection program, refer to Section 12.5 of the Site Addendum.

CALLAWAY - SA12.0-iTABLE OF CONTENTSCHAPTER 12.0RADIATION PROTECTION Section Page12.1ENSURING THAT OCCUPATIONAL RADIATION EXPOSURE S ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA).........

................

........................12.1-112.1.1POLICY CONSIDERATIONS.....

..............

..............

..............

...................12.1-112.1.1.1Management Policy............

..............

..............

...............

.....................12.1-112.1.1.2Organizational Structure...

................

..............

...............

.....................12.1-112.1.1.3Application of the Un ion Electric ALARA Policy to Plant Operations.12.1-112.4DOSE ASSESSMENT................

................

..............

...............

.....................12.4-112.5RADIATION PROTECTION PROGRAM..................

...............

.....................12.5-112.5.1ORGANIZATION.......

..............

..............

..............

...............

.....................12.5-112.5.1.1Radiation Protection Program Objectives...........

..............

.................12.5-112.5.1.2Organization of the R adiation Protection Department

........................12.5-212.5.1.3Personnel Training.........

................

................

...............

.....................12.5-212.5.2EQUIPMENT, INSTRUMENTATION AND FACILITIES.....

.....................12.5-212.5.2.1Equipment and Instrumentation..............

.................

..........................12.5-212.5.2.1.1Installed Radiation Monitoring Equipment...........

..........................12.5-312.5.2.1.2Radiation Protection Laboratory Instrumentation..

........................12.5-312.5.2.1.3Portable Radiation Detection Instrumentation.....

..........................12.5-312.5.2.1.4Portable Air Sampling Equipment............

.................

.....................12.5-312.5.2.1.5Instrument Calibration..

...............

..............

...............

.....................12.5-412.5.2.1.6Protective Clothing..................

................

.................

.....................12.5-412.5.2.1.7Respiratory Protection Equipment.........

................

........................12.5-512.5.2.1.8External Dosimetry......

................

..............

...............

.....................12.5-512.5.2.1.9Internal Dosimetry.....

................

................

...............

.....................12.5-512.5.2.2Facilities...........

................

................

..............

...............

.....................12.5-612.5.2.2.1Radiation Protection Facilities...............

................

........................12.5-612.5.2.2.2Access Control Facilities

.............

..............

...............

.....................12.5-612.5.2.2.3Equipment Decontamination Facilities................

..........................12.5-612.5.2.2.4Personnel Decontamination Facilities.......

...............

.....................12.5-712.5.2.2.5Radiation Protection Count Room......

.................

..........................12.5-7 CALLAWAY - SATABLE OF CONTENTS (Continued)

Section Page12.0-ii12.5.2.2.6Laundry Decontamination Facility ....

..............

..............

.................12.5-712.5.2.2.7Radiation Protection Office...............

..............

..............

.................12.5-712.5.2.2.8Locker and Change Facilities...........

..............

..............

.................12.5-712.5.2.2.9Calibration Facilities...

.................

..............

...............

.....................12.5-712.5.2.2.10Dosimetry Processing Facilities........

..............

..............

.................12.5-812.5.2.2.11Respiratory Protection Facilities............

................

........................12.5-812.5.2.2.12Respirator Issue and Storage Area...................

............

.................12.5-812.5.3PROCEDURES............

................

................

.................

..........................12.5-812.5.3.1Radiation Surveys..........

................

................

...............

.....................12.5-812.5.3.1.1Radiation - Surveys.....

................

..............

...............

.....................12.5-812.5.3.1.2Surface Contamination Surveys............

................

........................12.5-912.5.3.1.3Airborne Radioactivity Surveys..............

................

........................12.5-912.5.3.2Radiation Exposure Control.

.................

..............

.............................12.5-1012.5.3.3Contamination Control......

................

..............

...............

...................12.5-1212.5.3.4Radiation Protection Training..............

..............

..............

.................12.5-1212.5.3.5Personnel Dosimetry........

................

..............

...............

...................12.5-1312.5.3.5.1External Radiation Dosimetry................

................

......................12.5-1312.5.3.5.2Internal Radiation Dosimetry............

..............

.............................12.5-1312.5.3.6Airborne Radiation Evaluation and Control..............

........................12.5-1312.5.3.7Respiratory Protection Program.................

................

......................12.5-1312.5.3.8Radioactive Materials Handling.............

..............

.............................12.5-14 CALLAWAY - SA12.0-iiiRev. OL-1412/04LIST OF TABLES NumberTitle12.5-1Radiation Protecti on Laboratory Equipment12.5-2Portable radiati on protection Equipment CALLAWAY - SA12.0-ivRev. OL-1412/04LIST OF FIGURES NumberTitle12.5-1Deleted CALLAWAY - SA12.1-1Rev. OL-21 5/15CHAPTER 12RADIATION PROTECTION12.1ENSURING THAT OCCUPATIONAL RADIATION EXPOSUR ES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA)12.1.1POLICY CONSIDERATIONS12.1.1.1Management PolicyUnion Electric Company is co mmitted at the corporate level to maintaining occupational radiation as low as reasonably achievable.

This policy is promul gated in general terms by the Senior Vice President and Chief Nuclear Officer and is implemented by specific procedures and directives at the plant level by the Senior Director, Nuclear Operations. Implementation of the ALARA program will be in accordance with the requirements of 10CFR19 and 10CFR20 and will follow the intent of the guidance prov ided in applicable sections of Regulatory Guides 1.8, 8.8 and 8.10.12.1.1.2Organizational StructureThe Senior Director, Nuclear Operations, th rough the Radiation Protection Department staff, is ultimately responsible for development, implementation, and continuing conduct of the overall plant Radiation Protection program of which the ALARA program is a part. The Manager, Radiation Protection directs the Radiation Protection Department and is responsible to the Plant Di rector for the ALARA program. The Manager, Radiation Protection has the responsi bility and authority for de veloping, impl ementing and managing the plant ALARA program.

The organization of the Radiaton Protection Department is described in Section 12.5.1

.ALARA reviews can be performed by qualif ied nuclear engineeri ng personnel (i.e., Design Engineering, Technica l Support Engineering, Systems Engineering, or Radiation Protection) for plant modifications or maintenance activities.12.1.1.3Application of the Union Electric ALARA Po licy to Plant Operations The Callaway Plant ALARA program is implemented throug h the adherence to Administrative and Radiation Pr otection procedures which cover, but are not limited to, the following subjects:a.Qualifications of and the conduct of operations of the plant radiation protection department.b.ALARA program administration - re viewing job planni ng, job exposure data, radiation work permit comparisons wit h personnel exposure, CALLAWAY - SA12.1-2Rev. OL-21 5/15 procedure reviews, review of design change packages for installation and long term operations and maintenance concerns, and trends of dose data for potential progr ammatic changes.c.Training of personnel in the fundamentals of ra diation protection and in health physics exposur e control procedures.d.Radiation Work Permit Programe.Internal Pers onnel Monitoringf.External Personnel Monitoring g.Area Postingh.Radioactive Material Controli.Radiological Surveys j.Instrumentationk.Radiological Incidentsl.Radiation Work Practices m.Environmental Monitoringn.Respiratory Protection CALLAWAY - SA12.4-1Rev. OL-13 5/0312.4DOSE ASSESSMENTThis section contained the dose assessment to the constructi on workers during the construction of Callaway Plant Un it 2. With the ca ncellation of Callaway Plant Unit 2 in October 1981, this secti on is no longer applicable.

CALLAWAY - SA12.5-1Rev. OL-21 5/1512.5RADIATION PROTECTION PROGRAM12.5.1ORGANIZATION12.5.1.1Radiation Protecti on Program ObjectivesThe objective of the radiation protection program for the Callaway Plant is to maintain the occupational radiation exposure of personnel working at the pl ant as low as is reasonably achievable (ALARA). This objecti ve is accomplished by adherence to the requirements of Title 10 of t he Code of Federal R egulations, Part 20, by following the intent of guidance given in USNRC Regulatory Guides 8.2, 8.8, 8.10, 8.34, 8.35, 8.36 and 1.8, and by using industry acce pted radiation protection practices.

The radiation protection pr ogram is administered by t he organization described in subsections 12.5.1.2 and 13.1.2.2.2 to accomplish the following:a.Identify and review radiation protection training requirements for personnel assigned to work in radiological cont rolled areas commens urate with their duties, responsibilities, and the degree of radiation hazards anticipated.

Inform these personnel of methods for maintaining occupational radiation exposure ALARA and assist them in carrying out t heir radiation safety responsibilities.b.Evaluate and review the radiological status of the plant by monitoring radiation, contamination and airborne radioactivity levels to control or eliminate radiological hazards.c.Control external and in ternal radiation exposure of perso nnel through the implementation of radiological controls during operations and maintenance.d.Review and evaluate radiation prot ection records and appropriate plant operating and maintenance procedures fo r methods to reduce radiation exposure to personnel.e.Maintain and evaluate personnel exposure reco rds to ensure that occupational radiation exposures are maintained ALARA.f.Maintain control of radioactive mate rials on-site and maintain releases of radioactive materials in effluent s to unrestricted areas ALARA.g.Evaluate and review t he swipe samples from instruments containing radioactive sources that other Union Electric fossil power plants have provided.

CALLAWAY - SA12.5-2Rev. OL-21 5/1512.5.1.2Organization of the Radiation Protection Department The experience and qualificatio ns of the personnel respons ible for administering the radiation protection pr ogram are presented in Section 13.1.3.1. The responsibilities of the personnel are discussed in Section 13.1.2.2.2

.The Radiation Protection Department is comprised of personnel in the following classifications:a.Manager, Radiation Protectionb.Supervising Health Physicist c.Health Physicists/Nuclear Scientists d.Radiation Protection Supervisorse.Radiation Protection Techniciansf.Radiation/Chemical Helpers 12.5.1.3Personnel Training The training of plant per sonnel in radiation prot ection is described in Section 12.5.3.4. 12.5.2EQUIPMENT, INSTRUMENTATION AND FACILITIES12.5.2.1Equipment and InstrumentationRadiation Protection instrumentation and equipment will be availabl e for the assessment of plant radiological conditions and to support the oper ation of the Ca llaway Plant Radiation Protection Program.

The three classifications of equipment that will be utilized in radiological monitoring and su rveillance activities include:a.Installed Equipmentb.Laboratory Equipmentc.Portable EquipmentInstalled equipment includes bot h the in-plant area radiat ion monitoring system and the process and effluent radioacti vity monitoring system. Laboratory equipment consists of analytical instrumentation used to analyze and quantify radioactivity. Portable equipment includes portable radiation detection instrumentation, ai r samplers and personnel dosimetry used for performing radiation and contaminat ion surveys, airborne radioactivity monito ring and sampling and personnel monitoring.

CALLAWAY - SA12.5-3Rev. OL-21 5/15Portable and laboratory instrumentation used in the radiation protecti on program will be selected in accordance with the following criteria:a.Ability to measure the quantity of interest to an acceptable degree of precision and accuracy.b.Ease of operation, ma intenance, and calibration.c.Appropriate sensitivity and range for various operati onal situations, including normal operations, anticipated operati onal occurrences and accident conditions, as determined by the requirements of applicable regulations.d.Operational reliability.12.5.2.1.1Installed Radiat ion Monitoring EquipmentInstalled area radiation monito rs and continuous air monitors are described in Section 12.3. Process and effluent monitors are described in Section 11.5.12.5.2.1.2Radiation Protection Laboratory Instrumentation Radiation protection laboratory instrumentation will in clude a gamma spectroscopy system, a liquid scintillation counter, a gas flow pr oportional c ounter and a geiger-mueller type counting system. Additional info rmation on radiation protection laboratory equipment is provided in Table 12.5-1. The primary loca tion of radiation protection laboratory equipment will be the count room, 1984' elevati on of the Control Building. However, additi onal counting equipment will be available at other plant locations including the Emergency Operations Facility. 12.5.2.1.3Portable Radiation Detection InstrumentationPortable radiation detection instrumentation is described in Table 12.5-2 which includes information on the type, range, accuracy and typical quantities of instruments. Survey instruments will be available in-plant to facilitate access. Instruments requiring calibration or repair will be identified as out-of-service and segregated from in-use equipment.12.5.2.1.4Portable Ai r Sampling EquipmentPortable air sampling equipment includes high and low-volume air samplers capable of accepting particulate filters and charcoal cartridges for grab samp les of radioactive particulates and iodine. In addition, portable continuous air monitors, typical of that described in Table 12.5-2 , will be used for continuous surveill ance of airborne radioactivity levels. Air sampling instruments are calibrat ed periodically in accordance with an established calibration program.

CALLAWAY - SA12.5-4Rev. OL-21 5/1512.5.2.1.5Instrument CalibrationCalibration of portable and laboratory radi ation protection instrumentation will be performed in accordance with an established calibration program. Requirements of this program include as a minimum:a.written procedures for operation a nd calibration of each instrument.b.a mechanism for tracking and recall of instruments for calibration.c.established calibration frequencies fo r each instrument and recalibration following maintenance and repair that could change the instrument performance characteristics establishe d during the prev ious calibration.d.use of calibration standar ds which are traceable to the National Institute of Standards and Tec hnology (NIST).e.required calibration a ccuracies and tolerances.f.periodic response checking of instruments to verify continuing proper operation.g.maintenance of records document ing calibration activities.h.tagging and labeling of instruments.12.5.2.1.6Protective Clothing Protective clothing is prescribed and issued to plant personnel by the radiation protection staff based upon the actual or potential radiological condit ions expected for the job assignment. Protective clothing stations are established at strategic loca tions within the plant as required to ensure efficient operations and to preclude the spreading of contamination. Protective clothing avai lable at the plant in cludes the following:a.Coveralls.b.Caps and hoods.c.Shoe covers.

d.Plastic, rubber, and cotton gloves.

e.Plastic suits.f.Face shields/masks CALLAWAY - SA12.5-5Rev. OL-21 5/1512.5.2.1.7Respiratory Protection Equipment Respiratory protection equipment is available to plant personnel and is also prescribed and issued to individuals as required by actual or potentia l radiological conditions of the work assignment. The Callaway Plant Resp iratory Protection Pr ogram follows the guidance of Regulatory Guide 8.15 and complies with the requirements of 10 CFR Part

20. Respiratory devices av ailable at the Callaway Plant include th e following:a.Full-face respirators with high-efficiency partic ulate and/or charcoal filters.b.Full-face respirators with supplied air.c.Hoods with supplied air.d.Full-face respirators in self-contained breathing apparatuses. e.Half-face respirators with HEPA and/or charcoal filters.Purchases of respiratory protection equipment such as air purifying respirators, supplied air respirators, self-contained breathing apparatuses and accessory equipment are made following the guidance given by 30 CFR Part 11 and the NIOSH Certified Equipment Manual.12.5.2.1.8External Dosimetry

All personnel entering the plant's Radiological Controll ed Area (RCA), are monitored for occupational radiation dose in accordance with the requirements of 10CFR20. Optical Stimulated Luminescent Dosi metery (OSL) is the primar y method of monitoring occupational radiation dose from exposure to beta and photon radiat ion. CR-39 is the primary method of monitoring occupational radiation dose from ex posure to neutron radiation. Primary dosimetry is processed by a vendor that is accredited by the National Voluntary Laboratory Accr ediation Program (NVLAP) in accordance with the requirements of 10CFR20. An appropriate secondary moni toring device is issued to each individual who enters the RCA. Portable neutron survey instruments and stay time calculations are used as a secondary method for determi ning neutron dose. Exposure records for each individual will be maintained in accor dance with Regulatory Guide 8.7. 12.5.2.1.9Inter nal Dosimetry Internal radiation exposure is assessed using either in vi vo counting, sp ecimen analysis or calculational techniques based on surveillance program data. The methodology for internal radiation exposure assessment will follow the guidance of the International Commission on Radiological Pr otection (ICRP). The concepts, models, equations and assumptions used for internal radiation exposure assessment will reflect th e appropriate recommendations of the ICRP.

CALLAWAY - SA12.5-6Rev. OL-21 5/15An in vivo counter is located at the plant for measurement of plant per sonnel, visitors, or support personnel. The in vivo counter will provide prelim inary background information, periodic evaluation, and emergency capability for detecting internal exposure conditions. Assessment of internal radiation exposure of those individuals who regularly enter areas where the potential exists for inhalation, inges tion, or absorption of radioactive material will be performed annually.12.5.2.2Facilities12.5.2.2.1Radiation Protection Facilities The radiation protection facili ties consist of a radiation protection office, counting room, hot laboratory, and personnel decontamination shower lo cated at elevation 1984' of the Control Building.12.5.2.2.2Access Control Facilities Radiation protection facilities at the 1984' level of the Cont rol Building are designed to function as the primary access point for entry to the radiological c ontrolled areas of the plant. The plant arr angement is such that traffic between radiological controlled areas and uncontrolled areas is routed through this access cont rol area. The purpose of the access control point is to prov ide positive control over acce ss to radiologica l controlled areas for exposure control purposes and to prevent the spread of contamination to uncontrolled areas of the plant. Contamination control features of access control include controlled entrance and exit locations with contamination m onitoring provisions at the exit of the radiologi cally controlled area.

In addition to the pr imary access control point described above, auxiliary radiological control points can be establis hed at necessary loca tions within the pl ant for the purpose of personnel and contaminati on control. These radiological control points are equipped with protective clothing, portable survey equipment and other radiation protection materials and are established at locations of strategic importance for contamination and exposure control.12.5.2.2.3Equipment Decontamination Facilities Equipment used for the cleaning and maintenance of contaminated parts, instruments and equipment is located at elev ation 2000' adjacent to the Auxiliary Building. When hand decon is not practical (such as for high level of contam ination or ALARA constraints), special cleaning equipmen t will be used for removing radioactive contamination from items requiring repair or maintenance.

CALLAWAY - SA12.5-7Rev. OL-21 5/1512.5.2.2.4Personnel Decontamination FacilitiesA personnel decontamination area separated from normal showerin g facilities is provided on the 1984' leve l of the Control Building in the access control area. Facilities provided include decontamination sinks, shower, and a drying and monitoring area.12.5.2.2.5Radiation Pr otection Count RoomA counting room is located near access control and is used for radioactivity analyses. A separate hot laboratory is provided for sample preparation. The count room is equipped to perform routine analyses required for per sonnel protection, su rveys, and related radiation protection functi ons. The counting room is equipped with the necessary instrumentation to perform r outine counting on plant radioactivity samples (water, air, swipe survey, etc.). 12.5.2.2.6Laundry Decont amination Facility The laundry decontamination fa cility is located on the 2 000 elevation adjacent to the Auxiliary Building. It is equipped with washer-extractors, dryers, a clean laundry monitor and clothes sorting tables.12.5.2.2.7Radiation Protection OfficeThe radiation protection office is located adjacent to the entrance to the access control area and will allow radiati on protection personnel to obse rve personnel entering and leaving the radiological controlled area. The radiation protection office provides work space for radiation protection personnel to perform routine administrativ e activities such as completing survey records and generating RWP'

s. It also provi des a location from which plant radiation protec tion activities can be coordi nated and plant personnel can obtain information concerning work activi ties or plant radi ological conditions.12.5.2.2.8Locker an d Change Facilities Change areas consisting of lockers and ben ches are provided with in the plant. A storage area for protective clothing is prov ided adjacent to t he change areas, thus allowing personnel to dress-out for activities near the contaminated areas.12.5.2.2.9Calibration Facilities Instrument calibration faciliti es are provided for the calibration, rout ine maintenance and storage of portable radiation protection instrumentatio
n. The radiation protection calibration facilities will be equipped with fixed and portable calibr ation sources of sufficient range and strength to allow calibration of most portable radiation detection instruments and personnel dosimetry devic es used at the Callaway Plant.

CALLAWAY - SA12.5-8Rev. OL-21 5/1512.5.2.2.10Dosimetry Pr ocessing Facilities Primary dosimetry is processed by a vendor.12.5.2.2.11Respiratory Protection Facilities On-site facilities se t aside for respiratory protec tion functions in clude areas for conducting quantitative fit te sting and an area for adminis tering pulmonary function testing as part of the medical qualific ation program for respirator users.

Other on-site areas ar e set aside for washing, drying, repair and storage of respirators which are used for supporting day to day work. The plant wa rehouse serves as the bulk storage point for receiving and storing respiratory protec tion equipment held in reserve for future use.12.5.2.2.12Respirator Issue and Storage Area A Respirator Issue Area is provided within the Access Control Area for issuing of respirators to individua ls as required by actual or potent ial radiological conditions of the work assignment.12.5.3PROCEDURES The radiation protection administrative, departmental and technical procedures used at the Callaway Plant are an integral part of the plant ALARA program. The provisions and guidance of Regulatory Guides 8.2, 8.7, 8.8, 8.10, 8.13, 1.8, and 1.39, are used in the development of the plant radiation protection procedures. Exception to Reg. Guide 1.33 has been submitted under separate correspondence to the NRC.

Procedures for personnel radiat ion protection shall be prep ared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving pers onnel radiation exposure.12.5.3.1Radiation Surveys Radiological surveys are per formed periodically during operation and shutdown at the Callaway Plant for the assessm ent of radiation-fields, radioactive contamination, and airborne radioactivity levels. Portable instruments, equi pment, and techniques for surveys are addressed in depart ment and technical procedures.12.5.3.1.1Radiation - Surveys Routine measurements of radiation field intensities are performed in accessable areas of the plant, using portable instrumentation appropriate for the type of radiation present.

These surveys are performed on a schedule which is determined by:

CALLAWAY - SA12.5-9Rev. OL-21 5/15a.Actual or potentia l radiation levels.b.The variability of radiation level.c.The occupancy fact or of the location.

Radiation surveys are con ducted to monitor and detect any significant changes in radiation levels and to evaluate the effectiveness of radiolog ical controls. High radiation areas are usually surveyed upon entry and periodically thereafte r while work in the area is in progress. Additional radiation surveys are performed as necessary to evaluate and minimize personnel radiatio n exposure during operationa l and maintenance functions.Records are maintained of the results of these surveys, by location, so that trends in the radiation level are readi ly identified. Result s of these radiation surveys are correlated with the readings of the Area Radiation Monitoring System, where appropriate.

Prior to the initiation of any operation for which a radiation work permit is required, a survey will be made of the r adiation field in the vicinity where the operat ion is to be performed. The results of this survey are recorded on the r adiation work permit.12.5.3.1.2Surface Contamination SurveysContamination surveys are per formed on a regularly schedul ed basis in all accessible areas during operation and shutdown to evaluate the hazard due to removable and non-removable radioactive cont amination. Locations of importance for controlling the potential spread of contaminat ion are surveyed at a frequ ency commensurate with the contamination hazard present using the "smear" technique or an appropriate portable instrument. These survey frequencies could increase or decrease depending on factors such as the actual or potential radioactive concentration, occupancy factor, location, and plant status.Contamination surveys are also made on personnel, equipment, and materials from time to time as necessary to ensure complete control over the levels and spread of removable contamination. Appr opriate techniques, instruments, and frequen cy of surveys will be delineated in the radiati on protection procedures.The results of these surveys are recorded and tabulated by location so that trends in the data may be readily observed.12.5.3.1.3Airborne Radioactivity Surveys Surveys to assess airborne r adioactivity levels are perform ed to ensure 10 CFR Part 20 limits are not exceeded, engine ering controls are functioning, and resp iratory protection techniques are adequate.

CALLAWAY - SA12.5-10Rev. OL-21 5/15The results of routin e airborne radioactivity surveys ar e recorded by location so that trends in the radiation leve ls can be readily identified.

Air sampling surveys are performed on a routi ne basis in areas with both a high occupancy factor and a high potential for the existence of airborne radioactivity. Ot her locations are surveyed periodically depending on the significance of the location as a source of occupational radiation exposure from airborne radioactivity. These survey frequencies could increase or decrease depending on factors such as the actual or potentia l radioactive concentration, occupancy factor, location, and plant status.

Airborne samples are collect ed and analyzed for operations and maintenance activities which are expected to produce airborne radioactivity.12.5.3.2Radiation Exposure ControlIn order to control occupa tional radiation exposures and the spread of radioactive contamination, varyin g degrees of access to plant areas will be established. Area access and exposure time for personnel will be determined by the necessity for such access and by radiation and contamination levels.Areas of the plant are subject to access control restrictions proprortionate to the potential for radiation exposu re in each area. The particular access control requirements for each area will be specified by Radiation Protection Supervision. A radiological controlled area is any area where actual or potential radiological hazards exis t in the form of radiation, contamination, airborne radioactivity, or stored r adioactive materials.

Radiation areas, contamination areas, high radiation areas, very high radiation area, airborne radioactivity areas and radioactive material areas will be cons idered radiological controlled areas.

Radiological controlled areas will be posted and controlled in accordance with the requirements of 10CFR20. A ccess to such areas is cont rolled as warranted by the degree of radiological hazard involved via use of access control, auxiliary radiological control points a nd the radiation work permit program.

The plant arrangement is such that traffi c between radiol ogical controlled areas and uncontrolled areas is r outed through access c ontrol. The access control facilities for normal traffic are located in the control building where peopl e entering and leaving all controlled areas will check for contamination. Within t he controlled area, access is controlled by utilizing lock ed and/or annunciated doors and gates, fences, alarms, rope barriers, and the posting of signs.

High radiation areas in whic h the radiation field is greater than 100 mrem/hr but less than 1000 mrem/hr wil l be posted as a high radiation area and entrance into the area will be controlled by the radiation work permit program. High radiation areas in which the radiation fiel d is greater t han or equal to 1000 mr em/hr will be provided with locked doors to prevent unauthorized entry.

For individual areas accessible to personnel that are lo cated within large areas, such as the Containment, where no enclosure exists for purpose of locking , and where no enclosure can be reasonably constructed, the individual area shall be barricaded, conspicuously posted and a flashing CALLAWAY - SA12.5-11Rev. OL-21 5/15 light activated as a warning de vice. The flashing light may be omitted if positive control of access to the area is provided.Very high radiation areas with radiation levels in excess of 500 rads in one hour at one meter from the radiation sour ce will be conspicuously posted as a very high radiation area. Entrances to these areas will be barr icaded and locked.

Administrative measures for access cont rol in high radiat ion areas include the requirement for issuance of a radiation work permit for any operation to take place in such an area. Depending upon the operation, a radiation protection technician may be assigned to supervise stay-tim e and make appropriate surve ys while the operation is in progress.The radiation work permit (RWP) program will be established as an integral part of the ALARA policy implementation.

This program will be the responsibility of the Manager, Radiation Protection. In addi tion to entries into and work performed in high radiation areas, RWP's may be required for the following type activities:a.Entries into designated airborne radi oactivity areas, r adiation areas, contamination areas and radioacti ve material storage areas. b.Work involving maintenance or other adjustments to any system or component which contains, stores, transports or collects radioactive materials.c.Other activities which in the judgement of Radiation Protection Supervision warrant the issuance of a radiation wo rk permit prior to initiating the operations.

The radiation work permit st ipulates the purpose of entry , work location, radiation conditions, surveillance and dosimetry requirements, stay-t ime, protective clothing and equipment, and other procedural requirement s and precautions.

The primary objectives of the radiation wo rk permit program are to insure that:a.Radiological conditions are k nown as accurately as possible.b.Proper protective measures are taken to safely perform the r equired duties.c.Each person involved in these operations acknowledges his understanding of the radiation condition s, the protective and sa fety measures required, and his willingness to follow the radiation work permit requirements.d.Appropriate supervisors are aware of the task being performed, the radiation conditions, and the prescribed pr otective measures.

CALLAWAY - SA12.5-12Rev. OL-21 5/15e.A means for maintaining the account ability of personnel is provided.12.5.3.3Contamination ControlAreas that may be contaminat ed with radioactive material will be decontaminated to a level that is reasonably achi evable using available met hods and techniques. Since the complete removal of surface contamination from parts of the plant is a practical impossibility, certain plant areas may be designated as "contamination areas". These areas will be posted with the proper warni ng placards and barricaded.

Entry to these areas will be contro lled by radiation prot ection personnel and allowed only through the issuance of an appr opriate radiation work permit. Personnel, equipm ent, and material exiting from contaminated areas will be monitored to prevent the spread of contamination to clean areas. All contam inated equipment will be properly packa ged and identified before removal from a contami nated area. At access contro l facilities near the radiation protection office, a final survey will ensur e that all per sonnel, material, and equipment are free of significant contam ination, and thus will provide assurance that no radioactive material will spr ead to the uncontrolled areas of the plant.12.5.3.4Radiation Protection TrainingThe Manager, Technical Training is responsi ble for the radiation protection training program at the Callaway Plant. Plant personnel, both permanently assigned and temporary, receive training in the principles of radiation protecti on commensurate with the individual's job function and the anticipat ed radiation hazards.

The radiation protecti on training program will include as a minimum the following topics:a.Fundamentals of radiation and radioactivity.b.Biological effects of radiation on humans.c.Measurement of radiation and radioactivity.d.Principles and techniques of radiation protection.e.Use of protective clothing and equipment.f.General regulatory and s pecific facility license radiation protection requirements.g.Emergency planning.

h.ALARA program - concepts and methods.

The radiation protecti on training program will maintain the proficiency of these employees through periodic retr aining lectures and exercises.

CALLAWAY - SA12.5-13Rev. OL-21 5/1512.5.3.5Personnel Dosimetry12.5.3.5.1External Radiation Dosimetry See Section 12.5.2.1.8

.12.5.3.5.2Internal Ra diation Dosimetry See Section 12.5.2.1.9

.12.5.3.6Airborne Radiati on Evaluation and Control Airborne radioactivity will be routinely assessed using local sampling, portable continuous air monitors , and the fixed radiation monitoring system. Airborne radioactive materials (particulates, noble gases, halogens, tritium) will be sampled and analyzed using appropriate techniques. Since local sampling will provide better estimates of airborne contamination levels existing in a work area than wi ll a monitor reading, such special air sampling will be us ed in the radiation work per mit program to keep radiation exposure due to airborne radioactivity ALARA. Portable continuous air monitors and fixed airborne radioactivity monitors will be used to provide alarm indications and additional information which will be used wit h local sampling for the assessment of airborne radioactivity.

Control of airborne radioactivity levels wi ll be assured through the use of the plant's heating, ventilation, and air conditioning (HVAC) systems, portable air movers and filters. The HVAC systems provide contro lled air movement and filtra tion for those areas with a high potential for airborne radioactivity problems. Specia l control techniques can be used, such as plastic enclosu res which isolate and vent ai rborne radioactivity arising from special work projects. Respiratory pr otection equipment will be available for use in those situations where airbor ne radioactivity hazards exis t and other cont rol measures are inadequate at t he location and time.12.5.3.7Respiratory Protection ProgramThe respiratory protection program will be developed through the guidance of Regulatory Guide 8.15 and will satisfy the requirements of 10CFR20. As a minimum, the respiratory protection program will provide the following:a.Procedures to implement the selection, use, maintenance and storage of respiratory protection devices.b.Adequate facilities to support the storage, issue, cleaning and maintenance of respiratory protection devices.c.Use of only NIOSH certified or NRC approved equipment.

CALLAWAY - SA12.5-14Rev. OL-21 5/15d.Medical certification prog ram for respirator users.e.Periodic review of the overall effectiveness of the respiratory protection program via the inter nal dosimetry program.12.5.3.8Radioactive Materials Handling Methods and procedures will be developed to control, handle, and store by-product, source, and special nuclear material in accordance with regulatory requirements.Subjects to be covered by these procedures will incl ude the following:a.Storage of radioactive material in appropriately shielded, labeled and secured containers.b.Minimizing the distance that radi oactive samples are transported by personnel and the use of special ex tension and remote handling tools when applicable.c.Use of shielded sample transporters as appropriated.Periodic testing to verify the integrity of the sealed material.e.Emergency procedures which detail the proper actions to be taken in the event of leakage and spills.f.Accountability of any by-produ ct or special nuclear material.

CALLAWAY - SA Rev. OL-18 12/10TABLE 12.5-1 RADIATION PROTEC TION LABORATORY EQUIPMENTINSTRUMENTRADIATIONDETECTEDDETECTOR TYPICAL QUANTITYLOCATIONREMARKSGamma Spectroscopy SystemGammaHPGe w/graded shield4Counting Room and/or Maintenance Training Annex/

Operations Support FacilityProvides Gamma Isotopic analysis capabilities Gas Proportional CounterAlpha, Beta, Gamma Gas Flow Proportional1Counting RoomUsed for counting smears, effluent and radio chemistry

samples.Liquid Scintillation CounterBeta---1Counting LabTritium determinationIn Vivo CounterGammaNaI (TI) and/or HPGe1In Vivo Count Room Central Processing FacilityPersonnel In Vivo Counting; MultidetectorsNote:Sensitivities of this analytical instrumentation are dependent upon counting parameters such as sample geometry, count time and background, however, equipm ent from various manufacturers will be evaluated to ensure that the equipment when purchased will be sufficiently sensitive to perform the tasks for which it was intended. Instruments will be selected to satisfy m easurement and reporting requirements in effect at the ti me the equipment is ordered.

CALLAWAY - SA Rev. OL-15 5/06TABLE 12.5-2 PORTABLE RADIATION PROTECTION EQUIPMENT TYPE TYPICAL DETECTOR TYPE 1APPROXIMATE RANGE ACCURACY TYPICAL QUANTITYLOCATIONPortable Count Rate Meter/

Frisker Geiger-Mueller/ Scintillation 0-500,000 CPM 10% FS10See Note 3Low-Range Survey MeterGeiger-Mueller/

IIonization Chamber 0-2 R/hr/

0-5 R/hr 10% FS25 High-Range Extendable Probe Survey MeterGeiger-Mueller0-1000 R/hr 10% FS 4Mid-Range Survey MeterIGeiger-Mueller/

IIonization Chamber 0-50 R/hr 0-100 R/hr 10% FS5High-Range Survey MeterIGeiger-Mueller/

IIonization Chamber 0-1000 R/hr IIndication 20% FS3Neutron Survey MeterBF Tube Inside Polyethylene Sphere 0.2 mrem/hr 10rem/hr 10% FS4 Ultra-High Range Survey Meter Geiger-Mueller/

IIonization Chamber 0-10,000 R/hr 10% FS2Alpha Count Rate Meter/

FriskerScintillation0-500,000 CPM 10% FS5Portal Monitor/Personnel Contamination Monitor Gamma Scintillation/

Gas Flow ProportionalVariable-----6Beta-Gamma Count Ratemeter/FriskerGeiger-Mueller0-500,000 CPM 10% FS20Electronic DosimeterSolid State0-100 R/hr/ 0-1000R 10% FS250 CALLAWAY - SATABLE 12.5-2 (Sheet 2)

Rev. OL-15 5/06 Electronic Dosimeter ReaderNot ApplicableNot ApplicableNot Applicable 8See Note 3Calibration Transfer InstrumentIIonization ChamberVarious 5%1Portable Area MonitorIGeiger-Mueller/

Ionization Chamber/Solid State 0-2 R/hr 0-100 R/hr 25% Max.5Gamma Survey MeterNaI0-500k CPM 10%2High Volume Air SamplerNot Applicable> 15 CFM 20% 5Low Volume Air SamplerNot Applicable0-3 CFM 20%15Portable Continuous Air Monitor (3 channel type)NaI (I-131); Beta Scintillation (Particulate and

Gaseous)1E-11 to 1E-5 (I-131 and particulate) 1E-11

to 1E-1 (gaseous) 20% of Indication 4Note 1:The detector types listed are typical of the instrumentation used. As instrumentati on technology improves, an equivalent or better type of instrumentation may be substituted.Note 2:The quantities listed for each type of instrumentation ar e the typical numbers calibrated and ready for use.Note 3:Instrumentation is located in the RP Facilities on 1984' elevation of the Control Building, the Calibration Facilities, Dosimetry Processing Facilities or at specific work lo cations in the pl ant while in use.Note 4:The above information is referenc ed in the Radiological Em ergency Response Plan, Sect ion 6.7.3, and complies with Regulatory Guide 1.97.

TYPE TYPICAL DETECTOR TYPE 1APPROXIMATE RANGE ACCURACY TYPICAL QUANTITYLOCATION

Retrieved from "https://kanterella.com/w/index.php?title=ML17061A157&oldid=837245"