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{{#Wiki_filter:WOLF CREEK
{{#Wiki_filter:}}
 
CHAPTER 12.0
 
RADIATION PROTECTION
 
Section                                                    Page
 
12.1                                                  ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA)                                                  12.1-1
 
12.1.1 POLICY CONSIDERATIONS                                                            12.1-1
 
12.1.1.1          ALARA and Planning Committees                                                                                                                                                                                              12.1-1
 
12.1.1.2 Health Physics                                                                                                                                                                      12.1-1
 
12.1.2 DESIGN CONSIDERATIONS                                                            12.1-2
 
12.1.2.1 Plant Design                                                                                                                                                                                12.1-3 12.1.2.2 Scale-Model Program                                                                                                                                            12.1-3 12.1.2.3          Second-Level Design Reviews by the Lead Architect-Engineer 12.1-4 12.1.2.4          Design Reviews by SNUPPS                                                                                                                                                                                                                                                12.1-5 12.1.2.5          Examples of Radiation Protection Design Reviews 12.1-6 12.1.2.6 Decommissioning                                                                                                                                                                12.1-8
 
12.1.3 OPERATIONAL CONSIDERATIONS                                                12.1-8 12.1.4                              QUALITY ASSURANCE OF MAINTENANCE OF ALARA                                                                      12.1-9 12.
 
==1.5 REFERENCES==
12.1-10
 
12.2                                                  RADIATION SOURCES                                                                                                                                                                                                                                                                                                                      12.2-1
 
12.2.1 CONTAINED SOURCES                                                                      12.2-1
 
12.2.1.1 Containment                                                                                                                                                                                    12.2-1 12.2.1.2          Auxiliary Building                                                                                                                                                                                                                                                                                                            12.2-3 12.2.1.3          Fuel Building                                                                                                                                                                                                                                                                                                                                                              12.2-4 12.2.1.4 Turbine Building                                                                                                                                                            12.2-5 12.2.1.5 Radwaste Building                                                                                                                                                      12.2-5 12.2.1.6          Sources Resulting from Design Basis Accidents 12.2-6 12.2.1.7 Stored Radioactivity                                                                                                                                        12.2-6
 
12.2.2                              AIRBORNE RADIOACTIVE MATERIAL SOURCES                                                                                                              12.2-6
 
12.2.2.1          Model for Calculating Airborne Concentrations                              12.2-8
 
12.
 
==2.3 REFERENCES==
12.2-9
 
12.0-i                      Rev. 29 WOLF CREEK
 
TABLE OF CONTENTS (Continued)
 
Section                                                  Page
 
12.3                                                  RADIATION PROTECTION DESIGN FEATURES                                                                                                              12.3-1
 
12.3.1                              FACILITY DESIGN FEATURES                                                                                                                                                                                                                                      12.3-1
 
12.3.1.1          Plant Design Description for as Low as is Reasonably Achievable (ALARA)                                                                                                                                                                                    12 3-1 12.3.1.2          Radiation Zoning and Access Control                                                                                                                        12 3-8
 
12.3.2 SHIELDING                                                                                        12.3-9
 
12.3.2.1 Design Objectives                                                                                                                                                12.3-9 12.3.2.2          General Shielding Design                                                                                                                                                                                                                                      12.3-10 12.3.2.3          Shielding Calculational Methods                                                                                                                                                                12.3-14
 
12.3.3 VENTILATION                                                                                  12.3-15
 
12.3.3.1          Design Objectives                                                                                                                                                                                                                                                                                                            12 3-15 12.3.3.2          Design Criteria                                                                                                                                                                                                                                                                                                                                12 3-15 12.3.3.3 Design Guidelines                                                                                                                                                12.3-16 12.3.3.4          Design Description                                                                                                                                                                                                                                                                                                  12 3-18 12.3.3.5          Air Cleaning System Design                                                                                                                                                                                                                  12.3-19
 
12.3.4                              AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION                                                                                                                                                                                                                  12.3-20
 
12.3.4.1          Area Radiation Monitoring                                                                                                                                                                                                                            12.3-20 12.3.4.2          Airborne Radioactivity Monitoring                                                                                                                                            12.3-25
 
12.
 
==3.5 REFERENCES==
12.3-40
 
12.4 DOSE ASSESSMENT                                              12.4-1
 
12.4.1                              EXPOSURES WITHIN THE PLANT                                                                                                                                                                                                                  12.4-1
 
12.4.1.1          Direct Radiation Dose Estimates                                                                                                                                                                12.4-1 12.4.1.2          Airborne Radioactivity Dose Estimates                                                                                                    12.4-4 12.4.1.3          Illustrative Examples of Dose Assessment                                                                      12.4-5
 
12.4.2                              EXPOSURES AT LOCATIONS OUTSIDE PLANT                                                                                                              12.4-8 STRUCTURES
 
12.4.2.1          Direct Radiation Dose Estimates                                                                                                                                                                12.4-8 12.4.2.2          Exposures Due to Airborne Radioactivity                                                                                12.4-8
 
12.
 
==4.3 REFERENCES==
12 4-8
 
12.0-ii                    Rev. 29 WOLF CREEK
 
TABLE OF CONTENTS (Continued)
 
Section                                                  Page
 
12.5                                                  HEALTH PHYSICS PROGRAM                                                                                                                                                                                                                                                          12.5-1
 
12.5.1 ORGANIZATION                                                                                12.5-1 12.5.2                              EQUIPMENT, INSTRUMENTATION AND FACILITIES                                                            12.5-3
 
12.5.2.1          Health Physics Equipment and Instrumentation                              12.5-3 12.5.2.2          Health Physics Facilities                                                                                                                                                                                                                            12.5-5
 
12.5.3 PROCEDURES                                                                                      12.5-7
 
12.0-iii                    Rev. 29 WOLF CREEK
 
TABLE OF CONTENTS (Continued)
 
LIST OF TABLES
 
Number                        Title
 
12.1-1    Regulatory Criteria Applicable to the Operating Agent's Health Physics Program
 
12.2-1    Neutron Fluxes on Inside Surface of the Primary Shield Wall at the Core Centerline (100% Power)
 
12.2-2    Gamma Fluxes on Inside Surface of the Primary Shield Wall at the Core Centerline (100% Power)
 
12.2-3    Pressurizer Shielding Source Terms
 
12.2-4    Spent Fuel Shutdown Sources (Full Core)
 
12.2-5    Radiation Sources Residual Heat Removal System
 
12.2-6    Chemical and Volume Control System Sources Letdown Mixed Bed Demineralizer
 
12.2-7                  Fuel Storage Pool Water Activities
 
12.2-8    Secondary System Activities
 
12.2-9    Conservative Basis Accumulated Radioactivity in the Gaseous Waste Processing System After Forty Years Operation
 
12.2-10  Fort Calhoun Operating Data
 
12.2-11  Parameters and Assumptions for Calculating Airborne Radioactive Concentrations
 
12.2-12  Airborne Radioactivity Concentrations
 
12.3-1    List of Computer Codes Used in Shielding Design Calculations
 
12.3-2    Area Radiation Monitors
 
12.3-3    Inplant Airborne Radioactivity Monitors
 
12.3-4    Power Supplies for Area and In-Plant Airborne Monitors
 
12.4-1    Illustrative Examples of Dose Assessment
 
12.0-iv                    Rev. 14 WOLF CREEK
 
TABLE OF CONTENTS (Continued)
 
LIST OF TABLES
 
Number                        Title
 
12.4-2    Average Number of Personnel per PWR Unit for the Period 1969-1977
 
12.4-3    Average Occupational Radiation Exposure (Man-Rem Dose) per PWR Unit for the Period 1969-1977
 
12.4-4    Average Occupational Radiation Exposure (Man-Rem Dose)
Based on PWR Plant Age
 
12.4-5    Distribution of the Number of Personnel (>100 Millirem/Yr) According to Work Function
 
12.4-6    Distribution of Personnel (>100 Millirem/Yr) According to Employee Category
 
12.4-7    Percentages of Personnel Dose by Work Function
 
12.4-8    Annual Occupational Exposures for Various PWR Vendor's Units
 
12.4-9    Average Annual Occupational Exposure and Power for Indi-vidual PWRs
 
12.4-10  Average Individual Exposure Based on PWR Plant Age
 
12.4-11  Cumulative Average of Annual Exposure by Years of Oper-ation - PWRs
 
12.4-12  Estimates of Occupancy Times in Plant Radiation Areas and Gamma Doses to Plant Personnel
 
12.4-13  Distribution of Direct Radiation Man-Rem Doses According to Work Functions
 
12.4-14  Annual Occupancy in Plant Areas Containing Airborne Radioactivity
 
12.4-15  Doses to Plant Personnel Caused by Airborne Radio-activity
 
12.5-1    Health Physics and Lab Equipment
 
12.5-2    Portable Health Physics Equipment
 
12.0-v                      Rev.25
 
WOLF CREEK
 
CHAPTER 12 - LIST OF FIGURES
*Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
 
Figure #                                              Sheet                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            Title                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            Drawing #*
12.3-1                                                                                          1                                                                                                                                Typical Valve Compartment Arrangement 12.3-1                                                                                          2                                                                                                                                Typical Valve Compartment Arrangement 12.3-1                                                                                          3                                                                                                                                Typical Valve Compartment Arrangement 12.3-1                                                                                          4                                                                                                                                Typical Valve Compartment Arrangement 12.3-1                                                                                          5                                                                                                                                Typical Valve Compartment Arrangement 12.3-2                                                                                          1                                                                                                                                Radiation Zones for Normal Operation El. 1974                                                                                                                                                              10466-A-1701 12.3-2                                                                                          2                                                                                                                                Radiation Zones for Normal Operation El. 2000                                                                                                                                                              10466-A-1702 12.3-2                                                                                          3                                                                                                                                Radiation Zones for Normal Operation El. 2026                                                                                                                                                              10466-A-1703 12.3-2                                                                                          4                                                                                                                                Radiation Zones for Normal Operation El. 2047-6                                                                                                              10466-A-1704 12.3-2                                                                                          5                                                                                                                                Radiation Zones for Normal Operation Turbine Bldg El.                                                  10466-A-1705 1983 & 2000 12.3-2                                                                                          6                                                                                                                                Radiation Zones for Normal Operation Turbine Bldg El.                                                  10466-A-1706 2033 & 2065 12.3-3                                                                                          0                                                                                                                                Control Room Isometric 12.3-4 0    Decontamination System                                                  M-12HD01 12.4-1                                                                                          1                                                                                                                                Cumulative Average of Annual Exposure by Years of Operation - PWRs 12.4-1                                                                                          2                                                                                                                                Cumulative Average of Annual Exposure by Years of Operation - PWRs 12.4-1                                                                                          3                                                                                                                                Cumulative Average of Annual Exposure by Years of Operation - PWRs 12.4-1                                                                                          4                                                                                                                                Cumulative Average of Annual Exposure by Years of Operation - PWRs 12.5-1 0                                                                                  Health Physics Area in the Walter P. Chrysler Support Complex 12.5-2                                                                                          0                                                                                                                                Health Physics Area in the Control Building 12.5-3                                                                                          0                                                                                                                                Sample Lab Facilities in the Radwaste Building 12.5-4                                            0                                      Dosimetry office in Olive Ann Beech Building
 
12.0-vi    Rev.25 WOLF            CREEK
 
NRC        QUESTIONS        PERTAINING        TO        CHAPTER        12.0
 
Section                                                                                                                      Question        Number            Section                                                  Question        Number
 
12.1.2.5b                                                                                                                                          331.1 12.2.1.3                                                                                                                                                    331.2 12.2.1.2.3                                                                                                                                331.3 T12.2-7                                                                                                                                                              331.4 12.3.4.2.2.2.2                                                                                        331.5
 
12.0-vii                                                                                                                                                                                                      Rev.        0 WOLF CREEK
 
CHAPTER 12.0
 
RADIATION PROTECTION
 
12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA)
 
12.1.1  POLICY CONSIDERATIONS
 
The Operating Agent is committed to a company policy that supports the applicable Regulatory Guides (see Table 12.1-1) and parts of Title 10 of the Code of Federal Regulations in maintaining occupational radiation exposure (ORE) as low as reasonably achievable (ALARA). A Radiation Protection Operative Policy, initiated by the President and Chief Executive Officer through the Operating Agent Corporate Policy Manual has been implemented to meet this commitment. The Manager Radiation Protection is responsible for the technical content of the program and ensuring professional ALARA input. The Operating Agent establishes radiation protection practices for applicable plant activities and provides a qualified health physics organization to accomplish this goal. Utility management recognizes and emphasizes the importance of each individual's responsibilities to maintain occupational radiation exposures (ORE) - ALARA.
 
12.1.1.1  ALARA and Planning Committees
 
Under the ALARA Program, the ALARA Committee is responsible for integrating regulatory requirements with management policy by providing a multidisciplined forum for the discussion of radiological problems. A member of management serves as chairman of the ALARA Committee.
 
The station ALARA Coordinator participates in outage planning and scheduling of maintenance and testing activities during operations to evaluate dose reduction mechanisms and provide the necessary directives to ensure OREs, through the support of a functional onsite Health Physics Program, are maintained ALARA.
 
12.1.1.2  Health Physics
 
The ALARA Program maintains, through the Radiation Protection Manual and the Operative Policy contained in the Corporate Policy Manual:  1) that personnel work closely together to ensure that each Nuclear Division's ALARA responsibilities are carried out and that the site is updated to current regulatory standards;  2) an adequate training program to educate all nuclear station personnel in the areas of ALARA practices; 3) a method of reviewing the effectiveness of the station ALARA Program; 4) the formation of an ALARA committee to resolve areas of activity which offer the potential for increasing radiation exposures.
 
12.1-1                        Rev. 25 WOLF CREEK
 
The Health Physics ALARA group is responsible for analyzing the latest regulatory criteria and providing general Health Physics support. This section also performs ALARA reviews of non-expedited plant modification requests.
Expedited plant modification requests (Category I) ALARA reviews will be performed by Engineering.
 
The plant manager is responsible for ensuring that a station ALARA program is developed and that all station personnel support it. The plant manager is responsible for all operational aspects of the station, including the WCGS Health Physics Program, and support of the station Radiation Protection Manager.
 
The Radiation Protection Manager is responsible for developing and implementing the WCGS Health Physics Program and maintaining plant occupational radiation exposures ALARA. This manager is the site expert on radiation protection and is responsible for supporting operations and maintenance through radiological safety procedures, manpower and the provision of radiological protection equipment in controlled areas. This manager participates in site planning activities such as, but not limited to, the emergency plan, outages, equipment procurement, and personnel training, which require radiological input.
 
To ensure freedom from operation and maintenance pressures, this manager retains the administrative freedom to report technical information of an immediate nature directly to the plant manager. Any ALARA concerns can be appealed to the ALARA Committee and reported to corporate management.
 
The WCGS Health Physics organizations works to consider the radiological implications of WCGS from long range perspectives. The organization ensures the application of sound Health Physics, radiation protection, and ALARA philosophies.
 
12.1.2  DESIGN CONSIDERATIONS
 
A major objective in the design of the WCGS powerblock has been to limit the potential radiation exposures of operating personnel. This objective has been emphasized to the lead architect engineer (Bechtel) from the outset of the project, through reviews of project design criteria and frequent design reviews as the detailed design of the powerblock has evolved. Specific design considerations and the guidelines employed in developing the WCGS design are presented in Sections 12.3.1.1 and 12.3.3.3. Estimated occupational radiation doses are given in Section 12.4. Significant elements of the design program to ensure that ORE are ALARA have been implemented as described in the following sections.
 
For ALARA design considerations associated with the ISFSI including fuel loading, transfer and storage operations, refer to the NUHOMS EOS System UFSAR, Docket 72-1042 and the applicable site-specific dose calculations as documented in the 72.212 Evaluation Report.
 
12.1-2                        Rev. 35 WOLF CREEK
 
12.1.2.1  Plant Design
 
The design engineers and first level supervisors assigned by Bechtel to WCGS 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 WCGS. Bechtel design engineers are also made aware of other operating experience through Licensee Event Reports, NRC IE Bulletins and Information Notices, 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 Reference
: 1. Other recommendations 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 Reference 1.
 
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. 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 secondary waste evaporator and 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."
 
12.1.2.2  Scale-Model Program
 
The use of scale models is an innovative aspect of the WCGS 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.
 
12.1-3                        Rev. 8 WOLF CREEK
 
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 the Operating Agent 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.
 
The 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.
 
12.1.2.3  Second-Level Design Reviews by the Lead Architect-Engineer
 
The 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 5 to 20 years of nuclear power plant design experience.
 
12.1-4                        Rev. 8 WOLF CREEK
 
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.
 
12.1.2.4  Design Reviews by SNUPPS
 
Operating Agent reviews of safety-related systems, structures, and equipment were coordinated through the SNUPPS Technical Committee, which was composed of senior-level utility engineers, one from each SNUPPS utility (Wolf Creek and Callaway). It was 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 duration of the SNUPPS project, the Technical Committee 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 averaged about 50. The SNUPPS technical director participated in Technical Committee meetings.
 
Assisting the Technical Committee were various plant review groups, which are ad hoc groups of Operating Agent personnel selected to review and recommend action on specific aspects of design. The total meeting days per year averaged about 15. A SNUPPS staff member participated in each meeting.
 
Radiation protection was an important aspect of the reviews by the Technical Committee and the plant review groups and was the subject of numerous specific reviews, as discussed further in Section 12.1.2.5. At least six distinct reviews of the scale models specifically included consideration of radiation protection during plant operation and maintenance. At these reviews, representatives of SNUPPS included, in addition to Technical Committee members and SNUPPS staff, the health physics superintendent from Ginna Station, several licensed senior reactor operators, operations and maintenance supervisors from other nuclear power plants, and Operating Agent engineering personnel experienced in nuclear plant operation. Several design changes resulted from these reviews. Examples are described in Section 12.1.2.5.
 
During design and construction of the first SNUPPS unit (Callaway), SNUPPS staff and qualified personnel from WCGS 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 WCGS participated in ALARA reviews of Wolf Creek.
 
12.1-5                        Rev. 8 WOLF CREEK
 
12.1.2.5  Examples of Radiation Protection Design Reviews
: a. Radiation Zone Drawings
 
Every 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 WCGS 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 powerblock. These drawings were reviewed by the Operating Agent and SNUPPS staff.
Participants in those reviews included the Technical Committee, Operating Agent engineering personnel, Operating Agent operating and health physics personnel, and SNUPPS staff.
: b. Reactor Cavity
 
WCGS 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 WCGS) for access to the outside of the reactor vessel for performance of inservice inspection. WCGS through 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 had numerous design review meetings with Bechtel. Participants for SNUPPS have included:  SNUPPS staff (executive director, technical director, and licensing manager), the Technical Committee, and Operating Agent 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 Section 3.8.3.1.4 and 12.3.2.2.1.
: c. Radioactive Demineralizers
 
As a result of a review of a study model of the auxiliary building, the following design changes were made:
 
12.1-6                        Rev. 8 WOLF CREEK
: 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 Containment
 
Airborne 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 WCGS plant:
: 1. A 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).
: 4. Provision of a containment atmospheric control system (Section 9.4.6) to remove airborne iodine from the containment. This reduces the potential airborne exposure to operators entering the containment and also results in reduced environmental concentrations of iodine when purging the containment.
: e. Steam Generator Maintenance
 
A 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.
 
12.1-7 Rev. 13 WOLF CREEK
 
12.1.2.6  Decommissioning
 
The following features of the plant design will assist decommissioning crews to maintain ORE-ALARA during the eventual decommissioning of WCGS.
: 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, provisions for remote flushing of equipment, ability to use remote handling equipment, and component design features to minimize crud buildup.
: c. Specifications and limitations on cobalt content in equipment components serve to limit radiation doses from crud buildup during both operation and subsequent decommissioning.
: d. The amount of potentially radioactive buried pipe is limited.
 
12.1.3  OPERATIONAL CONSIDERATIONS
 
The Health Physics staff works with other WCGS groups to coordinate their input and ensure that proper radiological surveillance and controls for maintenance, operations, waste handling, inservice inspection, decommissioning and other activities are performed in a manner that maintains occupational exposures ALARA. This includes work on such systems as the NSSS, the Residual Heat Removal System, the Fuel Handling System, the Liquid Waste Management Systems, the Gaseous Waste Management Systems, the Solid Waste Management Systems, the Fuel Pool Cooling and Cleanup System, the ISFSI and other systems that collect, store, contain or transport liquid, gaseous and solid radioactive material.
 
Routine operational practices used at WCGS to promote an ALARA philosophy and objectives are; the employment of a radiological staff that meets Regulatory Guide 1.8 requirements, the proper training of personnel and the preparation and preplanning of potentially high radiation exposure jobs. Briefings, surveys, critiques, etc, are administrative tools, which are used to ensure that station doses are maintained ALARA. The development and implementation of radiation protection procedures, practices and techniques in conjunction with adequate supervisory and radiation protection surveillance, provides a system that ensures ALARA is adequately reflected in station activities.
 
12.1-8 Rev. 35 WOLF CREEK
 
The operation and calibration procedures developed and implemented by the WCGS Health Physics staff provide specific guidelines, techniques and methods that incorporate guidance from Regulatory Guides 8.8 and 8.10. Past operational experience of the Health Physics personnel and information from other Nuclear Power Plants were used in program development and implementation to minimize occupational exposure problems. A description of the Health Physics program is given in Section 12.5.3.
 
The Manager Radiation Protection through a Supervisor Radiation Protection has the primary responsibility for developing and implementing the WCGS Health Physics Program. Implementation of the program is through the use of administrative exposure controls and procedures, Health Physics procedures, employee training, Health Physics review of plant procedures, job preparation, pre-planning of work which may produce significant exposures, and a Radiation Work Permit (RWP) program by which Health Physics regulates access to controlled areas. A description of the RWP System is given in Section 12.5.3.
 
To ensure that Health Physics is an integral part of the WCGS plant operation, the Manager Radiation Protection is a member of the Plant Safety Review Committee.
 
The Manager Radiation Protection through a Supervisor Radiation Protection periodically reviews exposure records for the purpose of identifying exposure by category, location, job function, etc., to enable the recommendation of changes to plant procedures, operating methods, etc., where such changes may reduce exposure.
 
12.1.4  Quality Assurance of Maintenance of ALARA
 
To verify that radiation protection functions are being performed as required and that a high level of radiological safety is maintained, review and evaluation through audit of the radiation protection program is performed biennially by the Operating Agent Quality Assurance Division. The performance of audits and review of radioactivity monitoring (fixed and portable),
radioactivity sampling, contamination measurement and analysis, internal and external personnel monitoring, instrument storage, calibration and maintenance, decontamination, and the respiratory protection program including testing and contamination control to assure program effectiveness is consistent with the position of NUREG 0761.
 
The audit program is conducted in accordance with established auditing principles:
: 1. Use of inspectors/auditors with training and expertise in the area being audited.
: 2. Use of audit personnel with no responsibilities or "vested interests" in the areas being reviewed.
 
12.1-9                        Rev. 25 WOLF CREEK
: 3. Documentation of audit results and findings, and review by management.
: 4. Performance of corrective or followup actions as appropriate.
 
Additional quality for these items is built into HP procedures utilizing quality provisions from regulatory guides.
 
Adequacy of permanent and temporary biological shielding within the Radiation Controlled Areas (RCAs) is verified by periodic radiation surveys.
 
The programs which control the monitoring activities are administered to meet the requirements of 10 CFR 20.
 
12.1.5  REFERENCE
: 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.
 
12.1-10    Rev. 8 WOLF CREEK TABLE 12.1-1
 
Regulatory Criteria Applicable to The Operating Agent's Health Physics Program
 
Number                Title                        Revision    Issuance
 
1.8          Personnel Selection and Training    Draft Rev. 2      2/79
 
8.2          Guide For Administrative Practices        -        2/2/73 in Radiation Monitoring
 
8.4          Direct-Reading and Indirect Reading      -        2/26/73 Pocket Dosimeters
 
8.7          Occupational Radiation Exposure Records  4        5/2018 System
 
8.8          Information Relevant to Ensuring that    3          6/78 Occupational Radiation Exposures at Nuclear Power Stations will be As Low As is Reasonably Achievable (ALARA)
 
8.9          Acceptable Concepts, Models, Equations,  -          9/73 and Assumptions for a Bioassay Program
 
8.10        Operating Philosophy for Maintaining      1-R        9/75 Occupational Radiation Exposures As Low As is Reasonably Achievable
 
8.13        Instruction Concerning Prenatal          1          11/75 Radiation Exposure
 
8.14        Personnel Neutron Dosimeters              1          8/77
 
8.15        Acceptable Programs for Respiratory      -          10/76 Protection
 
8.26        Applications of Bioassay for Fission      -          9/80 and Activation Products
 
ANSI        American National Standard                -          1972 N13.5-1972  performance specfications for direct reading and indirect reading pocket dosimeters for x - and gamma radiation
 
ANSI        American National Standard                R1972      1972 N13.6-1966  practice for occupational radiation exposure records systems
 
ANSI/ANS    American National Standard                            1978 3.1-1978    for selection and training of Nuclear Power Plant Personnel
 
On June 24, 1997 an exemption from the requirements of 10CFR70.24 was granted by the NRC, therefore Regulatory Guide 8.12 is no longer applicable. On November 12, 1998 the NRC issued 10CFR50.68, which provides eight criteria that may be followed in lieu of criticality monitoring per 10CFR70.24 and revised 10CFR70.24 to make any exemption ineffective so long as the licensee elects to comply to 10CFR50.68.
 
Rev. 37 WOLF CREEK
 
12.2  RADIATION SOURCES
 
The 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.
 
12.2.1  CONTAINED SOURCES
 
The 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 Section 1.2.
 
12.2.1.1  Containment
 
12.2.1.1.1  Reactor Core
 
The primary radiation within the containment during normal operation is 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.
 
12.2.1.1.2  Reactor Coolant System
 
Sources 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 Section 11.1.
 
12.2-1                        Rev. 0 WOLF CREEK
 
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 Table 12.2-3.
 
12.2.1.1.3  Secondary Coolant Cycle
 
Under 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 11.1 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 Section 11.1.
 
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.
 
12.2.1.1.4  Auxiliary Systems
 
Residual heat removal system - see Section 12.2.1.2.1.
 
Chemical and volume control system - see Section 12.2.1.2.2.
 
Boron Recycle system - see Section 12.2.1.5.
 
12.2.1.1.5  Spent Fuel Handling and Transfer
 
The 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 100 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 Table 12.2-4.
 
For radiation sources associated with the ISFSI including fuel loading, transfer and storage operations, refer to the NUHOMS EOS System UFSAR, Docket 72-1042 and the applicable site-specific dose calculations as documented in the 72.212 Evaluation Report.
 
12.2-2                        Rev. 35 WOLF CREEK
 
12.2.1.2  Auxiliary Building
 
12.2.1.2.1  Residual Heat Removal System
 
The 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 Table 12.2-5.
 
12.2.1.2.2  Chemical and Volume Control System
 
The CVCS source activity is the reactor coolant inventory which is provided in Table 11.1-4. More than 1 minute 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 heat exchanger, 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.
 
12.2.1.2.3  Nuclear Sampling System
 
The sampling systems for (a) reactor coolant, (b) containment sump water, and (c) containment atmosphere have been evaluated to meet normal sample conditions and the guidelines in WCAP-14986-A, Rev. 2, Post Accident Sampling system Requirements: A Technical Basis, to assure that a recovery sample can be taken if needed. Steps will be taken to assure that exposure is minimized while taking recovery samples.
 
By Amendment 137, the in-line post-accident sampling system is no longer used, and the equipment has been abandoned in place. The sample panel for this system is located in Room 1312 of the auxiliary building.
 
12.2-3                        Rev. 29 WOLF CREEK
 
The existing sampling systems, which provide the capability to make required analyses under normal conditions, are retained. Tables 9.3-3, 9.3-4, 9.3-5, and 9.3-6 list the various systems which are sampled. 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 Section 12.3.4 and are shown in Figure 12.3-2.
 
The 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.
 
12.2.1.3  Fuel Building
 
12.2.1.3.1  Spent 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 Section 12.2.1.1.5. For shielding design, the fuel storage pool assumptions are given in Section 9.1.2. The major radionuclide concentrations in the water are provided in Table 12.2-7.
 
12.2.1.3.2  Fuel Pool Cooling and Cleanup System
 
Sources 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.
 
12.2-4                        Rev. 20 WOLF CREEK
 
12.2.1.4  Turbine Building
 
12.2.1.4.1  Main Steam Supply and Power Conversion Systems
 
Potential 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 Section 12.2.1.1.3. This radioactivity is sufficiently low that no radiation shielding for equipment in the turbine building, except the condensate demineralizers, is required in order to meet the shielding design requirements. The isotopic concentrations for a condensate demineralizer bed and other secondary system sources are listed in Table 12.2-8.
 
12.2.1.5  Radwaste Building
 
12.2.1.5.1  Boron Recycle and Liquid Radwaste Systems
 
The 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, demineralizers, and evaporators, 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 Sections 11.1 and 11.2.
 
12.2.1.5.2  Gaseous Radwaste System
 
Radiation sources for each component of the waste gas system are based on operation under the conditions given in Sections 11.1 and 11.3. Tabulation of the activities is shown in Table 12.2-9.
 
12.2.1.5.3  Solid Radwaste System
 
Radiation 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 Table 11.4-3.
 
12.2-5                        Rev. 5 WOLF CREEK
 
12.2.1.6  Sources Resulting from Design Basis Accidents
 
The 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 Chapters 11.0 and 15.0.
 
12.2.1.7  Stored Radioactivity
 
The principal sources of activity not stored inside the plant structures are the reactor makeup water storage tank (RMWST) and the refueling water storage tank (RWST). The RMWST is 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 10 mrem/hr when the water is returned from the refueling pool. This is rapidly reduced by processing through the fuel storage pool purification filter and demineralizer. Spent fuel is stored in the fuel storage pool until prepared for shipment offsite. Storage space is allocated in the radwaste building for the storage of spent filter cartridges and solidified spent resins, evaporator bottoms, and chemical wastes. Radioactive wastes stored inside designed storage areas are shielded so anticipated access outside the structure will meet Zone A specifications (see Figure 12.3-2). If it becomes necessary to store radioactive wastes outside the plant structures, adequate radiation protection measures are taken by the health physics staff. Tabulations of the activities within these tanks are provided in Table 11.1-6 (Sheets 1,2, and 3).
 
12.2.2  AIRBORNE RADIOACTIVE MATERIAL SOURCES
 
This section identifies the models, parameters, and sources required to evaluate potential long-term 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 DACs for various stations in the auxiliary building (these stations include the waste handling area and the waste gas decay tank rooms, which for WCGS are located in the radwaste building) indicates that the concentration to DAC ratios in all stations is well within the limits of 10CFR20.1204(G) to consider these airborne activities as insignificant. It is expected that WCGS will not have airborne radioactivity concentrations significantly
 
12.2-6                        Rev. 14 WOLF CREEK
 
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, are normally not occupied. Also the ventilation systems in the auxiliary and radwaste buildings are designed in such a manner that airborne contamination from high radiation zones does 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 (see Table 12.2-10). 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 10-13          P Ci/cc and 6.9 x 10-12          P Ci/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 fuel storage pool.
During power operation, the airborne radioactivity in the fuel building is almost all due to tritium, since the operation of the fuel pool cleanup system reduces concentrations of other isotopes in 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 Table 12.2-11. The concentrations in these buildings are listed in Table 12.2-12. 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 are a small fraction of the 10 CFR 20 limits for occupation exposures.
 
Airborne radioactivity is monitored inside the plant, as described in Section 12.3.4, and in process equipment and effluents, as described in Section 11.5.
 
12.2-7                        Rev. 14 WOLF CREEK
 
12.2.2.1  Model For Calculating Airborne Concentrations
 
For 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 peak or equilibrium airborne concentration of the radioisotope in the regions can be calculated, using the following equation:
 
Ci  (t) = (LR)i Ai  (PF)i  (1-e-OTit        )                    (1)
VOTi where
 
(LR)i      = Leak or evaporation rate of the ith radioisotope in gm/sec, in the applicable region and
 
Ai        = activity concentration of the ith leaking or evaporating radioisotope in P Ci/gm
 
(PF)i      = partition factor or the fraction of the leaking activity that is airborne for the ith radioisotope
 
OTi        = Total removal rate constant for the ith radioisotope in sec-1 from the applicable region
 
                = (Odi  + Oe )
 
Odi  +                      Oe are the removal rate constants in sec-1 due to radioactive decay for the ith radioisotope and the exhaust from the applicable region
 
t = time interval between the start of the leak and the time at which the concentration is evaluated in seconds
 
V  = free volume of the region in which the leak occurs in cc
 
Ci(t)      = airborne concentration of the ith radioisotope at time t in P Ci/cm3 in the applicable region
 
12.2-8                        Rev. 0 WOLF CREEK
 
From the above equation, it is evident that the peak or equilibrium concentration,  CEqi, of the ith radioisotope in the applicable region is given by the following expression:
 
CEqi  =  (LR)i  Ai  (PF)i  /V W Ti              (2)
 
With high exhaust rates, this peak concentration is reached within a few hours.
 
12.
 
==2.3  REFERENCES==
: 1. C. M. Lederer, et. al., Table of Isotopes, Lawrence Radiation Laboratory, University of California (March, 1968).
: 2. Reactor Physics Constants, Argonne National Laboratory, ANL-5800 (July, 1963).
: 3. H. Soodak, Reactor Handbook, 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 & G Idaho, Inc., July 1978.
: 7. NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous & Liquid Effluents from Pressurized Water Reactors, USNRC, April 1976.
 
12.2-9                        Rev. 1 WOLF CREEK
 
TABLE 12.2-1
 
NEUTRON FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)
 
Neutron Flux Energy Group                      (neutrons/cm2-sec)
 
1(E > 1.0 Mev)                        7.6 x 108
 
2(5.53 Kev < E < 1.0 Mev)              1.2 x 1010
 
3(0.625 ev < E < 5.53 Kev)            7.1 x 109
 
4(E < 0.625 ev)                        1.8 x lO9
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-2
 
GAMMA FLUXES ON INSIDE SURFACE OF THE PRIMARY SHIELD WALL AT THE CORE CENTERLINE (100% POWER)
 
Group              Flux(Mev/cm2-sec)        Group Energy (Mev/  )
 
1                    3.7 x 109                      7.5
 
2                    3.3 x 109                      4.0
 
3                    1.7 x 109                      2.5
 
4                    1.0 x 109                      0.8
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-3
 
PRESSURIZER SHIELDING SOURCE TERMS
 
Liquid Volume                    Liquid Volume
 
Isotope      Activity              Isotope    Activity
(&#xb5;Ci/gm)                          (&#xb5;Ci/gm)
 
N-16        1.8 (max)            Cs-135      1.04E-10 Cr-51        1.68E-03              Cs-136      2.10E-02 Mn-54        3.10E-04              Cs-137      3.79E-02 Fe-55        1.61E-03              Ba-137M    3.58E-02 Fe-59        9.30E-04              Ba-140      3.55E-04 Co-58        1.53E-02              La-140      3.43E-04 Co-60        2.02E-03              Ce-141      1.32E-04 Br-83        2.53E-04              Ce-143      2.19E-05 Br-84        3.08E-05              Ce-144      6.86E-05 Br-85        3.39E-07              Pr-143      8.67E-05 Rb-86        1.48E-04              Pr-144      6.86E-05 Rb-88        1.32E-03 Sr-89        6.84E-04                Steam Volume Sr-90        2.llE-05 Sr-91        1.29E-04              Isotope    Activity Y-89M        6.16E-08                          (&#xb5;Ci/cc)
Y-90        1.35E-05 Y-91M        8.44E-05              Kr-83M      9.74E-04 Y-91        1.33E-04              Kr-85M      1.14E-02 Y-93        7.05E-06              Kr-85      1.13E+00 Zr-95        5.72E-05              Kr-87      1.92E-03 Nb-95        4.84E-05              Kr-88      1.36E-02 Nb-95M      2.92E-05              Kr-89      6.89E-06 Mu-99        7.34E-02              Xe-131M    1.31E-01 Tc-99        3.68E-09              Xe-133M    1.35E-01 Ru-103      4.14E-05              Xe-133      1.59E+01 Ru-106      1.00E-05              Xe-135M    9.03E-05 Te-125M      2.75E-05              Xe-135      6.79E-02 Te-127M      2.73E-04              Xe-137      1.49E-05 Te-127      3.26E-04              Xe-138      2.69E-04 Te-129M      1.27E-03 Te-129      8.23E-04 Te-131M      6.13E-04 Te-131      1.15E-04 Te-132      1.24E-02 I-129        8.55E-13 I-130        5.17E-04 I-131        3.83E-01 I-132        1.71E-02 I-133        1.46E-01 I-134        9.09E-04 I-135        2.68E-02 Cs-134      5.24E-02
 
Rev. 0
 
WOLF CREEK
 
TABLE 12.2-5
 
RADIATION SOURCES RESIDUAL HEAT REMOVAL SYSTEM
 
Isotope        Activities          Isotope    Activities
(&#xb5;Ci/gm)                      (&#xb5;Ci/gm)
 
Cr-51          1.51E-03            I-131      4.37E-01 Mn-54          2.48E-04            I-132      6.36E-02 Fe-55          1.28E-03            I-133      5.46E-01 Fe-59          7.98E-04            I-134      3.17E-03 Co-58          1.28E-02            I-135      2.06E-01 Co-60          1.60E-03            Xe-131M    3.94E-02 Br-83          2.49E-03            Xe-133M    2.07E.01 Br-84          2.28E-05            Xe-133    1.06E+01 Br-85          negligible          Xe-135M    3.56E-02 Kr-83M        1.34E-02            Xe-135    5.42E-01 Kr-85M        1.20E-01            Xe-137    negligible Kr-85          1.68E-02            Xe-138    8.llE-07 Kr-87          1.48E-02            Cs-134    4.10E-02 Kr-88          1.57E-01            Cs-135    4.48E-13 Kr-89          negligible          Cs-136    2.llE-02 Rb-86          1.39E-04            Cs-137    2.95E-02 Rb-88          1.75E-01            Cs-138    4.47E-04 Rb-89          5.28E-08            Ba-137M    2.79E-02 Sr-89          5.73E-04            Ba-140    3.58E-04 Sr-90          1.64E-05            La-140    2.54E-04 Sr-91          8.01E-04            Ce-141    1.14E-04 Y-89M          5.15E-08            Ce-143    6.03E-05 Y-90          2.58E-06            Ce-144    5.41E-05 Y-91M          5.30E-04            Pr-143    8.18E-05 Y-91          1.06E-04            Pr-144    5.41E-05 Y-93          4.25E-05 Zr-95          4.79E-05 Nb-95          4.00E-05 Nb-95M        1.45E-06 Mo-99          1.32E-01 Tc-99          2.00E-10 Ru-103        3.59E-05 Ru-106        8.00E-06 Te-125M        2.32E-05 Te-127M        2.24E-04 Te-127        5.63E-04 Te-129M        1.12E-03 Te-129        7.66E-04 Te-131M        1.82E-03 Te-131        3.34E-04 Te-132        2.08E-02 I-129          2.63E-14 I-130          2.75E-03
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-6
 
CHEMICAL AND VOLUME CONTROL SYSTEM SOURCES
 
LETDOWN MIXED BED DEMINERALIZER
 
Isotope                          Activity
(&#xb5;Ci/cc)
 
Cr-51                            3.30E+01 Mn-54                            3.36E+01 Fe-55                            2 23E+02 Fe-59                            2.80E+01 Co-58                            6.94E+02 Co-60                            2.96E+02 Br-83                            6.26E-01 Br-84                            7.47E-02 Br-85                            8.17E-04 Rb-86                            1.15E+00 Rb-88                            1.78E+00 Sr-89                            2.30E+01 Sr-90                            3.25E+00 Sr-91                            3.42E-01 Y-89M                            2.07E-03 Y-90                            3.21E+00 Y-91M                            2.25E-01 Y-91                            5.13E+00 Y-93                            1.88E-02 Zr-95                            2.40E+00 Nb-95                            3.44E+00 Nb-95M                          2.40E+00 Mo-99                            3.02E+02 Tc-99                            9.73E-04 Ru-103                          1.12E+00 Ru-106                          1.14E+00 Te-125M                          1.04E+00 Te-127M                          1.72E+01 Te-127                          1.74E+01 Te-129M                          2.99E+01 Te-129                          1.92E+01 Te-131M                          1.95E+00 Te-131                          3.64E-01 Te-132                          5.49E+01 I-129                            1.12E-06 I-130                            1.41E+00 I-131                            2.83E+03 I-132                            6.73E+01 I-133                            4.30E+02 I-134                            2.21E+00 I-135                            6.92E+01 Cs-134                          3.88E+03 Cs-135                          1.73E-05
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-6 (Sheet 2)
 
Isotope                          Activity
(&#xb5;Ci/cc)
 
Cs-136                          1.22E+02 Cs-137                          3.26E+03 Ba-137M                          3.08E+03 Ba-140                          3.66E+00 La-140                          3.98E+00 Ce-141                          2.96E+00 Ce-143                          7.16E-02 Ce-144                          7.20E+00 Pr-143                          9.56E-01 Pr-144                          7.20E+00
 
Reactor Coolant Filter
 
Specific Source Gamma Energy              Strength (Mev/  )                (Mev/cc-sec)
 
0.8                5.7 x 107 1.3                1.5 x 107
 
NOTE:
 
All other demineralizers and filters throughout the plant are shielded with these same source terms, since these are the most radioactive.
 
Rev. 0
 
WOLF CREEK
 
TABLE 12.2-6 (Sheet 5)
 
Boric Acid Tanks
 
Isotope                          Activities
(&#xb5;Ci/cc)
 
Cr-51                            4.16E-06 Mn-54                            8.56E-07 Fe-55                            4.49E-06 Fe-59                            2.41E-06 Co-58                            4.08E-05 Co-60                            5.63E-06 Br-83                            2.54E-08 Br-84                            1.92E-12 Br-85                            negligible Rb-86                            8.40E-06 Rb-88                            1.87E-09 Rb-89                            5.94E-14 Sr-89                            7.32E-06 Sr-90                            5.90E-08 Sr-91                            6.39E-08 Y-89M                            6.58E-10 Y-90                            5.00E-08 Y-91M                            4.27E-08 Y-91                            3.54E-07 Y-93                            3.61E-09 Zr-95                            1.52E-07 Nb-95                            1.35E-07 Nb-95M                          1.18E-07 Mo-99                            8.72E-05 Tc-99                            1.45E-11 Ru-103                          1.06E-07 Ru-106                          2.77E-08 Te-125M                          7.23E-08 Te-127M                          7.40E-07 Te-127                          7.68E-07 Te-129M                          3.21E-06 Te-129                          2.05E-06 Te-131M                          5.16E-07 Te-131                          9.41E-08 Te-132                          1.59E-05 I-129                            5.08E-15 I-130                            3.06E-07 I-131                            7.46E-04 I-132                            1.67E-05 I-133                            1.09E-04 I-134                            2.26E-09 I-135                            1.06E-05 Cs-134                          3.56E-03 Cs-135                          3.71E-09
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-6 (Sheet 6)
 
Boric Acid Tanks
 
Isotope                          Activities
(&#xb5;Ci/cc)
 
Cs-136                          1.llE-03 Cs-137                          2.59E-03 Cs-138                          7.90E-08 Ba-137M                          2.45E-03 Ba-140                          7.64E-07 La-140                          8.17E-07 Ce-141                          3.31E-07 Ce-143                          1.92E-08 Ce-144                          1.90E-07 Pr-143                          1.93E-07 Pr-144                          1.90E-07
 
Rev. 0 WOLF            CREEK
 
TABLE            12.2-7
 
FUEL          STORAGE          POOL            WATER            ACTIVITIES**
 
Isotope*                                                                                                                                  Activities
(&#xb5;Ci/cc)
 
Co-58                                                                                                                                                                                    1.3E-04
 
Co-60                                                                                                                                                                                    4.0E-04
 
Cs-134                                                                                                                                                                        2.0E-05
 
Cs-137                                                                                                                                                                        2.0E-04
* Other            isotopes            will            be            present            in            much            lower            concentrations.
 
**            Typical            spent            fuel            pool            concentrations            at            operating            plants with            similar            cleanup            systems.
 
Rev.          14
 
WOLF CREEK
 
TABLE 12.2-9
 
CONSERVATIVE BASIS ACCUMULATED RADIOACTIVITY IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATION
 
Activity at Plant Shutdown Isotope                    (curies)
 
Kr-85                    15317.5
 
Kr-85m                  10.9
 
Kr-87                    1.05
 
Kr-88                    10.6
 
Xe-131m                  128.5
 
Xe-133                  16536
 
Xe-133m                  145
 
Xe-135                  82
 
Xe-135m                  0.035
 
Xe-138                  0.04
 
I-131                    0.207
 
I-132                    0.00095
 
I-133                    0.04
 
I-134                    0.000187
 
I-135                    0.0065
 
This 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.
 
Rev. 0
 
WOLF CREEK
 
TABLE 12.2-10 (Sheet 3)
 
FORT CALHOUN OPERATING DATA III. AUXILIARY BUILDING SAMPLING STATION FEEDS (REF. 6)
 
Station 1
: 1. Letdown heat exchanger room
: 2. Mechanical penetration area
: 3. Shutdown heat exchanger room
: 4. Valve room
: 5. Pipe penetration area
: 6. Personnel air lock area
: 7. Sampling room*
 
Station 2
: 1. Cask decon. room
: 2. Fuel arrival area
: 3. Fuel storage area
: 4. Drum storage area
: 5. Waste baler room
: 6. Spent resin storage room
: 7. Volume control tank room
: 8. Waste evaporator room
: 9. Waste holdup tank rooms
: 10. Spent fuel heat exchanger room
: 11. Safety injection pump rooms
: 12. Charging pump room
: 13. Charging pump valve room
: 14. Fuel pool area
 
Station 3
: 1. Waste decay tank rooms
: 2. Shutdown cooling heat exchanger room
: 3. Shutdown cooling heat exchanger valve room
: 4. Component heat exchanger room
 
Station 4
: 1. Spent fuel heat exchanger room
: 2. Safety injection pump rooms
: 3. Charging pump room*
: 4. Charging pump valve room
 
        *Principal source in the station.
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-11
 
PARAMETERS AND ASSUMPTIONS FOR CALCULATING AIRBORNE RADIOACTIVE CONCENTRATIONS (REF. 7)
 
A. Leak Rates                                    Pounds/Day
: 1. Equivalent reactor coolant leak into containment during power for noble gases                        5,300
: 2. Equivalent reactor coolant leak into containment for halogens                                      5.3
: 3. Equivalent reactor coolant leak into containment for other isotopes                                240 B. Evaporation Rates                              Pounds/Hr
: 1. From refueling pool into containment atmosphere (based on pool temperature of 120qF, and building air temperature of 70qF and 50 percent relative humidity and pool surface area of 1,500 square feet and 30 ft/minute flow parallel to the pool surface)                          499.5
 
canal, and connecting slots, into fuel  2.            From fuel storage pool, transfer building atmosphere during re-fueling (based on pool temperature of 137qF and building air temperature of 110qF and 95 percent relative humidity
: 2.                  1024 and with pool surface area of 2111.5 ft
: 3. From fuel storage pool into fuel building atmosphere during power (based on pool temperature of 95qF and building air temperature of 70qF and 50 percent RH and
: 2. 409 pool surface area of 2111.5 ft
 
Rev. 14 WOLF CREEK
 
TABLE 12.2-11 (Sheet 2)
C. Partition Factors            Halogens    Particulates    Tritium
: 1. Containment during power      1          .001            0.35
: 2. Fuel storage and refueling pool surfaces                1        Negligible        1
 
D. Ventilation Rates                                CFM
: 1. Containment during power                    4,000
: 2. Containment during refueling                20,000
: 3. Fuel building during power                  20,000
: 4. Fuel building during refueling              20,000  (Note 1)
E. Volumes of the Regions                            CF
: 1. Containment                                2.5 x 106
: 2. Fuel building                              8.2 x 105 F. Maximum Annual Individual Occupancy                                      Hrs/yr
: 1. Containment during power 5 hr/wk for 50 wks/year                      250
: 2. Containment during fuel handling 10 hr/day for ~ 6.25 days when the refueling pool is full of water          62.5
: 3. Fuel building during power 5 hr/week for 50 weeks/year                  250
: 4. Fuel building during refueling 10 hrs/day for ~ 10 days and 8 hrs/day for ~ 3 days                      125 G. Miscellaneous Information
: 1. Failed fuel percentage for fission products                            0.12
: 2. Reactor coolant specific activities        Table 11.1-1
 
Rev. 19 WOLF CREEK
 
TABLE 12.2-11 (Sheet 3)
: 3. Refueling and fuel storage pool con-        I-131      I-133 centrations ( Ci/gm).  (These are the maximum concentrations during refueling)                  3.21x10-5  2.32xlO-6
: 4. Fuel storage pool cleanup rate during power (gpm) (for conservatism, no cleanup of fuel storage pool or refueling pool is assumed during refueling)                                  300
: 5. Decay of isotopes in the pools              Not included for during refueling                              conservatism
: 6. Duration of refueling pool evaporation (hrs)                            150
: 7. Duration of fuel storage pool evaporation during refueling (hrs)          320
: 8. Duration of fuel storage pool evaporation during power (hours/year)        8440
: 9. 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)            24
: 10. Tritium release to environment via fuel building ventilation exhaust during refueling (Ci) (same bases as given for Item 9)                        104
: 11. Tritium release to environment via fuel building ventilation exhaust during power (Ci) (same bases as given for Item 9)                                                                                                                                                                                                                                                                        1094
 
NOTES:
: 1.                              The emergency exhaust flow rate may vary resulting in proportionally higher or lower airborne concentrations in the fuel building. Any difference is insignificant when compared to the allowable limits of 10 CFR 20.
 
Rev. 18 WOLF CREEK
 
TABLE 12.2-12
 
AIRBORNE RADIOACTIVITY CONCENTRATIONS
 
( Ci/cc)
 
Containment  Containment    Fuel Building    Fuel Building Nuclide      (power)    (refueling)      (power)        (refueling)
 
H-3        2.34E-7        4.7E-6        3.10E-6          1.66E-5 Cr-51      1.25E-12      negligible    negligible      negligible Mn-54      2.07E-13      negligible    negligible      negligible Fe-55      1.07E-12      negligible    negligible      negligible Fe-59      6.63E-13      negligible    negligible      negligible Co-58      1.06E-11      negligible    negligible      negligible Co-60      1.33E-12      negligible    negligible      negligible Br-83      1.76E-11      negligible    negligible      negligible Br-84      2.61E-12      negligible    negligible      negligible Kr-83m    6.60E-8        negligible    negligible      negligible Kr-85m    6.02E-7        negligible    negligible      negligible Kr-85      1.18E-7        negligible    negligible      negligible Kr-87      1.39E-7        negligible    negligible      negligible Kr-88      8.34E-7        negligible    negligible      negligible Kr-89      5.83E-10      negligible    negligible      negligible Rb-88      8.01E-7        negligible    negligible      negligible Rb-89      5.63E-10      negligible    negligible      negligible Sr-89      3.54E-12      negligible    negligible      negligible Sr-91      2.49E-13      negligible    negligible      negligible Mo-99      5.06E-11      negligible    negligible      negligible Te-127m    1.86E-13      negligible    negligible      negligible Te-127    4.01E-13      negligible    negligible      negligible Te-129m    9.26E-13      negligible    negligible      negligible Te-129    6.58E-13      negligible    negligible      negligible Te-131m    1.34E-12      negligible    negligible      negligible Te-131    2.69E-13      negligible    negligible      negligible Te-132    1.65E-11      negligible    negligible      negligible I-130      1.95E-11      negligible    negligible      negligible I-131      3.82E-9        2.14E-10      negligible      7.6E-10 I-132      3.66E-10      negligible    negligible      negligible I-133      4.14E-9        1.55E-11      negligible      5.5E-11 I-134      7.41E-11      negligible    negligible      negligible I-135      1.34E-9        negligible    negligible      negligible Xe-131m    2.75E-7        negligible    negligible      negligible Xe-133m    1.35E-6        negligible    negligible      negligible Xe-133    7.24E-5        negligible    negligible      negligible Xe-135m    7.64E-9        negligible    negligible      negligible Xe-135    2.55E-6        negligible    negligible      negligible Xe-137    1.26E-9        negligible    negligible      negligible Xe-138    2.22E-8        negligible    negligible      negligible
 
Rev. 0 WOLF CREEK
 
TABLE 12.2-12 (Sheet 2)
 
Containment  Containment    Fuel Building    Fuel Building Nuclide      (power)    (refueling)      (power)        (refueling)
 
Cs-134    1.67E-11      negligible    negligible      negligible Cs-136    8.48E-12      negligible    negligible      negligible Cs-137    1.20E-11      negligible    negligible      negligible Cs-138    2.07E-8        negligible    negligible      negligible Ba-137m    1.14E-11      negligible    negligible      negligible Ba-140    1.43E-13      negligible    negligible      negligible La-140    1.07E-13      negligible    negligible      negligible
 
NOTES:
: 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.
 
Xe-133 and I-131 air concentrations in the containment ventilation exhaust duct are expected to be <2 x 10 -2 and 1.6 x 10-6  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.
: 3.                              Airborne concentrations in the fuel building are based upon 9,000 cfm ventilation exhaust airflow rate during refueling. Actual flow rate may be lower resulting in proportionally higher concentrations. Any difference is insignificant when compared to the allowable limits of 10 CFR 20.
 
Rev. 19 WOLF CREEK
 
12.3  RADIATION PROTECTION DESIGN FEATURES
 
12.3.1  FACILITY DESIGN FEATURES
 
In this section, specific design features for maintaining personnel exposures ALARA, commensurate with the guidance given in Regulatory Guide 8.8, are discussed. For ISFSI design features, refer to the NUHOMS EOS System UFSAR, Docket 72-1042.
 
12.3.1.1  Plant Design Description for as Low as is Reasonably Achievable (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).
 
12.3.1.1.1  Common Equipment and Component Designs for ALARA
 
This 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 - To reduce exposure, spent liquid system radioactive filters are normally handled with a remote tool during removal from the filter housing and during transfer for packing and shipment from the site for disposal. 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 and compartment drainage capabilities.
The ventilation return ducts are equipped with a removable access panel through which a portable radiation monitor can be lowered into the filter compartment to obtain radiation levels within the compartment. Care has been taken to ensure that adequate space is provided to allow removing, cask loading, and transporting the cartridge to the solid radwaste building.
 
12.3-1 Rev. 35 WOLF CREEK
 
DEMINERALIZERS - Demineralizers for radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to spent resin storage tanks prior to processing 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 tank which, when removed, allows insertion of a portable radiation monitor to obtain 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 shielding design 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
 
12.3-2                        Rev. 5 WOLF CREEK
 
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 is not 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 extenders or shielding are provided so that personnel need not enter the radiation area for valve operation.
 
For infrequently operated equipment in design Zone E, most of the manual valves associated with safe operation, shutdown, and draining of the equipment are provided with remote-manual operators or reach rods. For a definition of the design 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.
 
All manually-operated valves in the filter and demineralizer valve compartments required for normal operation and shutdown have provisions for reach rods extending through or over the valve gallery wall. The valve gallery shield walls are designed for maximum expected filter activities.
 
12.3-3                        Rev. 16 WOLF CREEK
 
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 does not require entry and immediate replacement of the defective lamp, since sufficient illumination is still 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. 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 to minimize personnel exposure during sampling. The counting room and laboratory facilities are described in Section 12.5.
 
12.3-4                                                                                                                                                                                                                        Rev. 5 WOLF CREEK
 
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.
 
12.3.1.1.2  Common 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 permits easy decontamination.
 
PIPING - Pipes carrying radioactive materials are routed through controlled-access areas based on the anticipated 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 designed low radiation zones, shielded pipeways are provided. Whenever possible and practical, valves and instruments are not placed in the radioactive pipeways. 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 located in appropriate radiation design zone and restricted areas. Process piping is monitored where required to ensure that access is controlled to limit exposure (Section 12.5).
 
Piping is designed to minimize low points and dead legs where practicable.
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 piping joints is
 
12.3-5                        Rev. 12 WOLF CREEK
 
prohibited to eliminate potential crud traps for radioactive materials, unless specifically required by the piping design. Butt welds are used in lieu of socket welds in resin slurry and evaporator bottoms piping, unless specifically required by the piping design. Piping carrying resin slurries or evaporator bottoms is run vertically as much as possible, and large radius bends are utilized instead of elbows, unless specifically required by the piping design.
 
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.
 
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.
 
In controlled access areas where contamination is expected, airborne radiation monitoring equipment is provided (Section 12.3.4). 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.
 
12.3-6                        Rev. 5 WOLF CREEK
 
A decontamination facility, used for decontamination of removable components, is located adjacent to the auxiliary building. The decontamination system, shown in Figure 12.3-4, consists of a cask washdown pit in the Fuel Building, wash tanks, pumps, filters, spray booth, ultrasonic generator, 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 areas. Control panels are located in low designed radiation zones (Zones A or B).
 
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 areas for maintenance.
 
Labyrinth entranceway shields or shielding doors are provided for certain compartments from which radiation could stream to access 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 may be used, as needed. When this is not practicable, access to the high radiation areas are under the direct supervision of the unit health physics personnel in accordance with approved radiation work permits, when applicable.
 
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 possible interferences are taken into account.
 
12.3-7                        Rev. 19 WOLF CREEK
 
12.3.1.2  Radiation Zoning and Access Control
 
Access to areas inside the plant structures and plant yards is regulated and controlled as described in Section 12.5.2. 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 the plant buildings 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. These criteria were then used as the basis for the radiation shielding design. 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 Figure 12.3-2 (Sheets 1-6).
All frequently accessed areas, i.e., corridors, were shielded for Zone A or Zone B access. Periodically dose levels in various areas will exceed their radiation zone levels from plant conditions or stored items. During plant operation and refueling conditions the Health Physics staff will evaluate area access, monitor entry into areas, and update posting and entry requirements in accordance with 10 CFR 20. Quality Control and Procurement Quality may update postings that are required to perform Radiography and metal examination.
 
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 described in Sections 12.5.2 and 12.5.3.
 
Any area accessible to individuals which could result in an individual receiving a dose equivalent in excess of 5 mrem in 1 hour at 30 centimeters from the radiation 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 radiation areas where practicable. Locations of these barriers are shown on Figure 12.3-2. Any area accessible to individuals which could result in an individual receiving a dose equivalent in excess of 100 mrem in one (1) hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates is posted with the radiation symbol and the words, "CAUTION, HIGH RADIATION AREA."  High radiation areas are kept locked or barricaded if greater than 1000 mrem in one hour at 30 centimeters except during periods when access to the area is required in which case positive control is exercised over each individual entry. For these areas, in excess of 1000 mrem in one hour at 30 centimeters, that are located within large areas, such as containment, where no enclosure exists and where no enclosure can be reasonably constructed around the individual area, that area shall be barricaded, conspicuously posted and a flashing red light shall be activated as a warning device. Any area accessible to individuals which could result in an individual receiving an absorbed dose in excess of 500 rads in one hour at one meter from the source or from any surface that the radiation penetrates shall be locked or barricaded and posted "GRAVE DANGER, VERY HIGH RADIATION AREA."
 
12.3-8                        Rev. 19 WOLF CREEK
 
Whenever practicable, the measured radiation level and the location of the source is posted at the entry to radiation or high radiation areas.
 
The anticipated locations of radioactive equipment and the shield wall thicknesses are given on the general arrangement drawings in Section 1.2.
 
The normal radiation level in the counting room is the natural background level. Fly ash was specifically excluded from all concrete used for the counting room.
 
12.3.2  SHIELDING
 
The bases for the nuclear radiation shielding and the shielding configurations are discussed in this section. For shielding associated with the ISFSI including fuel loading, transfer and storage operations, refer to the NUHOMS EOS System UFSAR, Docket 72-1042 and the applicable site-specific dose calculations as documented in the 72.212 Evaluation Report.
 
12.3.2.1  Design Objectives
 
The 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 limits of 10 CFR 20. Shielding and equipment layout and design are considered in ensuring that exposures are kept ALARA during anticipated personnel activities in 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.
 
12.3-9                                                                                                                                                                                                                                            Rev. 35 WOLF CREEK
: c. To reduce potential equipment neutron activation and mitigate the possibility of radiation damage to materials.
: d. The control room is sufficiently shielded such that the direct dose plus the inhalation dose (calculated in Chapter 15.0) does not exceed the limits of GDC-19 (see Section 6.4 for a more detailed discussion).
 
12.3.2.2  General Shielding Design
 
Shielding is provided, as necessary, to attenuate postulated 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 Section 12.3.1.1.2.
 
The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 147 lb/ft3. 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).
 
12.3.2.2.1  Reactor Building Interior Shielding Design
 
During reactor operation, several areas inside the reactor building are High Radiation or Very High Radiation areas 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
 
12.3-10 Rev. 9 WOLF CREEK
 
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 shield design utilizes two types of shielding material contained in neutron shield segments hung from the support bars attached to the bottom of the seal plate of the permanent cavity seal ring (PCSR) at the RPV seal ledge. The bottom four inches of the shield is constructed of Reactor Experiments Type 207 material which is a flame resistant borated polyethylene.
The upper ten inches is type 277 which is a refractory material that resembles concrete in its final form. This shield configuration possesses bulk shielding characteristics superior to the shielding provided by the analyzed 12 inch thick water shield. MORSE analyses have been performed which predict that limited personnel access may be allowed at the operating floor during power operation. The MORSE analysis was performed using a water bag design. Since the shielding properties were not reduced by replacing water bags with the configuration described above, the analysis is still considered valid.
 
Components of the letdown system are located in shielded areas which are normally restricted access areas. Shielding is provided for N-16 delay piping, the excess letdown heat exchanger, and the regenerative heat exchanger.
 
After 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, but not limited, 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 a major 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.
 
12.3.2.2.2  Auxiliary Building Shielding Design
 
During 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 design zoning requirements of adjacent areas (Figure 12.3-2).
 
12.3-11    Rev. 6 WOLF CREEK
 
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 designed for Zone C 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 may temporarily reach high radiation 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 anticipated radiation levels of the adjacent areas.
 
12.3.2.2.3  Fuel Building Shielding Design
 
Spent 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 and the fuel transfer canal to ensure that radiation levels remain consistent with anticipated levels specified for adjacent areas. Water provides the shielding above the spent fuel assemblies during fuel handling operations.
 
The fuel pool cooling and cleanup system (FPCCS) (Section 9.1.3) shielding is based on the maximum activity discussed in Section 12.2.1 and the access requirements of adjacent areas. Equipment in the FPCC system shielded includes the FPCC heat exchangers, pumps, piping, filters, and demineralizers.
 
12.3.2.2.4  Radwaste Building Shielding Design
 
Shielding 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 pumps
: b. Liquid waste monitor tanks and pumps
: c. Chemical drain tank and pump
 
12.3-12    Rev. 14 WOLF CREEK
: d. Liquid waste, boron recycle, and secondary liquid waste evaporators
: e. Liquid Radwaste Demineralizer Skid
: f. Solid radwaste disposal and storage areas
: g. Evaporator bottoms tanks
: h. Liquid radwaste piping
: i. Radwaste filters and demineralizers
: j. Spent resin storage tank and pump
: k. Gaseous radwaste surge tank, recombiners, and compressors
: l. Gas decay tanks
 
Shielding is based upon operation with maximum activity conditions, as discussed in Sections 11.1, 11.2, 11.3, and 11.4.
 
Depending on the equipment in the compartments, the designed shielding varies from Zones B through E. Corridors are shielded for Zone B access, and operator areas for valve compartments are  designed for Zone C 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
 
12.3.2.2.5  Turbine Building Shielding Design
 
Radiation shielding is not required for any process equipment located in the turbine building, except for the condensate demineralizers. All other areas in the turbine building are classified Zone A.
 
12.3.2.2.6  Control 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
 
12.3-13    Rev. 8 WOLF CREEK
 
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.
 
12.3.2.2.7  Diesel Generator Building Shielding Design
 
There are no radiation sources in the diesel generator building. Therefore, no shielding is required within the building.
 
12.3.2.2.8  Miscellaneous Plant Areas and Plant Yard Areas
 
Sufficient shielding is designed for all plant buildings containing radiation sources so that anticipated radiation levels at the accessible outside surfaces of the buildings are maintained to meet Zone A levels. Plant yard areas which are frequently occupied by plant personnel are fully accessible during normal operation and shutdown. These areas are within the Radiation Control Area boundary and closed off from areas accessible to the general public. Radiation control Area boundaries established outside the restricted area are maintained less than or equal to 0.6 mrem/hr. The Radiation Control Area boundary for the ISFSI is maintained to this limit.
 
12.3.2.3  Shielding Calculational Methods
 
The shielding thicknesses provided to ensure compliance for plant radiation zoning are designed to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in Section 12.2.1. 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 pipes, 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 Table 12.3-1 (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
 
12.3-14 Rev. 35 WOLF CREEK
 
radiation dose through any wall. The shielding thickness was designed such that the aggregate postulated radiation level from all contributing sources, at this point, would be attenuated to meet the criteria of the adjoining radiation zone specifications.
 
Where shielded entryways to compartments containing high radiation sources are anticipated, 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 consistent with the radiation zone specified for each plant area.
 
12.3.3  VENTILATION
 
The 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.
 
12.3.3.1  Design Objectives
 
The 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."
 
12.3.3.2  Design Criteria
 
Design 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 10 CFR 20.
: b.                    During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary are ALARA and within the limits specified in 10 CFR 20 and 10 CFR 50.
: c.                    The plant siting dose guidelines of 10 CFR 50.67 are satisfied, following those hypothetical accidents described in Chapter 15.
 
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: 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 Section 6.4.
 
12.3.3.3  Design Guidelines
 
In order to accomplish the design objectives, the following guidelines are followed, wherever practicable.
 
12.3.3.3.1  Guidelines to Minimize Airborne Radioactivity
: a. 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.
: d. 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 graphite packing. 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.
 
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12.3.3.3.2  Guidelines to Control Airborne Radioactivity
: a. 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 than 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 Section 12.3.3.5.
: 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 post-accident safe shutdown 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.
 
12.3.3.3.3  Guidelines to Minimize Personnel Exposure from HVAC Equipment
: a. 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.
 
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: 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 is provided around each unit for anticipated maintenance, testing, and inspection.
: 2. Local cooling equipment servicing the normal building requirements is located in areas of low contamination potential (radiation Zones A or B)
(refer to Figure 12.3-2).
 
12.3.3.4  Design Description
 
The ventilation systems serving the following structures are considered to be potentially radioactive and are discussed in detail in Section 9.4:
: 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)
: f. 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 Section 6.4).
 
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Other structures (e.g., pump intake structures, administrative building, etc.)
contain no potential source of radioactivity and are not addressed in this chapter.
 
12.3.3.5  Air Cleaning System Design
 
The 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 Section 9.4.
 
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 and/or the general area are checked for radioactivity on a periodic basis with portable survey 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 is not 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.
 
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: 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.
: e.          The clear space for doors is a minimum of 3 feet by 7 feet.
: f.          The filters are designed with replaceable 2 feet by 2 feet 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.
 
12.3.4  AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION
 
12.3.4.1  Area Radiation Monitoring
 
The area radiation monitoring system (ARMS) is provided to supplement the personnel and area radiation survey provisions of the plant health physics program described in Section 12.5 and to ensure compliance with the personnel radiation protection guidelines of 10 CFR 20, 10 CFR 50, 10 CFR 70, and Regulatory Guides 8.2, 8.8, and 8.12.
 
12.3.4.1.1  Design Bases
 
The principal objectives and criteria of the ARMS are provided below.
 
12.3.4.1.1.1  Safety Design Bases
 
The 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 50.67 and, therefore, has no safety design bases. See Appendix 7A for a discussion of Regulatory Guide 1.97.
 
12.3.4.1.1.2  Power Generation Design Bases
 
POWER 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.
 
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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.
 
POWER GENERATION DESIGN BASIS FOUR - Certain monitors located near the fuel storage pool, new fuel storage vault, and cask handling area were originally installed to act as criticality alarm monitors in conformance with the requirements of 10 CFR 70, Regulatory Guides 8.5 and 8.12 and Standards ANSI/ANS-8.3-1979 and USAS N2.3-1967. The NRC issued an exemption to the requirements of 10CFR70.24 to WCNOC on June 24,1997. On November 12, 1998 the NRC issued 10CFR50.68, which provides eight criteria that may be followed in lieu of criticality monitoring per 10CFR70.24. One of these criteria require that radiation monitors are provided in storage areas when fuel is present to detect excessive radiation levels. These monitors will remain in place to provide prompt warning of high radiation. The monitors provide a hi-hi radiation alarm of 15 mrem/hr which will give prompt warning if high radiation occurs. These monitors are provided in accordance with GDC-63.
 
12.3.4.1.1.3  Codes and Standards
 
Codes and standards applicable to the area radiation monitors are indicated in Table 3.2-1.
 
12.3.4.1.2  System Description
 
12.3.4.1.2.1  General Description
 
The 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. Two ARMS (Technical Support Center (TSC) SDRE0043 and Emergency Offsite Facility (EOF) SDRE0044) are stand-alone units and do not transmit a signal or alarms to the control room.
 
12.3.4.1.2.2  Criteria 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 the Post-Accident sample room radiation monitor which has a range from 10 to 105 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.
 
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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 2.5 meV.
 
DIRECTIONAL DEPENDENCE - The dose rate readout does not vary more than 10 percent when exposed to a single point radiation source of approximately 1.0 MeV from any point in the frontal hemisphere of a horizontal plane.
 
ENVIRONMENTAL 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:
 
Temperature, F      60 to 120 Humidity, %RH      5 to 95 Pressure            Atmospheric
 
For instruments located inside the containment:
 
Temperature, F      50 to 150 Humidity, %RH      5 to 100 Pressure            2 psig
 
EXPOSURE LIFE - Each monitor located inside the containment maintains its characteristics up to an integrated dose of 107 rads. Each monitor located outside the containment maintains its characteristics up to an integrated dose of 106 rads.
 
12.3.4.1.2.3  Alarms
 
Each 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 causes an alarm on all three-alarm circuits.
The alarms must be manually reset and can be reset only after the alarm condition is corrected.
 
12.3.4.1.2.4  Check 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. The check source for most monitors can be operated from the control room. However, the TSC Monitor (SDRE0043) and the EOF Monitor (SDRE0044) must be operated locally.
 
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12.3.4.1.2.5  Power Supplies
 
The power supplies for all of the monitors are given in Table 12.3-4.
 
12.3.4.1.2.6  Calibration and Maintenance
 
The area radiation monitors are calibrated by the manufacturer, using a Cs-137 source. The manufacturer's calibration standards are traceable to National Institute of Standards and Technology primary calibration standard sources and are accurate to at least 5 percent. The 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. The frequency of calibration is established in station procedures.
 
The count rate response of each monitor to the remotely positionable check source supplied with each monitor is recorded by the manufacturer after the primary calibration.
 
Check source response and counter background is maintained in accordance with station procedures. Following repairs or modifications, the monitors are recalibrated at the plant with the secondary radionuclide standard.
 
12.3.4.1.2.7  Sensitivities
 
Each 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 10 mrem/hr.
 
12.3.4.1.2.8  Criteria 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 to unknowingly 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:
 
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: a. Zone A 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. For areas in which fuel is stored or radioactive waste systems and handling equipment are located, area radiation monitors are provided to detect conditions 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, Appendix A.
 
Other than the existing Area Monitors located in the fuel building, no additional Area Monitors are required for the ISFSI including fuel loading, transfer and storage operations.
 
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 corridors
: c.                    Auxiliary building corridors
: d.                    Fuel storage and handling area
: e.                    Radwaste pipe tunnel
: f.                    Railway access (fuel building)
 
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: g. Hot machine and hot instrument shops
: h. Containment
: i. Radwaste solidification area
: j. Sampling rooms and laboratory
 
12.3.4.1.2.9  Setpoints
 
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 setpoints for the individual area radiation monitors are provided in Table 12.3-2.
 
12.3.4.1.2.10  Safety 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 for most detectors. Most of these monitors cause an audible and visual alarm at the detector and in the main control room if the radiation levels exceed preset limits. However, the TSC Monitor (SDRE0043) and the EOF Monitor (SDRE0044) only provide local indication and alarm. 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.
 
12.3.4.2  Airborne Radioactivity Monitoring
 
Monitoring for the presence of airborne radioactivity inside the plant is necessary for the protection of plant personnel, in compliance with Regulatory Guide 8.2 and within the limits established by 10 CFR 20.
 
The airborne radioactivity monitors provide information necessary to ensure that gaseous, particulate, and iodine radioactivity do not exceed 10 DAC hours in areas occupied by the station personnel.
 
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The systems consist of permanently installed, continuous monitoring devices together with a program and provisions for specific sample collections and laboratory analyses.
 
12.3.4.2.1  Design Bases
 
The principal objectives and criteria of the airborne radiological monitoring systems (AiRMS) are provided below.
 
12.3.4.2.1.1  Safety Design Bases
 
SAFETY 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, Sections 6.2.4, 7.3.2, 9.4, and 11.5.
: b. Fuel building ventilation isolation, Sections 7.3.3, 9.4.2, and 11.5.
: c. Control room intake isolation, Sections 6.4.1, 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, and 11.5.
 
The safety evaluation of these systems is discussed in Section 7.3.
 
These monitors also serve for inplant worker protection, and this function is discussed at length in this section.
 
12.3.4.2.1.2  Power 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.
 
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POWER GENERATION DESIGN BASIS TWO - The airborne radioactivity monitors are designed to detect 10 DAC-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 Section 5.2.5. The containment purge monitors serve as a backup to the containment atmosphere monitors while the purge is in operation.
 
12.3.4.2.1.3  Codes and Standards
 
Codes and standards applicable to the airborne radioactivity monitors are indicated in Table 3.2-1. The monitors listed in Section 12.3.4.2.1.1 have additional codes and standards applied due to their safety-related functions discussed in other sections, as noted above.
 
12.3.4.2.2  System Description
 
12.3.4.2.2.1  General Description
 
12.3.4.2.2.1.1  Data 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.
 
The local microprocessors for monitors which perform safety functions (control room ventilation, fuel building ventilation, containment atmosphere, and containment purge monitors, refer to Section 11.5) are wired directly to individual indicators located on the seismic Category I 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 are not affected by signals or conditions existing in the nonsafety portion of the system.
 
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The control room microprocessor provides controls and indication for the AiRMS.
Indication is via a CRT located in the control room. The signals from each monitor may also be recorded on a system printer. All of the monitors are recorded on a tape cassette. The safety-related monitors are also recorded on analog strip chart recorders.
 
12.3.4.2.2.1.2  Selection Criteria for Airborne Monitors
 
12.3.4.2.2.1.2.1  Introduction
 
The 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.
 
12.3.4.2.2.1.2.2  Sampling 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.
 
12.3.4.2.2.1.2.3  Detection 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.
 
12.3.4.2.2.1.2.4  Instrumentation 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.
 
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The 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 Table 12.3-3. The high-alarm points are based on the most restrictive isotopes which are expected to be present.
 
12.3.4.2.2.1.3  Criteria for Airborne Radioactivity Monitor Locations
 
The 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 the exhaust from normally accessible personnel operating areas for which there is a potential for airborne radioactivity.
: b. Areas not normally accessed may be 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.
 
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: 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 10 CFR 20 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 airborne radioactivity limits in the area with the lowest ventilation flow.
: e.                    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, Figure 12.3-2.
 
Other than the existing airborne radioactivity monitors located in the fuel building, no additional monitors are required for the ISFSI including fuel loading, transfer and storage operations.
 
12.3.4.2.2.1.4  Alarms
 
Each 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 causes 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 plant computer (audible and visual), and the AiRMS CRT (visual). In addition, the safety-related channels have individual alarm lights on the safety-related indicators on the AiRMS control panel. The 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.
 
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12.3.4.2.2.1.5  Check 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.
 
12.3.4.2.2.1.6  Power Supplies
 
All Class IE inplant AiRMS are powered from Class IE motor control centers.
The power supplies for all of the monitors are given in Table 12.3-4.
 
12.3.4.2.2.1.7  Calibration and Maintenance
 
The 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 Institute of Standards and Technology primary calibration standard sources and are accurate to at least 5 percent.
The source detector geometry during this primary calibration 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. The frequency of calibration is established in station procedures.
 
The count rate response of each continuous monitor to remotely positionable check sources supplied with each monitor is recorded by the manufacturer after the primary calibration. Check sources response and counter background is maintained in accordance with station procedures. Following repairs or modifications, the monitors are recalibrated at the plant with the secondary radionuclide standards.
 
12.3.4.2.2.1.8  Sensitivities
 
The AiRMS is capable of detecting 10 DAC-hours of airborne radioactivity.
 
The most restrictive isotope for each type of monitor is that isotope with the lowest worker derived air concentration (WDAC), as defined in Table 1, Column 3, of Appendix B to 10 CFR 20.1001 - 20.2402.
 
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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 Table 1, Column 3, of Appendix B to 10 CFR 20.1001 - 20.2402 or Technical Specification limits, considering dilution.
 
The sensitivity of the airborne radioactivity monitors is based on a 95-percent confidence level for 0.5 MeV beta or gamma radiation in a 1 mr/hr gamma radiation background at standard pressure and ambient temperature.
 
The fixed volume noble gas detector assemblies have a minimum detectable concentration of 2 x 10-7          P Ci/cc, using Kr-85/Xe-133 as the limiting isotope.
 
The fixed filter particulate detector assemblies have a minimum detectable concentration of 1 x 10-11          P Ci/cc, using Cs-137 as the limiting isotope. The filter assembly has a collection efficiency of 99 percent for particles of 0.3 micron or larger.
 
The charcoal filter halogen detector assemblies have a minimum detectable concentration of 1 x 10-11          P Ci/cc, using I-131 as the limiting isotope. The charcoal filter assembly has a collection efficiency of at least 95 percent for iodine.
 
12.3.4.2.2.1.9  Ranges and Setpoints
 
The ranges of the various airborne radioactivity monitors were chosen based on the detection of radioactivity in concentrations ranging from 10 DAC-hours or lower in compartments served up to those from postulated accidents.
 
The fixed volume noble gas detector assemblies have a range of 10-7 to 10-2 P Ci/cc.
 
The fixed filter particulate detector assemblies have a range of 10-12 to 10-7 P Ci/cc.
 
The charcoal filter halogen detector assemblies have a range of 10-11 to 10-6 P Ci/cc.
 
The setpoints are chosen to alert the operators to airborne radioactivity that might be present so that 10 CFR 20 limits on worker exposure or Technical Specification limits are not exceeded.
 
12.3-32 Rev. 13 WOLF CREEK
 
The setpoints for control of ventilation are discussed in Section 7.3.
 
The setpoints on the monitors used for reactor coolant pressure boundary leakage detection are discussed in Section 5.2.5.
 
The ranges and setpoints for the airborne radioactivity monitors are provided in Table 12.3-3.
 
12.3.4.2.2.1.10  Expected System Parameters
 
The expected ranges of system parameters, such as flow rate, composition, and concentrations, are summarized in Table 12.3-3. Detailed information on the individual HVAC systems can be found in Section 9.4.
 
12.3.4.2.2.2  Monitoring Systems
 
The systems discussed in the following sections are summarized in Table 12.3-3.
 
12.3.4.2.2.2.1  Access Control Area Ventilation Exhaust Radioactivity Monitor
 
The 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.
 
Further determination of the source and the corrective action to be taken are based on monitoring with portable detection and sampling equipment.
 
12.3-33 Rev. 5 WOLF CREEK
 
Indication for this system is provided on the AiRMS CRT in the control room.
 
12.3.4.2.2.2.2  Radwaste Building Ventilation Exhaust Radioactivity Monitor
 
The radwaste building ventilation exhaust radioactivity monitor, 0-GH-RE-22, continuously monitors for particulate radioactivity in the exhaust duct upstream of the radwaste exhaust filter adsorber. 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 is obtained. After passing through the fixed filter detector assembly and the pumping system, the sample is discharged back to the duct. The cartridge filter is removed periodically for laboratory isotopic analyses. Monitoring upstream of the filter adsorber provides the most rapid response to airborne radioactivity in the system.
 
The high and high-high alarms function to alert the operator to airborne particulate radioactivity in the radwaste building. Indication for this monitor is provided on the AiRMS CRT in the control room.
 
If required, grab samples are utilized to determine airborne radioactivity levels and iodine concentrations in specific areas to aid in the determination of the source of the release.
 
12.3.4.2.2.2.3  Waste Gas Decay Tank Area Ventilation Exhaust Radioactivity Monitor
 
The 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 DAC for the most restrictive isotope expected to be present (Kr-85 or Xe-133).
 
12.3-34 Rev. 7 WOLF CREEK
 
Back up for this monitor is provided by the radwaste building exhaust and effluent monitors.
 
Indication of this monitor is provided on the AiRMS CRT in the control room.
 
12.3.4.2.2.2.4  Auxiliary Building Ventilation Exhaust Reactivity 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 can be removed for laboratory isotopic analyses.
 
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 CRT in the control room.
 
If required, grab samples can be utilized to determine airborne radioactivity levels and iodine concentrations in specific areas to aid in the determination of the source of the release.
 
12.3.4.2.2.2.5  Containment Atmosphere Radioactivity Monitors
 
The 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 Sections 5.2.5, 7.3, 9.4, and 11.5.
 
12.3-35 Rev. 5 WOLF CREEK
 
The containment atmosphere radioactivity monitors are seismic Category I systems and completely redundant.
 
The high and high-high alarms alert the operators to high airborne radioactivity in the containment atmosphere.
 
12.3.4.2.2.2.6  Containment Purge System Radioactivity Monitors
 
The 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 Sections 5.2.5, 7.3, 9.4, and 11.5.
 
The purge monitors are seismic Category I and completely redundant.
 
The sample points are located outside the containment between the containment isolation dampers and the containment purge filter adsorber unit.
 
Each 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.
 
12.3-36 Rev. 5 WOLF CREEK
 
Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the AiRMS CRT 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.
 
12.3.4.2.2.2.7  Control Room Ventilation Radioactivity Monitors
 
The 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 Category I 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.
 
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 Sections 6.4, 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 CRT in the control room.
 
12.3.4.2.2.2.8  Fuel Building Ventilation Exhaust Radioactivity Monitors
 
The 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 Sections 9.4 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.
 
12.3-37 Rev. 8 WOLF CREEK
 
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 Sections 7.3 and 11.5.
 
Indication is provided by individual indicators on the AiRMS control panel and, through isolated signals, by the AiRMS CRT in the control room.
 
12.3.4.2.2.3  Safety Evaluation
 
Due 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 Category I with Class IE power supplies. These monitors all have safety-related control functions which are described in Section 7.3.
 
The following monitors are located upstream of filters and therefore, are effective for inplant personnel protection:
: a. Containment atmosphere
: b. Containment purge
: c. Control room supply
: d. Fuel building exhaust
: e. Auxiliary building exhaust
: f. Radwaste building exhaust
: g. Access control area exhaust
: h. Waste gas decay tank room exhaust
: i. Portable monitor
 
All the inplant areas where the potential for airborne radioactivity exists are, therefore, monitored. The process and effluent radioactivity monitors are discussed in Section 11.5.
 
The AiRMS is adequate and sufficient to ensure personnel protection from airborne radioactivity. The system provides indication to the operator that airborne radioactivity exists in several possible areas. The location of the airborne radioactivity can then be further identified by using portable air samplers to collect general area air samples.
 
12.3-38 Rev. 5 WOLF CREEK
 
The combination of the AiRMS, in conjunction with administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns throughout 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 very radioactive 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 restricts the possibility of personnel exposure to airborne radioactivity in continuous occupancy areas.
: c.                    Prior to entry for work in airborne, or potentially airborne areas, Health Physics will take appropriate air samples. Authorization must be obtained before entry.
Prior to entry, a high-volume portable air sampler may be used by the health physics group to collect a representative air sample. Gaseous, iodine, and particulate activity of the area will be analyzed before entry as applicable.
: d.                    Health physics programs are discussed in Section 12.5.
 
12.3-39                                                                                                                                                                                                                                  Rev. 5 WOLF CREEK
 
12.
 
==3.5  REFERENCES==
: 1. R. L. Walker and M. Grotenhuis, A Summary of Shielding Constants for Concrete, ANL-6443 (November 1961).
: 2. C. M. Lederer, et. al., Table of Isotopes, Lawrence Radiation Laboratory, University of California (March 1968).
: 3. R. G. Jaeger, et. al., Engineering Compendium on Radiation Shielding, Shielding Fundamentals and Methods, I (1968).
: 4. G. W. Goldstein, X-Ray Attenuation Coefficients from 10 keV to 100 MeV, National Bureau of Standards Circular 583 (issued April 10, 1957).
: 5. Reactor Physics Constants, Argonne National Laboratory, ANL-5800 (July 1963).
: 6. J. F. Kircher and R. E. Bowman, Effects of Radiation on Materials and Components (March 1964).
: 7. T. Rockwell, Reactor Shielding Design Manual, D. Van Nostrand Co., New York (1956).
: 8. C. R. Tipton, Jr., Reactor Handbook, Vol. I, Materials, second edition (1962).
: 9. H. Soodak, Reactor Handbook, Vol. III, Part A, Physics, second edition (1962).
: 10. E. P. Blizzard and L. S. Abbott, Reactor Handbook Vol.
111, Part B, Shielding, second edition (1962).
: 11. N. M. Schaeffer, Reactor Shielding for Nuclear Engineers, TID-25951 (1967).
: 12. D. S. Duncan and A. B. Spear, Grace I -An IBM 704-709 Program Design for Computing Gamma Ray Attenuation and Heating in Reactor Shields, Atomics International (June 1959).
: 13. D. S. Duncan and A. B. Spear, Grace II - An IBM 709 Program for Computing Gamma Ray Attenuation and Heating in Cylindrical and Spherical Geometries, Atomics International (November 1959).
: 14. W. W. Engle, Jr., A User's Manual for ANISN:  A One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering, Union Carbide Corporation, Report No. K-1693 (1967).
: 15. E. D. Arnold and B. F. Maskewitz, SDC, A Shielding Design Code for Fuel Handling Facilities, ORNL-3041 (March 1966).
 
12.3-40 Rev. 5 WOLF CREEK
: 16. Richard E. Malenfant, "QAD, A Series of Point-Kernel General-Purpose Shielding Programs," Los Alamos Scientific Laboratory, LA 3573 (October 1966).
: 17. D. A. Klopp, NAP - Multigroup Time-Dependent Neutron Activation Prediction Code, IITRI-A6088-21 (January 1966).
: 18. E. A. Straker, P. N. Stevens, D. C. Irving, and V. R.
Crain, MORSE - A Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code, ORNL-4585 (September 1970).
: 19. W. A. Rhoades and F. R. Mynatt, The DOT III Two-Dimensional Discrete Ordinates Transport Code, ORNL-TM-4280 (1973).
: 20. R. E. Malenfant, "G3 A General Purpose Gamma-Ray Scattering Program," Los Alamos Scientific Laboratory, LA 5176 (June 1973).
: 21. M. J. Bell, "ORIGEN - The ORNL Generation and Depletion Code," Oak Ridge National Laboratory, ORNL-4628 (May 1973).
 
22                      Letter from J. C. Stone, USNRC, to O. L. Maynard, WCNOC dated June 24, 1997, Request For Exemption From 10 CFR 70.24 Criticality Monitoring Requirements - Wolf Creek Generating Station (TAC NO.
M89161).
 
12.3-41    Rev. 11 WOLF CREEK
 
TABLE 12.3-1
 
LIST OF COMPUTER CODES USED IN SHIELDING DESIGN CALCULATIONS
 
GRACE I        Multigroup, multiregion, gamma-ray attenuation code used to compute gamma heating and gamma dose rates in slab geometry (Ref. 12).
 
GRACE II      Multigroup, 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).
 
ANISN          Multigroup, multiregion code solving the Boltzman transport equation for neutrons or gamma-rays in one dimensional slab, cylindrical, or spherical geometry (Ref. 14).
 
SDC            Multigroup, multiregion, Kernal integration gamma-ray, shield design code which calculates dose rates for 13 geometry options (Ref. 15).
 
QAD            Multigroup, multiregion, three-dimensional, point Kernal code which calculates fast neutron and gamma-ray dose and heat generation rates (Ref. 16).
 
NAP            Determines activation emission source strengths as a function of neutron exposure and decay time (Ref.
17).
 
MORSE-CG      Three-dimensional Monte Carlo neutron and gamma ray general transport code (Ref. 18).
 
DOT III        Two-dimensional neutron, gamma ray, discrete ordinate, transport code (Ref. 19).
 
ORIGEN        Isotope generation and depletion code which solves equations of radioactive growth and decay for isotopes of arbitrary coupling (Ref. 21).
 
G3            A general purpose gamma-ray scattering code (Ref.
20).
 
Rev. 0
 
WOLF CREEK
 
TABLE 12.3-4
 
POWER SUPPLIES FOR AREA AND IN-PLANT AIRBORNE MONITORS
 
Area Radiation Monitors
 
Normal                                                                                                                            Restored After Power                                                                                                                                      Loss of Offsite Monitor Number                                                                                                                  Supply    Power                                Remarks
 
0-SD-RE-1                                                                                                                                                              Supplied by                                                                                                                            Yes 0-SD-RE-2                                                                                                                                                              regulated in-0-SD-RE-3                                                                                                                                                              strumentation 0-SD-RE-4                                                                                                                                                              ac power which 0-SD-RE-5                                                                                                                                                              is supplied from 0-SD-RE-6                                                                                                                                                              the diesel gener-0-SD-RE-7                                                                                                                                                              tors on loss of 0-SD-RE-8                                                                                                                                                              offsite power.
0-SD-RE-9 0-SD-RE-10 0-SD-RE-11 0-SD-RE-12 0-SD-RE-13 0-SD-RE-14 0-SD-RE-15 0-SD-RE-16 0-SD-RE-17 0-SD-RE-18 0-SD-RE-19 0-SD-RE-20 0-SD-RE-21 0-SD-RE-22 0-SD-RE-23 0-SD-RE-24 0-SD-RE-25 0-SD-RE-26 0-SD-RE-27 0-SD-RE-28 0-SD-RE-29 0-SD-RE-30 0-SD-RE-31 0-SD-RE-32 0-SD-RE-33 0-SD-RE-34 0-SD-RE-35 0-SD-RE-36 0-SD-RE-37 0-SD-RE-38 0-SD-RE-39 0-SD-RE-40 0-SD-RE-41 0-SD-RE-42 0-SD-RE-47 0-SD-RE-43                                                                                                                                                      Supplied by TSC                                                                  Yes emergency generator on loss of offsite power.
0-SD-RE-44                                                                                                                                                      Supplied by EOF                                                                  Yes emergency generator on loss of offsite power
 
Rev. 13 WOLF CREEK
 
TABLE 12.3-4 (Sheet 2)
 
In-Plant Airborne Radioactivity Monitors (Class IE)
 
Normal                                                                  Restored After Power                                                                          Loss of Offsite Monitor Number                                                                                                                  Supply                                                Power                                                                                Remarks
 
Containment                                                                                                                                            Class IE MCCs                                                Yes atmosphere O-GT-RE-31 O-GT-RE-32
 
Containment                                                                                                                                            Class IE MCCs                                                Yes purge system O-GT-RE-22 O-GT-RE-33
 
Fuel building                                                                                                                          Class IE MCCs                                                Yes exhaust O-GG-RE-27 O-GG-RE-28
 
Control room                                                                                                                                    Class IE MCCs                                                Yes air supply O-GK-RE-04 O-GK-RE-05
 
In-Plant Airborne Radioactivity Monitors (Non-IE)
 
Auxiliary building                                                                            Non-IE MCCs                                                  No                                                                                                                                    Power is lost to ventilation                                                                                                                                                                                                                                                                                                                                                                                                                                                        system also, so exhaust                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            monitor reading is O-GL-RE-60                                                                                                                                                                                                                                                                                                                                                                                                                                                                  not meaningful.
 
Radwaste building                                                                                      Non-IE MCCs                                          No                                                                                                                                    Power is lost to exhaust                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            system also, so O-GH-RE-22                                                                                                                                                                                                                                                                                                                                                                                                                                                                  monitor reading is not meaningful.
 
Access control                                                                                                                  Non-IE MCCS                                                  No                                                                                                                                    Power is lost to area ventilation                                                                                                                                                                                                                                                                                                                                                                                                            system also, so exhaust                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            monitor reading O-GK-RE-41                                                                                                                                                                                                                                                                                                                                                                                                                                                                  is not meaningful.
 
Waste gas decay                                                                                                        Non-IE MCCs                                                  No                                                                                                                                    Power is lost to tank area venti-                                                                                                                                                                                                                                                                                                                                                                                                            system also, so lation exhaust                                                                                                                                                                                                                                                                                                                                                                                                                              monitor reading O-GH-RE-23                                                                                                                                                                                                                                                                                                                                                                                                                                                                  is not meaningful.
 
Rev. 0
(                                                                                                                                                                                                                        c
 
  ;----    _..,.,. -""-"" ~ ..... '\\                                                        .                    ~LF      CREEK
                                                                                                                /        l~
I                                                    -',~~~~              N
                                \\sTEEL        SHIE.LO OOOR.                                                        ''1***    I                                                      *,*:          -
..,._ -    *-  -------------                                                                                          -            ____ .__,__ t:*f!.~:t                                          r- Rt:MOTE:    "--EC~.O.'IIC~L OPE RA.TOR    (. TYP)
I                                                ~
r-H
 
I                                                                                          r$ f: E      Of. 'TAll. '~>..'
lif.<EET  I (Tll"\\        I I                        I
:!f*~~    ;; 0                            *,.,. *). .
                                                                                                                                                                                              .. Jt..
* I                                                                              . '    I I                                                                        Iilli:                                                                        I                                I~  ;r                                                )
I: II H J~UJ    I                                          I                                                                                              J                                                                                                                                                                                            \\ '
 
                                                                                                                                                                    ~    1:::-:::.::
                                                                                                                                                          ----_.,r"      t-  ---          1-
                                                                                                                                              ~"'r---.  '                                                                                        I
 
t-----                              J                                    )
1------*
                                                                                                                                                        *~        ~r      t-----                                            I                    '
tl.. 'Zood* 0"    J        n                                      :.J: b.. ...ltl ~.
f' :t.1                                                                I  !  lef    ~ I !~ I I                        ~
                                                                                                                    ~~-.,.:*  I                                                              1          I ~ 1: : I Y**        I                                                                                                              ~ ~,
                                                                                                                    . ... ~*'~* ,                              SI.EE\\If. tT'fP.}~          : I :::    I I I* I I                          "* ~
f            . <<                                                        ILL    ~:1  j_ I .! I I
                                                                                                                    *.! -~~                        .                                              !        1    ll
-Ht--+-+----                -----+-                  f-+"'~--              ---It----+-++                      a    ~ :.~-~:*.                      '                            J                                                        l -~
                                                        ~;    -~                                              <'                                                                    SECTION              ~-/\\
 
tHERMAL      tlE<ioENUATIO~        DEMU.IUA.L11.U.,~          A ....J DEMINERALIZEB PLAN        YIEW AREA                        EL. AUX. BLQG. 2000'-o*
: l. Slf.lV!
* S(M. 90 PIPit                                                    frGUit(    12.l*l
                                                                                                                            . SIC ETCH *11.*                                                      hrJCAl      VAL V(  COIIPUTI'I(llt TYPICAl ltAI>IOo\\CTIVl SMA.FT $\\.,UV! StR.VtCl fOit                                      ~IIUIIG(tUU ISNHf II
 
Rev.        0 WOLF  CREEK                          REMOTE  MECHANICAL OPERATOR (TYP.)
                                                                    ~                                      VALVE  COMPARTMENT
.:~ ....                                                                                                    FILTER VAULT
/tt~;* ,..
 
I              I I
L-,          r-J I          I I          I I          I I
I          I
 
SLEEVE          (TYP.)
 
SECTION "A-All
 
WOLF    CREEK UPDATED    SAFETY    ANALYSIS      REPORT PLAN      VIEW    AUX.        BLDG.                                                          FIGURE                  12.3-1 FILTER      AREA                EL. 2000'-0"                                          TYPICAL      VALVE    COMPARTMENT ARRANGEMENT (SHEET            2)
Rev. 0 c                                                                                c                                                                                    c
 
                                                                                                  ~L.F    CREEK
                                                ......... ......... ~
                                                '. ... *. 0. "'
,_. ::::-* ':'~~-t:"t::!* ***+----*-*---
 
-  F .. .:_-: -:-; . .::_. ----o----------**
 
  -==.-:: -:-:.: ..
 
                                                                                          ... .;~."' ~ .            *;.-:
fo-                          I I
v~~-~~~------~~~~-4    I
 
                                                                                                ~liT IIUW SI.IIKL,.._                            ll'lNT ltl$111 lfotaK TO.NI(("'lluaaV}
 
                                                                                                                                        ~ ... ___ _
                                                                                          't                    I
 
IL!OOO'*o*      ...              ~      . ..... ..:
                                                                                          ~.~ .. u  :/ I            . .....
 
SECTION A-A
 
R.ev. 0                                            WOLF      CREEK UPDA~ED        SAFETY      ANALYSIS        REPORT SPENT PLAN RESIN VIEW STORAGE RAQWA5IE AREA BLDG,                                                              , EL.2000*0 *
(PRIMARY)                                                                                                            FIGURE      12.3-1 TYPICAL ARRANGEMENT VALVE COHPART~ENT (SHEET      3)
IDLF    CREEK
 
I
        *~~-~~.:~~----~--------;*~:.Fr------.
 
COMVR.a'!o\\QR. WJ.SU.Ga., 1
 
                                                                                                                                                  *:~~,. ______ ..._ __ -,..,_
        !                                  ~"*+----,----+
I ::*';------* -.. ...,---;---:.;.,. ~-~
 
eJ.T-.t.YrltMY~            I ltEeOM\\.1~
                                "'----4, .. J .. ,t--<---
 
                ~T~'=.;;::::=::====i~t:~:;t=: t* ===-N=::;;==t
 
A      I                                  "f**
Li                              ---+.:~:h~'-1---Z:".j.:-.::1 ====~=--=::i
 
                                                    ,--------,!.~.* ..... *: -~*
                                        . :=*:f; I
 
SECTION A*A
 
WOLF      CREEK UPDATED        SAFETY        ANALYSIS        REPORT Rev. 0                                        FIGURE        12.3-1
 
TYPICAL ARRANGEMENT VALVE COMPARTMENTTY (SHEET                4) c                                                  c
 
IDLF CREEK
                                                      ~1:..--~*l!, \\----.**. *"*r--------------; ...
 
                        \\
                      ~**
SM1PLE STATION SJ-14~
 
                ~-.
 
SAMPlE
_j                                        STATION SJ-143.
 
PLAN  VIEW -AUX. BUILDING .
SAMPLING ROOM    EL. 2000'-0                                SECTION@
 
WOLF  CREEK UPDATED  SAFETY  ANALYSIS  REPORT R.ev. o fiGURE  12.3-1 TYPICAL . ARRANGEMENT VALVE COMPARTMENT (SHEET  5)
_)'!) /
 
                  .. -- ..... -** I t T ~- ----
          -.. --                I L-.. ........                                                                                  .
                      *---i " ~      /' /    /
                      '/~f(~ ..- I        "'-
                      -<          I          ""-
                          --...... ............ I ' .........,
 
:*./"                                                                                                                              Rev. 0
                    .. -4\\ ,_                                                                                  WOLFCREEK UPDATED SAf.ETY ANALYSIS REPORT
 
FIGURE 12.3-3
 
CONTROL ROOM ISOMETRIC WOLF CREEK
 
12.4  DOSE ASSESSMENT
 
Radiation exposures in the plant are primarily due to direct radiation from components and equipment containing radioactive fluids. In addition, in some plant 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.
 
The ISFSI represents another source of direct radiation from the spent fuel placed in dry storage. For radiation exposures and dose estimates associated with the ISFSI including fuel loading, transfer and storage operations, refer to the NUHOMS EOS System UFSAR, Docket 72-1042 and the applicable site-specific dose calculations as documented in the 72.212 Evaluation Report.
 
Radiation exposures to operating personnel will be within 10 CFR 20 limits.
Radiation protection design features described in Section 12.3 and the health physics program outlined in Section 12.5 assure that the occupational radiation exposures (ORE) to operating personnel during operation and anticipated operational occurrences are as low as is reasonably achievable (ALARA).
 
12.4.1  EXPOSURES WITHIN THE PLANT
 
12.4.1.1  Direct Radiation Dose Estimates
 
Annual 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 manhour occupancy requirements for WCGS are given in Table 12.4-12.
 
12.4-1                                                                                                                                                                                                                              Rev. 35 WOLF CREEK
: e.        Table 12.4-13 provides an estimate of the distribution of the annual man-rem according to work function.
 
The precision of the man-rem estimate is of secondary importance. That estimate's relationship to actual man-rem doses received during subsequent plant operation will depend primarily on operating experience and maintenance and repair problems encountered rather than on design projections, however precise.The maximum and expected average dose rates in the plant radiation areas are given below:
 
Zone          Maximum Dose Rate  Expected Average Dose Rate
 
(mrem/hr)              (mrem/hr)
 
A                  0.5                        0.1 B                  2.5                        0.5 C                  15                          2.5 D                100                        15 E                >100                        100+
 
The maximum expected 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 minimizes crud buildup. However, it should be recognized that expected dose rates in various radiation zones and the actual 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 mrem/hr in the E Zone is used since these personnel generally are 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 Section 12.3 and the health physics program outlined in Section 12.5, it is expected that exposures due to special maintenance are minimized. However, an annual exposure of 150 man-rem 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. 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.
 
Assuming a work year of 2080 hours, the total estimated annual occupancy time for an individual in different work functions in various radiation zones is as follows:
 
12.4-2                                                                                                                                                                                                                                      Rev. 13 WOLF CREEK
: a.        Routine operation and surveillance:  The total occupancy time for an individual 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 occupancy times in various radiation zones is listed in Table 12.4-12.
: b.        Routine maintenance:  The total occupancy time for an individual involved in this work function in various radiation zones is expected to be 2,080 hrs/yr.
Individuals 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 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 an individual involved in this work function in various radiation zones is expected to be 320 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, health physics chemistry, and routine maintenance. The expected occupancy time for an individual involved in the refueling operation is about 160 hrs/yr. The average dose per individual in this category is estimated to be 0.5 rem over the occupancy period. The individuals 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 initially results in occupancies in radiation Zone C.
However, the operation of the cleanup systems results 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.
 
12.4-3                                                                                                                                                                                                                                      Rev. 14 WOLF CREEK
: e.        Special maintenance:  The total occupancy time for an individual involved in special maintenance problems, generally unanticipated, in various radiation zones is expected to be 320 hrs/yr, and the dose rate per individual 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 are performed by personnel involved in work functions such as routine operation and surveillance, health physics, chemistry, and routine maintenance. The total times expended by personnel in various radiation levels for the category are given in Table 12.4-1. The expected occupancy time for an individual involved in radwaste processing is about 2080 hrs/yr. The annual average dose per individual in this category is estimated to be about 3 rem.
 
12.4.1.2  Airborne Radioactivity Dose Estimates
 
12.4.1.2.1  Exposures Due to Airborne Radioactivity
 
As already discussed in Section 12.2.2, negligible airborne concentrations and consequently negligible airborne radioactivity 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 minimizes exposures to airborne activity and ensures 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 are 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.
 
12.4.1.2.2  Model for Calculating Exposures Due to Airborne Radioactive Sources
 
Thyroid and inhalation doses are calculated using the following equation:
 
DO  =      CI (BR).t.DFOI
 
i where
 
CI  = Airborne concentration of the ith nuclide in pci/cm3
 
12.4-4 Rev. 8 WOLF CREEK
 
and
 
BR  = Breathing rate for occupational worker in cm3/sec = 347
 
t  = Time duration of inhalation of radioactivity contaminated air in seconds
 
DFOI = Dose factor for adult for organ 0 (thyroid
 
or lung) via inhalation in mrem/pCi inhaled for the
 
ith isotope (These dose factors are taken from Regulatory Guide 1.109, Rev. 1)
 
DO = Dose in millirems to organ 0 due to inhalation
 
Total body submersion doses are calculated, using a finite cloud model.
 
Annual man-rem exposures due to airborne radioactivity are calculated using the following equation:
 
DO = (DR)O 10-3.h
 
where
 
(DR)O = Dose rate for organ in mrem/hr
 
and
 
h  = Annual occupancy in man-hours/yr D0 = Annual exposure in man-rem/yr
 
12.4.1.3  Illustrative Examples of Dose Assessment
 
Dose 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.
 
12.4-5 Rev. 5 WOLF CREEK
: a. Operation and Surveillance
 
The dose rates in the corridors and other normally occupied areas are expected to be much lower than the maximum radiation levels for each zone 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 annual exposures for this category range from 13-30 man-rem.
: b. Routine Maintenance
 
A number of examples of 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 Processing
 
Annual 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. Refueling
 
Based 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 Inspection
 
Inservice inspections at operating PWRs have taken much of the refueling outage, requiring from 3 to 38 men 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, performed per NEI 97-06, 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.
 
12.4-6 Rev. 24 WOLF CREEK
: f. Special Maintenance
 
Special 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 is used whenever practicable to minimize personnel exposures for this task.
Improvements in secondary system water chemistry and improved tube support plates have reduced the likelihood of the need to plug tubes.
 
The volatile chemical treatment to be employed for the secondary system greatly alleviates 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.
 
12.4-7 Rev. 5 WOLF CREEK
 
12.4.2  EXPOSURES AT LOCATIONS OUTSIDE PLANT STRUCTURES
 
12.4.2.1  Direct Radiation Dose Estimates
 
Direct radiation outside plant structures from the containment and the auxiliary, radwaste, and turbine buildings is negligible, compared to that from outside storage tanks. The principal sources of radioactivity not stored in the plant structures are the reactor makeup water storage tank, the refueling water storage tank, and the condensate storage tank. These tanks, shall be limited to the conditions outlined in Section 16.11.1.1. The dose rate at the nearest site boundary for the sectors in which these tanks are exposed has been calculated to be of the order of 10-5 mrem/yr. Using the projected populations for the year 2020 in the exposed sectors to a distance of 50 miles throughout the year results in a negligible population exposure of less than 10-3 man-rem.
 
12.4.2.2  Exposures Due to Airborne Radioactivity
 
Estimates of doses at the site boundary due to released activity are given in Section 11.3.3.
 
12.
 
==4.3  REFERENCES==
: 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.
 
12.4-8 Rev. 11 WOLF        CREEK
 
TABLE        12.4-1 ILLUSTRATIVE        EXAMPLES        OF        DOSE        ASSESSMENT
 
Operation        and        Surveillance:
 
Description        of                                                                                                  No.        of                                                                              No.        of                                                                                                                                                                                                                                                                  No.        of Task                                                                                                                                                                                                      Days          (1)                                                          Personnel          (2)                    Mrem/man-day                                                man-rem
 
Fuel        bldg.                                                                                                                                          Refueling                                                                              29                                                                                                  10-113                                                                                                                                6 Aux.        bldg.                                                                                                                                          Entire        yr                                                                    5-9                                                                                                            9-186                                                                                                                                2 Aux.        bldg.        equip.                                                                                                  18                                                                                                                      3                                                                                                                      7-65                                                                                                                                          0.2 test Ctmt.        (initial                                                                                                                                1-3                                                                                                  5-17                                                                                                                                                                                                                                                                  0.1-0.5 survey        after        SD)
Total        for        oper.        &                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              13-30 surv.
 
Routine        Maintenance
 
Misc.        instrument                                                                                                            7                                                                                                                                3                                                                                                            19-92                                                                                                                                          0.2 calibration        for pressurizer Aux.        bldg.        valves                                                                              19-22                                                                                        4-16                                                                                                  13-70                                                                                                                                                    2 Aux.        bldg.        general                                                                              8-45                                                                                        8-13                                                                                                            3-242                                                                                                            0.3-2 Aux.        bldg        instrument                                                                    41                                                                                                            11                                                                                                                      5-48                                                                                                                                          0.5maintenance calibration Ctmt.        decon.                                                                                                                      Refueling                                                                    33                                                                                                            21-170                                                                                                                                          7 Fuel        bldg.        decon.                                                                                                  46                                                                                                            5                                                                                                                                2-20                                                                                                                                          0.3 Radwaste        bldg.                                                                                                                                11                                                                                                            9                                                                                                                      28-137                                                                                                                                0.8 general        maintenance Radwaste        bldg.        decon.                                      21-24                                                                                                  3                                                                                                                      9-143                                                                                                                                                    4 Aux.        bldg.        decon.                                                                                        11                                                                                                                      4                                                                                                                                3-19                                                                                                                                          0.1 Ctmt.        valves                                                                                                                                          8-74                                                                              10-27                                                                                                  33-185                                                                                                            0.9-6 Ctmt.        instrument                                                                                                  8-65                                                                                        4-30                                                                                                            9-75                                                                                                                      0.3-5 calibration Replace        oil        in                                                                                                                                5                                                                                                                      5                                                                                                                      12-45                                                                                                                                          0.3 reactor        coolant pumps Remove        charcoal                                                                                                            4-5                                                                                                  7-36                                                                                                                                                                                                                                                                            0.1-0.5 filters        for        ctmt.
cleanup Replace        charcoal                                                                                                  6-45                                                                              2-20                                                                                                                                                                                                                                                                                      0.3-0.8 filters        for        ctmt.
cleanup Repair        dampers        &                                                                                                            8                                                                                                            4-5                                                                                                                                                                                                                                                                                      0.2-0.3 duct General        decon.        &                                                                                        29-68                                                                              33-71                                                                                                                                                                                                                                                                                                7-11 Decon.        refueling                                                                                        2-18                                                                                                  4-23                                                                                                                                                                                                                                                                                      0.1-6relamping canal
 
Rev.        0 WOLF        CREEK
 
TABLE        12.4-1        (Sheet        2)
Description        of                                                                                                  No.        of                                                                              No.        of                                                                                                                                                                                                                                                                  No.        of Task                                                                                                                                                                                                                  Days(1)                                                                      Personnel(2)                    Mrem/man-day                                                                        man-rem
 
Reactor        coolant                                                                                                            35                                                                                                  18-30                                                                                                                                                                                                                                                                                                          6 pump        motor        work Uncouple/couple                                                                                                            22                                                                                                  15-19                                                                                                                                                                                                                                                                                                0.8 reactor        coolant pumps Install/remove                                                                                                                      12                                                                                                            7-10                                                                                                                                                                                                                                                                                                0.3 reactor        coolant pump        scaffolding Repair        ctmt.        sump                                                                                        9-75                                                                                        9-30                                                                                                                                                                                                                                                                            0.4-4 pumps        &        level indicators Repair        pressurizer                                                                              5-31                                                                                        4-9                                                                                                                                                                                                                                                                                      0.2-0.3 relief        valves Replace        excore                                                                                                            23-42                                                                              15-30                                                                                                                                                                                                                                                                                                1-8 detectors Polar        crane                                                                                                                                                    9-78                                                                                        9-22                                                                                                                                                                                                                                                                            0.3-3 maintenance Ctmt.        pressure        test                                                                    60                                                                                                  26-32                                                                                                                                                                                                                                                                                                          2
          &        valve        repair Remove        &        clean        SI                                                                                                  3                                                                                                  14-25                                                                                                                                                                                                                                                                                                0.6 Check        &        repair                                                                                                            3-43                                                                                                  6-21                                                                                                                                                                                                                                                                            0.1-2system        strainers snubbers Remove        equip.                                                                                                                      5-10                                                                                                  2-9                                                                                                                                                                                                                                                        0.002-0.1 hatch Replace        equip.                                                                                                            2-3                                                                                                            3-7                                                                                                                                                                                                                                                                  0.03-0.04 hatch
 
Radwaste        Processing
 
Waste        drum        loading                                                          Refueling                                                                    7                                                                                                                      62-280                                                                                                                                          6 Loading        spent        resin                                                Refueling                                                          10                                                                                                                      78-760                                                                                                                                          4 casks Dry        waste        drum                                                                                                  Refueling                                                3-59                                                                                                                                                                                                                                                                                      0.2-4 handling Fuel          storage          pool                                                                                                  2                                                                                                            4                                                                                                                                                                                                                                                                                                                              0        6 filter        replace-ment Transferring        resin                                                                                        l                                                                                                            4-5                                                                                                                                                                                                                                                                                                          0.1
 
Refueling
 
Install        and        remove                                                                    7-16                                                                                                  7-30                                                                                                                                                                                                                                                                  0.4-0.9 reactor        cavity seal        ring Remove        transfer        tube                                      12-17                                                                                                  2-10                                                                                                                                                                                                                                                                  0.3-0.4flange
 
Rev.            14 WOLF        CREEK
 
TABLE        12.4-1        (Sheet        3)
Description        of                                                                                                  No.        of                                                                              No.        of                                                                                                                                                                                                                                                                  No.        of Task                                                                                                                                                                                                                  Days(1)                                                                      Personnel(2)                    Mrem/man-day                                                                      man-rem
 
Install        &        remove                                                                                        3-31                                                                                                  6-25                                                                                                                                                                                                                                                                                      0.2-3 upper        guide structure        lifting rig Rig        underwater                                                                                                            2-18                                                                                                  4-13                                                                                                                                                                                                                                                                  0.02-0.3 lights Flood        refueling                                                                                                            4                                                                                                                      5-9                                                                                                                                                                                                                                                                                                                    0.3 canal Fuel        handling                                                                                                            14-16                                                                                        49-92                                                                                                                                                                                                                                                                                                          3-7 (ctmt.)
Fuel        handling        (fuel                                                          5-28                                                                                        25-77                                                                                                                                                                                                                                                                                      0.5-1 bldg.)
Incore        instrumenta-                                                          3-36                                                                                        13-24                                                                                                                                                                                                                                                                                                          2-6 tion        removal Remove        &        replace                                                                                        6-12                                                                                                  9-24                                                                                                                                                                                                                                                                            0.2-0.3 missile        shield Remove        &        replace                                                                              11-13                                                                                                  9-27                                                                                                                                                                                                                                                                                      0.9-3 cable        &        ductwork to        reactor        head Remove        tool        access                                                          10-13                                                                                                  9-39                                                                                                                                                                                                                                                                                      0.8-9 flanges        &        un-couple        control rods Install        bullet                                                                                                            1-9                                                                                                            6-15                                                                                                                                                                                                                                                                                      0.7-2 noses Remove        bullet                                                                                                                      4-5                                                                                                            2-4                                                                                                                                                                                                                                                                            0.1-0.4 noses Remove        &        replace                                                                                        7-10                                                                                                  4-5                                                                                                                                                                                                                                                                            0.3-0.4 head        insulation RPV        studs        -        remove,                                                10-19                                                                                        26-40                                                                                                                                                                                                                                                                                                3-13 replace,        &        clean Install        &        remove                                                                                        2-11                                                                                                  3-12                                                                                                                                                                                                                                                                                      0.1-1 alignment        pins Install        lift        rig        &                                                                    2-10                                                                                        11-37                                                                                                                                                                                                                                                                                      0.6-3 remove        RPV        head Replace        RPV        head                                                                                        1-3                                                                                                            6-18                                                                                                                                                                                                                                                                                      0.4-7 Decon.        RPV        head                                                                                                  7-14                                                                                        10-24                                                                                                                                                                                                                                                                                                          3-5
          &        vessel        flanges
          &        replace        "0"        rings Remove        &        replace                                                                                        4-6                                                                                                  14-31                                                                                                                                                                                                                                                                                                3-8 holddown        ring Clean        stud        plugs        &                                                                    4-11                                                                                        12-26                                                                                                                                                                                                                                                                                                1-6 holes        and        install
          &        remove        plugs Clean        incore        in-                                                                                        9-28                                                                                        12-32                                                                                                                                                                                                                                                                                                2-6strument        flanges, studs,        &        nuts RPV        head        gasket                                                                                                  6-32                                                                                        19-35                                                                                                                                                                                                                                                                                                1-4 replacement
 
Rev.        0 WOLF        CREEK
 
TABLE        12.4-1        (Sheet        4)
Description        of                                                                                                  No.        of                                                                              No.        of                                                                                                                                                                                                                                                                  No.        of Task                                                                                                                                                                                                                  Days(1)                                                                      Personnel(2)                    Mrem/man-day                                                                    man-rem
 
Drain,        fill,        and                                                                                        2-9                                                                                                            5-10                                                                                                                                                                                                                                                                            0.1-0.7 vent        reactor coolant        system Install        handrails                                                                                        14                                                                                                            6-8                                                                                                                                                                                                                                                                                                                    0.1 around        refueling canal Total        for        refueling                                                15-30                                                                                                  8-61                                                                                                  32-347                                                                                                                                13-66
 
Inservice        Inspection
 
Steam        generator                                                                                                  7-51                                                                                        13-38                                                                                                  29-2500                                                                                                                                4-22 tubes Spent        fuel        sipping                                                                              20                                                                                                                      17                                                                                                            14-242                                                                                                                                                    3 Spent        fuel        inspection                                                57                                                                                                  22-30                                                                                                                                                                                                                                                                                                          1-2 Remove        &        replace                                                                                        4-14                                                                                        12-37                                                                                                                                                                                                                                                                                                          3-7 primary        manway covers Remove        &        replace                                                                                        7-8                                                                                                            4-16                                                                                                                                                                                                                                                                            0.4-0.8 secondary        man-Steam        generator                                                                                                  2-4                                                                                                  11-17                                                                                                                                                                                                                                                                            0.2-2way        covers secondary        side Install        primary                                                                                                            2                                                                                                                                3                                                                                                                                                                                                                                                                                                                              l loop        dams Clean        steam        generator                                                6                                                                                                                      4-12                                                                                                                                                                                                                                                                                                0.6 manway        studs Total        for        ISI                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                10-24
 
Special        Maintenance
 
Replace        upper        guide                                                69                                                                                                                      78                                                                                                                                17-785                                                                                                                      13 structure        on        control rods Sludge        lancing                                                                                                            5                                                                                                                      16                                                                                                                                19-105                                                                                                                                          2 of        steam        generator secondary        side Steam        generator                                                                                                  7-30                                                                                        5-10                                                                                                                                                                                                                                                                                      20-200 tube        plugging
 
Rev.        0 WOLF        CREEK
 
TABLE        12.4-1        (Sheet        5)
 
NOTES:
 
(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        a        particular        task.                  It        does        not        mean        that        all        of        the personnel        were        working        simultaneously        or        full        time        on        that        task.
 
Rev.        0 WOLF CREEK
 
TABLE 12.4-2
 
AVERAGE NUMBER OF PERSONNEL PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1,2,3)
 
Year            Number of Units    Average Number of Personnel/Unit
 
1969                    2                              131 1970                    2                              493 1971                    3                              250 1972                    4                              401 1973                    5                              774 1974                    10                            602 1975                    14                            548 1976                    25                            593 1977                    28                            642 1969-1977        Overall average                        580
 
NOTES:
: 1. Only PWRs at power levels >450 MWe have been considered, with the exception of San Onofre 1 (430 MWe)
: 2. Only PWRs that have been in commercial operation for at least 18 months at the end of the year have been considered. Multiple units have been considered for any year during which all units have been in commercial operation for at least 18 months prior to 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 data on the number of personnel were available for 1972.
 
Rev. 0 WOLF CREEK
 
TABLE 12.4-3
 
AVERAGE OCCUPATIONAL RADIATION EXPOSURE (MAN-REM DOSE)
PER PWR UNIT FOR THE PERIOD 1969-1977 (REF. 1,2,3)
 
Year            Number of Units      Average Man-Rem Dose/Unit
 
1969                    2                        74 1970                    2                        422 1971                    3                        274 1972                    5                        482 1973                    5                        610 1974                    10                      439 1975                    14                      460 1976                    25                      461 1977                    28                      422 1969-1977        Overall average                  441
 
NOTES:
: 1. Only PWRs at power levels >450 MWe have been considered, with the exception of San Onofre 1 (430 MWe).
: 2. Only PWRs that have been in commercial operation for at least 18 months at the end of the year have been considered. Multiple units have been considered for any year during which all units have been in commercial operation for at least 18 months prior to 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 data on the number of personnel were available for 1972.
 
Rev. 0 WOLF            CREEK
 
TABLE            12.4-4
 
AVERAGE            OCCUPATIONAL            RADIATION            EXPOSURE (MAN-REM            DOSE)            BASED            ON            PWR            PLANT            AGE (REF            1,            2,            3)
 
1/68      Conn.            Yankee                    -XXXXXXXXX 1/68      San            Onofre            1                    -XXXXXXXX 3/70      Ginna                                            XXXXXXXX 3/71      Robinson            2                          XXXXXXX 12/71      Pallisades                          XXXXXX 12/72      Maine            Yankee                    XXXXX 12/72      Surry            1                                    XXXXX 5/73      Surry            2                                    XXXXX 9/73      Fort            Calhoun            1              XXXX 6/74      Kewaunee                                                                                                                            X                              X                              X 9/74      Three            Mile            Isl.            1                X                              X                              X 12/74      Arkansas            1                                                                                                    X                              X                              X 4/75      Rancho            Seco                                                                                      -                              X                              X 8/75      Cook                                                                                                                                                                            X                              X 12/75      Millstone            Point            2                X                              X 5/76      Trojan                                                                                                                                                    X 12/76      St.            Lucie                                                                                                                X
 
NOTES:
: 1.      Only            PWRs            operating            at            power            levels            >450            MWe            have            been considered,            with            the            exception            of            San            Onofre            1            (430            MWe)
 
Rev.            0 WOLF CREEK
 
TABLE 12.4-4 (Sheet 2)
: 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).
: 3. Exposures reported to be in excess of 1,000 man-rem/year-unit are:
 
Years of Operation Reactor            at Year of Record      Man-Rem Dose
 
Palisades                  2                  1,109 Ginna                      3                  1,032 Surry 1 & 2                4                  3,165 Ginna                      5                  1,224 Robinson 2                  5                  1,142 Surry 1 & 2                5                  2,307
 
Rev. 0 WOLF CREEK
 
TABLE 12.4-5
 
DISTRIBUTION OF THE NUMBER OF PERSONNEL
(>100 MILLIREM/YR)
ACCORDING TO WORK FUNCTION
 
Percentage as Per  Percentage as Per NUREG-75/032 (for  NUREG-0109 (for the year 1974)      the year 1975)
Work Function                (Ref. 2)            (Ref. 1)
: 1. Reactor operations              19.2                  9.1
: 2. Routine maintenance              34.5                59.5
: 3. Inservice inspection              1.4                  4.1
: 4. Special maintenance              28.7                16.4
: 5. Waste processing                  2.1                  7.6
: 6. Refueling                        14.1                  3.3 See notes at end of Table 12.4-7.
 
Rev. 0 WOLF CREEK
 
TABLE 12.4-6
 
DISTRIBUTION OF PERSONNEL (>100 MILLIREM/YR)
ACCORDING TO EMPLOYEE CATEGORY
 
Percentage as Per    Percentage as Per NUREG-75/032 (for    NUREG-0109 (for the year 1974)        the year 1975)
Category                          (Ref 2)            (Ref 1)
: 1. Station employees                47.4                34.7
: 2. Utility employees                18.1                  6.1
: 3. Contract workers                34.5                59.2
 
See notes at end of Table 12.4-7.
 
Rev. 0 WOLF CREEK
 
TABLE 12.4-7 PERCENTAGES OF PERSONNEL DOSE BY WORK FUNCTION (REF 3)
 
Work Function                      Percent of Dose
 
1974      1975      1976      1977
 
Reactor operations      14.0%      10.8%      10.2%      10.6%
and surveillance
 
Routine maintenance      45.4%      52.6%      31.0%      28.9%
 
In-service inspection    2.7%      3.0%      6.0%      6.6%
 
Special maintenance      20.4%      19.0%      40.0%      41.4%
 
Waste processing          3.5%      6.9%      5.0%      5.9%
 
Refueling                14.0%      7.7%      7.9%      6.6%
 
Notes for Tables 12.4-5, 6 and 7:
: 1. PWRs and BWRs operating at all power levels were considered in compiling these tables.
: 2. Percentages for 1974 and 1975 are based on approximately 39 percent and 50 percent 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 in commercial 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 100 millirems, according to either work function or employee category, are not available for 1976 or 1977.
 
Rev. 0
 
WOLF CREEK
 
TABLE 12.4-10 AVERAGE INDIVIDUAL EXPOSURE BASED ON PWR PLANT AGE (REF. 1,2,3)
 
Year of                              Ave. Exposure  Ave. Exposure Operation  No. of  Ave. No. of      man-rem        per Individual (approx)    Units  Personnel/Unit    yr-unit          (rem-yr)
 
1        13        362              123              0.34
 
2        15        532              330              0.62
 
3        13        661              664              1.00
 
4          9        623              571              0.92
 
5          8        691              775              1.12
 
6          5        627              481              0.77
 
7          4        540              341              0.63
 
8          3        583              465              0.80
 
9          2        987              665              0.67
 
10        2        940              745              0.79
 
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:
 
Plant    Ginna        Surry 1 & 2        Robinson Year      Average      Average        Average Exposure Exposure      Exposure          in Rem/yr in Rem/yr    in Rem/yr
 
3        1.52            ---                ---
 
4        ---          1.15                ---
 
5        1.39          1.24                1.35
 
Rev. 0
 
WOLF CREEK
 
TABLE 12.4-12 ESTIMATES OF OCCUPANCY TIMES IN PLANT RADIATION AREAS AND GAMMA DOSES TO PLANT PERSONNEL Percentage              Hourly Dose    Yearly Dose of                    Rate          Rate        No. of      Annual Exposures Operation      Zone    Occupancy    Hrs/yr    (Rems/Hr)      (Rems/Yr)        men      (Man-Rem/yr-Unit)
 
Operation &      A          75        1,560    1 x 10-4          0.15          38              5.7 Surveillance B          21          437    5 x 10-4          0.21          38              8.3
 
C          3          62    2.5 x 10-3        0.15          38              5.9
 
D          0.9        19    1.5 x 10-2        0.28          38              10.8
 
E          0.1          2    1 x 10-1          0.20          38              7.6 Total                      100        2,080                      0.99                          38.3
 
Routine          A          75        1,560    1 x 10-4          0.15          28              4.4 Maintenance B          11          229    5 x 10-4          0.11          28              3.2
 
C          10.5        218    2.5 x 10-3        0.54          28              15.3
 
D          2.5        52    1.5 x 10-2        0.78          28              21.8
 
E          1          21    1 x 10-1          2.1            28              58.8 Total                      100        2,080                      3.68                          104
 
Radwaste        A          70        1,456    1 x 10-4          0.14            8              1.16 Processing B          20          416    5 x 10-4          0.2            8              0.68
 
C          8          166    2.5 x 10-3        0.41            8              3.28
 
D          1          21    1.5 x 10-2        0.31            8              2.52
 
E          1          21    1 x 10-1          2.1            8              16.8 Total                      100        2,080                      3.16                          24.4
 
Rev.        0 WOLF CREEK
 
TABLE 12.4-12 (Sheet 2)
Percentage              Hourly Dose    Yearly Dose of                    Rate          Rate        No. of      Annual Exposures Operation      Zone    Occupancy    Hrs/yr    (Rems/Hr)      (Rems/Yr)        men      (Man-Rem/yr-Unit)
 
Refueling        A          30          48    1 x 10-4          0.004          6              0.22
 
B          40          64    5 x 10-4          0.03            6              1.47
 
C          24          39    2.5 x 10-3        0.09            6              4.48
 
D          4            6    1.5 x 10-2        0.09            6              4.14
 
E          2            3    1 x 10-1          0.30            6              13.8 Total                      100          160                      0.51                          24.1
 
Inservice        A          55          176    1 x 10-4          0.02          46              0.81 Inspection (ISI)          B          36          115    5 x 10-4          0.06          46              2.64
 
C          7          23    2.5 x 10-3        0.06          46              2.61
 
D          1            3.5  1.5 x 10-2        0.05          46              2.30
 
E          1            3    2 x 10-1          0.67          46              30.67 Total                      100          320                      0.86                          39.03
 
Special          A          75          240    1 x 10-4          0.024        200              4.8 Maintenance B          20          64    5 x 10-4          0.03          200              6.4
 
C          3          10    2.5 x 10-3        0.025        200              5
 
D          1            3    1.5 x 10-2        0.04          200              9
 
E          1            3    2 x 10-1          0.6          200            120 Total                      100          320                      0.71                          145
 
Rev. 0 WOLF CREEK
 
TABLE 12.4-13
 
DISTRIBUTION OF DIRECT RADIATION MAN-REM DOSES ACCORDING TO WORK FUNCTIONS
 
Annual Exposures Operation                        (Man-rem/year)    Percentage
 
Operation and surveillance          38.3              10.5
 
Routine maintenance                104                28.5
 
Radwaste processing                  24.4                6.7
 
Refueling                            24.1                6.6
 
Inservice inspection                29.2                8
 
Special maintenance                145                  39.7
 
Total                              365                100
 
Rev. 0 WOLF CREEK
 
TABLE 12.4-14
 
ANNUAL OCCUPANCY IN PLANT AREAS CONTAINING AIRBORNE RADIOACTIVITY
 
hr/yr                    man-hours Building          per man    No. men          yr___
 
Containment
  -power          250          3            750
 
Containment
  -refueling      62.5        32          2,000
 
Fuel building
  -power          250          2            500
 
Fuel building
  -refueling      125          14          1,750
 
Rev. 0
 
WOLF                        CREEK 1400r-------------------------------------------------~~~-------------------------------------------,
 
UPDATED                                              SAFETY WOLF                            CREEK ANALYSIS                                                    REPORT 1200                                                                                                                                                                                                                                                                                                                                                    Rev.                            0                                                                                                FIGURE                    12.4-1 CUMULATIVE              AVERAGE                                                                                              OF                  ANNUAL EXPOSURE (TABLE                    12.4-11) BY                                YEARS                                  OF                  OPERATION PWRs                                                    -
(SHEET                      1)
 
1000
 
11.1      800 Cll 0
c
~
a:
z ct
:iE      600                                                                                                                                                                                                                                                                                                                                                                          Average For All                                            PWR's I
 
400
 
200
 
D                                                              1                                                                                                                                                                                                                                                            2                                                                                                                                                                                                                                                        3                                                                                                                                                                                                                                                          4 5                                                                                                                                                                                                                                                        6                                                                                                                                                                                                                                                          7                                                                                                                                                                                                                                                        8                                                                                                                                                                                                                                9 10 YEARS OF OPERATION 1400r--------------------------------------------                                                                                                                                                                                                                                        WOLF                    CREEK
 
WOLF                          CREEK UPDATED                                        SAFETY                                    ANALYSIS                                              REPORT 1200                                                                                                                                                                                                                                                                                                                                                                                                                                                FIGURE                      12.4-1 Rev.                          0                                  CUMULATIVE                                                        AVERAGE                                          OF              ANNUAL EXPOSURE (TABLE                      12.4-11) BY                        YEARS                              OF                OPERATION PWRs                                            -
(SHEET                                  2)
 
1000
 
w BOO V) 0 0
:E w
a:
z
<t
:E      600
 
Average  For All PWR 's
                                                                                                                                                                                                                                                                                                                                                                                              ---.....___.....____                                                            j 400
 
200
 
0                                                                                                                      2                                                                                                                                                                                                                                                                        3                                                                                                                                                                                                                                                                4                                                                                                                                                                                                                                                                          5 6                                                                                                                                                                                                                                          7                                                                                                                                                                                                                                                                8                                                                                                                                                                                                                                      9 10 YEARS OF                                OPERATION 14oor-------------------------------------------------                                                                                                                                                                                                                            WOLF                  CREEK
 
WOLF                      CREEK UPDATED                                    SAFETY                              ANALYSIS                                        REPORT
 
1200                                                                                                                                                                                                                                                                                    Rev.                      0                                                                                FIGURE                      12.4-1
 
r        CUMULATIVE                                                AVERAGE                                    OF              ANNUAL LXPOSURE (TABLE                      12.4-11) BY                YEARS                          OF              OPERATION PWRs                                        -
(SHEET                              3) 1000
 
Surry            1  &      2 w        800                                                                                                                                                                                                        Per      Unit (I) 0 c
:2 w
a;:
2
<{
:2      600
 
400
 
200                                                                                                                                                                    \\
Fort          Calhoun                    1
 
0                                                  1                                                                                                                                                                                                                                                                      2 3                                                                                                                                                                                                                                                                4 5 6                                                                                                                                                                                                                                          7                                                                                                                                                                                                                                                            8                                                                                                                                                                                                                                            9 10 YEARS                    OF OPERATION 1400~---------------------------------------------                                                                                                                                                                                                                                                  WOLF                    CREEK                    -------------------------------------,
 
WOLF                        CREEK UPDATED                                      SAFETY                                  ANALYSIS                                              REPORT
 
1200                                                                                                                                                                                                                                                                                                                                Rev.                          0                                                                          FIGURE                  12.4-1 CUMULATIVE              AVERAGE                                                                                OF              ANNUAL rXPOSURE                                              BY              YEARS                              OF                OPERATION                                                  -
{TABLE                              12.4-11)                                                PWRs
{SHEET                4)
 
1000
 
w
~
c          800
~
'f 2
~
 
600
 
Average For All                                      PWR's I
400
 
200
 
1                                                                                                                                                                                                                                                        2                                                                                                                                                                                                                                                          3                                                                                                                                                                                                                                                          4 5                                                                                                                                                                                                                                                          6 7                                                                                                                                                                                                                                                        8                                                                                                                                                                                                                                9 10 YEARS OF OPERATION WOLF CREEK
 
12.5  HEALTH PHYSICS PROGRAM
 
12.5.1  ORGANIZATION
 
The WCGS Health Physics Program is established to provide an effective means of radiation protection for station personnel, visitors and the general public.
The program consists of:  A management philosophy which supports radiation protection and ALARA concepts, a site organizational structure, (See Chapter 13.0), qualified personnel to direct and implement the Health Physics Program, written procedures outlining acceptable radiation protection practices and the appropriate equipment and facilities necessary to support a comprehensive Health Physics effort. The program is developed and implemented through the applicable sections of the Code of Federal Regulations, Regulatory Guides and ANSI standards.  (See Table 12.1-1 for the applicable criteria).
 
The Health Physics Section is responsible for providing technical support to the WCGS Health Physics Program. This section is headed by the Manager Radiation Protection. At the time of commercial operation at least one individual within the section met the qualifications of ANSI N18.1-1971, Regulatory Guide 1.8, 8.2, 8.8 and 8.10.
 
The resume of the Manager Radiation Protection is provided in Section 13.1.
The qualifications of the individual, who fulfills the requirements as the Radiation Protection Manager, as specified in TS 5.3.1.2, are also included in Chapter 13.1.
 
The major responsibilities of this section include:
: 1)  Providing technical support to WCGS in the area of Health Physics, radwaste, and emergency programs during normal plant operation, outages and unanticipated events.
: 2)  Assisting in long range planning for WCGS based on new and proposed regulatory changes.
: 3)  Performance of ALARA design reviews/cost benefit analysis, exposure data trending, ALARA committee activities and technical reviews.
 
12.5-1 Rev. 30 WOLF CREEK
: 4) coordinating with all other station departments to provide health physics coverage for all activities that involve radiation or radioactive materials.
: 5) assuring that radiation protection related training provided to personnel is technically accurate and properly describes Station Health Physics practices.
: 6) enacting the site ALARA program.
: 7) providing a personnel radiation monitoring and health physics records management program.
: 8) providing radiation surveys and appropriate posting of station areas, posting of radiation zones, supplying follow up monitoring and maintaining records.
: 9) maintaining Health Physics equipment functional and calibrated.
 
10)providing equipment and a respiratory protection program for the station personnel.
 
11)assuring proper shipment and receipt of radioactive material.
 
Procedures are developed with special attention given to  maintaining personnel exposures to ALARA levels. Work in a radiation area for activities such as maintenance, inspection, refueling, and nonroutine operations is appropriately planned prior to the initiation of work so as to minimize exposures to as low as practicable levels. Where circumstances allow, specific exposure reduction techniques may be utilized as part of the procedures for work in radiation areas. Examples of these techniques include:
: a. Minimizing source strength and contamination levels by flushing equipment prior to performing maintenance.
: b. Minimizing radiation levels in work areas by the use of temporary shielding.
: c. Using remote handling equipment and other special tools.
: d. Minimizing discomfort of workers so that efficiency is increased and less time is spent in radiation areas.
 
12.5-2 Rev. 7 WOLF CREEK
 
Each individual is responsible for obeying the applicable site radiation protection procedures. A thorough, demonstrable knowledge of these is required for unescorted individuals prior to entry in the RCA. Individuals are further charged with following the guidance of the Health Physics staff during their indoctrination, and normal work tasks and any activity noted to be in violation of procedures is reported to the appropriate supervisor for resolution.
 
12.5.2  EQUIPMENT, INSTRUMENTATION AND FACILITIES
 
12.5.2.1  Health Physics Equipment and Instrumentation
 
Laboratory, fixed and portable equipment, and instrumentation are selected on the basis of: job requirements, usefulness, and characteristics such as sensitivity, response time, type of radiation detected, accuracy, dependability, and lifetime. Factors considered before selection include previous experience, recommendations and comments from other utilities, and guidelines in applicable ANSI Standards and Regulatory Guides. Laboratory equipment for Health Physics is stored in the Health Physics offices and labs, the Hot Laboratory and the Counting Room. Details on laboratory counting equipment are given in Table 12.5-1. Some equipment and instrumentation in these areas is shared jointly by Health Physics and Radiochemistry and includes multi-channel analyzers, scintillation counters, fume hoods, and other ancillary equipment and supplies needed to perform adequate Health Physics and Radiochemistry sample preparation and analysis.
 
A list of portable instrumentation with respect to instrument type, location, type of detector, range, accuracy and quantity is given in Table 12.5-2.
 
Calibration of Health Physics portable equipment and instrumentation is performed by qualified trained personnel using written procedures and using standards that are traceable to the National Institute of Standards and Technology. The frequency of calibration is established in station procedures.
 
A record of the calibration is produced for each calibration and is kept on file at the station. Instruments that are calibrated have a current calibration sticker attached and are stored separately from instruments that are out of service.
 
12.5-3 Rev. 13 WOLF CREEK
 
Portable instrumentation for routine use is stored in the RCA and at various locations throughout the Site in emergency kits, lockers, etc., as designated by Health Physics to meet survey instrumentation requirements.
 
The Radiological Respiratory Protection Program is the responsibility of the Manager Radiation Protection and complies with 10 CFR 20 Subpart H.
 
Selection and use of respiratory equipment, i.e., self contained breathing apparatus, airline equipment, full face and hood respirators, chemical cartridges, etc., is according to regulations established in 42 CFR, part 84 and includes only equipment approved by the NRC or the National Institute for Occupational Safety and Health's Equipment Certification Manual. Respirator storage is at the entrance to the RCA, in emergency cabinets, and at locations designated by Health Physics. Adequate quantities and types of respiratory equipment is available to support peak respiratory demands.
 
Airborne Radioactivity Monitoring is normally performed by several methods.
Permanently installed particulate iodine and gas monitors are described in Section 12.3.4.2. Mobile continuous airborne monitors (CAMS) are available in the Radiological Controlled Area to provide airborne monitoring for routine operations/ maintenance or abnormal occurrences. Portable high and low volume "grab" type air samples are used to supplement or substitute for the CAMS as an additional method of determining airborne concentrations. Calibration of the monitors use National Institute of Standards and Technology traceable sources.
 
Forty-two Remote Area Radiation Monitors are located throughout the plant in locations that optimize their use and meet the recommendations of Regulatory Guide 8.12. These monitors are also calibrated according to acceptable methods using National Institute of Standards and Technology traceable sources. A further description of these monitors is provided in Section 12.3.4.1.
 
Personnel defined as Monitored workers by 10CFR20 are normally monitored for Beta, Gamma and Neutron (as necessary) by Record Dose Dosimeters (RDDs) that have been accredited through the National Voluntary Laboratory Accreditation Program (NVLAP). Backup methods such as dose rate/stay time calculations or Electronic Alarming Dosimeter (EAD) dose can be used when the RDD is lost or damaged. RDDs are processed by a NVLAP Accredited facility on a routine basis (normally annually), when an individual terminates their monitored status or when their exposure status is in question.
 
12.5-4 Rev. 30 WOLF CREEK
 
Pocket ion chambers (PICs) and/or electronic dosimeters, for determining Gamma exposure are issued to personnel who are entering the Radiological Controlled Area. Selection, use, care and testing of PICs and electronic dosimeters follow the guidelines of ANSI N13.5 1972 and Regulatory Guides 8.4 and 8.14. Extremity badges or other additional dosimetry are issued to personnel on an individual basis as determined by Health Physics. Personnel dosimetry records are maintained using guidelines established in ANSI N13.6 1966 (R1972) and Regulatory Guide 8.7 and meet the requirements of 10 CFR 20.2106.
 
12.5.2.2  Health Physics Facilities
 
Figures 12.5-1, 12.5-2, 12.5-3, and 12.5-4 show Health Physics Facilities located in the Walter P. Chrysler Support Complex, Control, Radwaste, and Olive Ann Beech buildings respectively.
 
The main Access Control Facility is located in the Control Building at standard plant elevation 1984 feet. During routine working hours and outages, personnel designated to sign personnel into and out of the RCA are located at the RCA entrance near the Health Physics Office. Personnel requiring assistance to enter the RCA on the off-shift hours, weekends or holidays can contact the Shift Health Physics Technician. Normal entrance to the RCA is via a corridor outside the Health Physics Office. Normal exit from the RCA is through Personnel Contamination Monitors. During major outages, an auxiliary access may be established for the access/egress of additional personnel. Separate male and female toilets are located adjacent to the Health Physics Office Adjacent to the egress corridor is the decontamination area. This area is equipped with a decontamination shower and sinks used for personnel decontamination. Decontamination supplies are normally stored in this room.
Drains from the hot shower and sinks in these rooms are connected to the Liquid Radioactive Waste System.
 
Across from Rm 3208 is the Health Physics Count Room (Rm 3202). See Figure 12.5-2.
 
12.5-5 Rev. 32 WOLF CREEK
 
The hot laboratory and counting rooms are adjacent to the Health Physics Office. Equipment and instrumentation stored and used therein are described in Section 12.5.2.1. The hot laboratory and counting room sinks and floor drains are connected to the radwaste system and the fume hoods and room ventilation system are connected to the access control area ventilation system.
 
A tool and equipment decontamination room is located in the Hot Machine Shop on elevation 2000'. Ultrasonic cleaners and steam spray booths are located in the Decontamination Room and may be used to remove radioactive material from items before maintenance or repair. Depending on contamination levels, air sampling and Respiratory Protection may be used during these decontamination activities.
The drains in this room are connected to the Liquid Radioactive Waste System.
 
Anti-contamination clothing will be maintained in sufficient quantity to support plant work activities. Protective clothing is stored in plant locations designated by Health Physics. Protective clothing requirements are specified by Health Physics and may include:
: 1)  Coveralls - Cotton and Disposable
: 2)  Laboratory Coats
: 3)  Caps and Hoods
: 4)  Rubber and Plastic Shoe Covers
: 5)  Cotton, Plastic and Rubber Gloves
: 6)  Plastic Suits
 
The RP Supervisory, ALARA, Radwaste, Operations, Technician Group and Dosimetry areas are on the first floor of the Olive Ann Beech building (See Figure 12.5-
: 4) and contains the RDD processing area and the HP Records Office. The Walter P. Chrysler Support Complex (See Figure 12.5-1) contains the HP instrument repair and calibration facility, Whole Body Counter, and respiratory facility.
Portable instrument calibration is performed with sources traceable to National Institute of Standards and Technology.
 
The Bioassay Program follows the guidelines of Regulatory Guide 8.9. Whole body counting is used as required to determine the effectiveness of the Respiratory Protection Program and to assess the internal exposure of individuals who are involved in activities that have the potential for inhalation, ingestion or absorption of radioactive material. Excreta analyses may also be used to verify the uptake of radioactive material. Whole body counting and bioassay analysis is performed by Health Physics or contract personnel.
 
12.5-6 Rev. 31 WOLF CREEK
 
An additional Sample Room is located in the Radwaste Building at elevation 2000 feet and is for obtaining samples of the Radwaste Systems for analysis. Sample analysis and counting equipment may be located in this area. For details see Figure 12.5-3.
 
Conformance to Regulatory Guide 8.2 is addressed in Section 12.3.4.1. The means by which the recommendations of Regulatory Guide 8.8 are implemented are discussed at length throughout Chapter 12.0. Regulatory Guide 1.97 is discussed in Table 7.5-4.
 
12.5.3  PROCEDURES
 
ALARA Regulatory Guides 8.8 and 8.10 are an integral part of Health Physics procedures and policy developed by the WCGS staff. The use of qualified and experienced personnel in developing and implementing procedures is a tool used to keep exposures ALARA. Detailed written procedures, including applicable instructions and precautions will conform to 10 CFR 20.
 
The type, frequency, and location, of radiation surveys are outlined in procedures and are conducted in a manner that assures that exposure is ALARA.
Survey requirements are delineated in Health Physics procedures and consist of combinations of radiation level, contamination level, and atmospheric particulate and/or radioiodine concentrations. Contamination surveys normally use the "smear" or "swipe" test and are taken at locations that are dependent on factors such as location, occupancy factor, and potential radiological hazard. Decontamination, using acceptable methods and techniques may be performed on areas and equipment to reduce personnel exposure and contamination levels. Areas that cannot be cleaned using these decontamination practices are posted and barricaded per Health Physics procedures. Entry and exit of these areas are controlled through the use of the RWP System. A posting of radiological conditions outside the Health Physics Office at the Controlled Access Area entrance point is available for the information of personnel entering the RCA. Placards, acceptable posting methods and physical controls are outlined in Health Physics procedures that specify proper methods that are in compliance with 10 CFR 20 and Regulatory Guide 8.38, section 1.5. The Health Physics Exposure Records System established is in accordance with federal regulations and follows guidelines established in Regulatory Guide 8.7.
 
Administrative procedures are employed to effectively control employee exposures and maintain station doses ALARA and will follow the guidance of Regulatory Guide 8.2.
 
12.5-7 Rev. 18 WOLF CREEK
 
The RWP System is used to specify personnel who may enter the RCA and prescribe the required clothing, dosimetry, respiratory equipment, special instructions, descriptions and information that is relevant to providing proper radiological surveillance and control. The RWP System is outlined in detail in the Health Physics procedures.
 
Airborne radioactivity is normally controlled through the use of engineering controls, i.e., the installed ventilation systems, HEPA and activated charcoal filters. The use of respiratory protection, decontamination, glove boxes, tents, etc., may be used to further reduce the possibility of personnel exposure to airborne activity in excess of 10 CFR 20 limits.
 
The Radiological Respiratory Protection Program is outlined in the Health Physics procedures, meets the requirements of 10 CFR 20 Subpart H and follows the guidance of Regulatory Guide 8.15. To ensure that the Respiratory Protection Program is functioning properly, a method of determining internal exposure, such as whole body counting or bioassay, as discussed in Section 12.5.2, is established.
 
A Radiation Worker Training Program is developed and implemented for instruction of personnel who have unescorted access to the RCA. The training involved is approved by the Manager Radiation Protection through a Supervisor Radiation Protection and includes information needed to perform work in the RCA. Training includes instruction on station rules and practices; state, local and federal regulations; the basics of radiological health; biological effects of radiation; and ALARA concepts and philosophies. Periodic retraining is conducted. Posting of notices, instructions and reports to the plant workers is in accordance with 10 CFR Part 19. Personnel who work at the site, whose duties do not require the handling of radioactive material or enter the RCA, are instructed as to why they may not enter such areas. Plant workers receive prenatal radiation exposure instructions as recommended in Regulatory Guide 8.13.
 
Personnel monitoring, including internal and external, with the associated record keeping system, is discussed in Section 12.5.2.
 
Conformance to Regulatory Guides 8.9 and 8.14 is discussed in Section 12.5.2.
Regulatory Guide 1.8 is discussed in Section 12.1.3 and Chapter 13.0.
Compliance with Regulatory Guide 1.16 is discussed in Appendix 3A, and report content is also discussed in WCGS Technical Specifications. Regulatory Guide 1.33 is discussed in Appendix 3A and Section 13.4. Appendix 3A provides compliance with Regulatory Guide 1.39.
 
12.5-8 Rev. 30
 
ELEVATOR--.......
 
COJtmNG CHEM ROOM              HOT LAICJtATCWt
 
HP SIGN            HP IN/OUT AREA        OFFICE ELECTR'CAL CHASE
 
HP              MEN's          EXIT CONF            TOILET        MONITOR ROOM                            AREA
                                                                                                -AREA        OF CHANGE HP LOUNGE        RP CAL LAB *2 ELECTRICAL                                                              CHASE CHASE ----1--L-
 
RP CAL LAB  *1              HP CONF ROOM AREA  OF                                                          ....,..__ __ __,_...,__ DECON NEA CHANGE_____.,.      HOT  JANillR CLOS&#xa3;T  --t-..L .                    CORRIDOR
 
RICAL
 
CONTROL      BUILDING AT    ELEVATION      1984
 
REV. 32 UPDATED  SAFETY WOLF                    CREEK ANALYSIS REPORT
 
FIGURE    12 .. 5-      2
 
HEALTH    PHYSICS    AREA    IN  THE CONTROL    BUILDING N                                            F
 
                *~. *:
      -------------------, I EQUIP. SPACE              I  I r-------------------~              -~ ,FD I
...      I I        ,...--,
I      r-L-.&.-,          I    SAMPLE I
I      I DESK  I      ~-      LAB
___ .J I  I        I        ,FO L.---~
 
NOO fAR  SAMPLE PANEL  ROOM
 
*-~**.
 
RADWASTE      BUILDING    AT  ELEVATION      2000
 
Rev .* 0 WOLFCREEK UPDATED SAFETY ANALYSIS              REPORT
 
FIGURE 12.5-3
 
SAMPLE LAB FACILITIES IN THE RADWAS'!"E BUILDING}}

Latest revision as of 17:00, 4 October 2024

Redacted Updated Safety Analysis Report (WCGS Usar), Revision 37, Chapter 12, Radiation Protection
ML24114A142
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 04/15/2024
From:
Wolf Creek
To:
Office of Nuclear Reactor Regulation
References
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