ML23067A164

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Technical Specifications, Chapter 5, Administrative Controls
ML23067A164
Person / Time
Site: Callaway Ameren icon.png
Issue date: 12/29/2022
From:
Union Electric Co, Ameren Missouri
To:
Office of Nuclear Reactor Regulation
Shared Package
ML23067A139 List: ... further results
References
ULNRC-06782
Download: ML23067A164 (1)
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Text

CHAPTER TABLE OF CONTENTS

CHAPTER 5.0

ADMINISTRATIVE CONTROLS

Section Page

5.1 Responsibility...................................................................................................... 5.0-1

5.2 Organization........................................................................................................ 5.0-2

5.2.1 Onsite and Offsite Organizations.................................................................. 5.0-2 5.2.2 Unit Staff........................................................................................................ 5.0-2

5.3 Unit Staff Qualifications....................................................................................... 5.0-4

5.4 Procedures.......................................................................................................... 5.0-5

5.5 Programs and Manuals........................................................................................ 5.0-6

5.5.1 Offsite Dose Calculation Manual (ODCM)..................................................... 5.0-6 5.5.2 Primary Coolant Sources Outside Containment............................................ 5.0-7 5.5.3 Not Used....................................................................................................... 5.0-7 5.5.4 Radioactive Effluent Controls Program......................................................... 5.0-7 5.5.5 Component Cyclic or Transient Limit............................................................ 5.0-9 5.5.6 Containment Tendon Surveillance Program................................................. 5.0-9 5.5.7 Reactor Coolant Pump Flywheel Inspection Program................................... 5.0-9 5.5.8 Not Used..................................................................................................... 5.0-10 5.5.9 Steam Generator (SG) Program.................................................................. 5.0-10 5.5.10 Secondary Water Chemistry Program......................................................... 5.0-14 5.5.11 Ventilation Filter Testing Program (VFTP).................................................. 5.0-14 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring P rogram.......... 5.0-16 5.5.13 Diesel Fuel Oil Testing Program................................................................. 5.0-17 5.5.14 Technical Specifications (TS) Bases Control Program............................... 5.0-18 5.5.15 Safety Function Determination Program (SFDP)........................................ 5.0-18 5.5.16 Containment Leakage Rate Testing Program............................................. 5.0-19 5.5.17 Control Room Envelope Habitability Program............................................. 5.0-21 5.5.18 Surveillance Frequency Control Program..................................................... 5.0-22

5.6 Reporting Requirements................................................................................... 5.0-23

5.6.1 Not Used..................................................................................................... 5.0-23 5.6.2 Annual Radiological Environmental Operating Report................................ 5.0-23 5.6.3 Radioactive Effluent Release Report.......................................................... 5.0-23 5.6.4 Not used...................................................................................................... 5.0-23

CALLAWAY PLANT 5.0-i CHAPTER TABLE OF CONTENTS (Continued)

Section Page

5.6.5 CORE OPERATING LIMITS REPORT (COLR).......................................... 5.0-24 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)......................................................................................... 5.0-26 5.6.7 Not used........................................................................................................ 5.0-26 5.6.8 PAM Report................................................................................................. 5.0-26 5.6.9 Not used...................................................................................................... 5.0-26 5.6.10 Steam Generator Tube Inspection Report.................................................. 5.0-27

5.7 High Radiation Area.......................................................................................... 5.0-28

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem /hour at 30 Centimeters from the Radiation Source or from any Surface Pen etrated by the Radiation:.................................................................................................... 5.0-28 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/h our at 30 Centimeters from the Radiation Source or from any Surface Pen etrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiat ion Source or from any Surface Penetrated by the Radiation:.......................................... 5.0-29

CALLAWAY PLANT 5.0-ii Responsibility 5.1

5.0 ADMINISTRATIVE CONTROLS

5.1 Responsibility

5.1.1 The plant manager shall be responsible for overall unit op eration and shall delegate in writing the succession to this responsibility during his abs ence.

The plant manager or his designee shall approve, prior to imple mentation, each proposed test, experiment or modification to systems or equipme nt that affect nuclear safety and are not addressed in the Final Safety Analys is Report (FSAR) or Technical Specifications.

5.1.2 The Shift Manager (SM) shall be responsible for the contro l room command function. During any absence of the SM from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor O perator (SRO) license shall be designated to assume the control room command function. During any absence of the SM from the control room while the unit is i n MODE 5 or 6, an individual with an active SRO license or Reactor Operator licen se shall be designated to assume the control room command function.

CALLAWAY PLANT 5.0-1 Amendment 173 Organization 5.2

5.0 ADMINISTRATIVE CONTROLS

5.2 Organization

5.2.1 Onsite and Offsite Organizations

Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite org anizations shall include the positions for activities affecting safety of the nu clear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent form s of documentation. These requirements including the plant-specific titles of those personnel fulfilling the re sponsibilities of the positions delineated in these Technical Specifications shall be documented in the FSAR;
b. The plant manager shall be responsible for overall safe opera tion of the plant and shall have control over those onsite activities neces sary for safe operation and maintenance of the plant;
c. A specified corporate officer shall have corporate responsibi lity for overall plant nuclear safety and shall take any measures needed to ensu re acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and
d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropria te onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff

The unit staff organization shall include the following:

a. An equipment operator shall be assigned when fuel is in the r eactor and an additional equipment operator shall be assigned when the unit i s in MODE 1, 2, 3, or 4.
b. Shift crew composition may be one less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.f for a period of tim e not to

(continued)

CALLAWAY PLANT 5.0-2 Amendment 155 Organization 5.2

5.2 Organization

5.2.2 Unit Staff (continued)

exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restor e the shift crew composition to within the minimum requirements.

c. A Radiation Protection Department technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hou rs, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
d. Not Used
e. The operations manager or assistant operations manager shall hold an SRO license.
f. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineer ing, and plant analysis with regard to the safe operation of the unit. This p osition shall be assigned when the unit is in MODE 1, 2, 3, or 4, unless the Shift Manager or the Operating Supervisor meet the qualifications as required by the NRC.

CALLAWAY PLANT 5.0-3 Amendment 193 Unit Staff Qualifications 5.3

5.0 ADMINISTRATIVE CONTROLS

5.3 Unit Staff Qualifications

5.3.1 Each member of the unit staff shall meet or exceed the min imum qualifications for the comparable position(s) addressed in the standard(s) tha t is referenced in the Callaway Plant Operating Qua lity Assurance Manual (OQAM), with exceptions specified in the OQAM.

5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Op erator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in ad dition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50.54(m).

CALLAWAY PLANT 5.0-4 Amendment 225 Procedures 5.4

5.0 ADMINISTRATIVE CONTROLS

5.4 Procedures

5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33,

Revision 2, Appendix A, February 1978;

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;
c. Quality assurance for effluent and environmental monitoring;
d. Not Used; and
e. All programs specified in Specification 5.5.

CALLAWAY PLANT 5.0-5 Amendment 206 Programs and Manuals 5.5

5.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

The following programs shall be established, implemented, and m aintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent mo nitoring alarm and trip setpoints, and in the conduct of the radiological envi ronmental monitoring program; and
b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 5.6.2 and Specification 5.6.3.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluat ions justifying the change(s), and
2. a determination that the change(s) maintain the levels of rad ioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoi nt calculations;
b. Shall become effective after the approval of the plant manage r; and
c. Shall be submitted to the NRC in the form of a complete, legi ble copy of the entire ODCM as a part of or concurrent with the Radioactive Eff luent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of th e page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

(continued)

CALLAWAY PLANT 5.0-6 Amendment No. 155 Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.2 Primary Coolant Sources Outside Containment

This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioacti ve fluids during a serious transient or accident to levels as low as practicable. The systems include the recirculation portion of the Containment Spray, Safety Inje ction, Chemical and Volume Control, and Residual Heat Removal. The program sha ll include the following:

a. Preventive maintenance and periodic visual inspection requirements; and
b. Integrated leak test requirements for each system at refuelin g cycle intervals or less.

5.5.3 Not Used

5.5.4 Radioactive Effluent Controls Program

This program conforms to 10 CFR 50.36a for the control of radioac tive effluents and for maintaining the doses to members of the public from rad ioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program sh all include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set point determination in accordance with the methodology in the ODCM;
b. Limitations on the concentrations of radioactive material rel eased in liquid effluents to unrestricted areas, conforming to 10 times the conc entration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402;

(continued)

CALLAWAY PLANT 5.0-7 Amendment No. 144 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.4 Radioactive Effluent Controls Program(continued)

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodol ogy and parameters in the ODCM;
d. Limitations on the annual and quarterly doses or dose commitm ent to a member of the public from radioactive materials in liquid efflu ents released to unrestricted areas, conforming to 10 CFR 50, Appendix I;
e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and curr ent calendar year in accordance with the methodology and parameters in the O DCM at least every 31 days;
f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the p rojected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;
g. Limitations on the dose rate resulting from radioactive mater ial released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:
1. For noble gases: A dose rate of 500 mrem/yr to the whole body and a dose rate of 3000 mrem/yr to the skin, and
2. For Iodine-131, Iodine-133, tritium, and for all radionuclide s in particulate form with half-lives greater than 8 days: A dose rat e of 1500 mrem/yr to any organ.
h. Limitations on the annual and quarterly air doses resulting f rom noble gases released in gaseous effluents from each unit to areas bey ond the site boundary, conforming to 10 CFR 50, Appendix I;
i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives > 8 days in gaseous effluents released to ar eas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

(continued)

CALLAWAY PLANT 5.0-8 Amendment No. 133 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.4 Radioactive Effluent Controls Program (continued)

j. Limitations on the annual dose or dose commitment to any memb er of the public, beyond the site boundary, due to releases of radioactiv ity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 1 90;
k. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

5.5.5 Component Cyclic or Transient Limit

This program provides controls to track the FSAR, Section 3.9(N).1.1, Design Transients, cyclic and transient occurrences to ensure that co mponents are maintained within the design limits.

5.5.6 Containment Tendon Surveillance Program

This program provides controls for monitoring any tendon degrad ation, including effectiveness of its corrosion protection medium, to ensure con tainment structural integrity. The program shall include baseline measurements pri or to initial operations. The Tendon Surveillance Program, inspection freque ncies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an exemption or relief has been aut horized by the NRC.

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.7 Reactor Coolant Pump Flywheel Inspection Program

This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle on e-half of the outer radius or a surface examination (MT and/or PT) of exposed surfa ces of the removed flywheels may be conducted at 20 year intervals.

(continued)

CALLAWAY PLANT 5.0-9 Amendment No. 163 Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.8 Not Used

5.5.9 Steam Generator (SG) Program

A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Genera tor Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of t he tubing (continued)

CALLAWAY PLANT 5.0-10 Amendment No. 222 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued)

with respect to the performance criteria for structural integri t y a n d a c c i d e n t induced leakage. The "as found" condition refers to the condit ion of the tubing during a SG inspection outage, as determined from the in service inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm t hat the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structu ral integrity, accident induced leakage, and operational LEAKAGE.
1. Structural integrity performance criterion: All in-service st eam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown), all anticipated transi ents included in the design specification, and design basis accident s.

This includes retaining a safety factor of 3.0 (3DP) against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 agai nst burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary t o secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 1 gpm total for all four steam generat ors.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

(continued)

CALLAWAY PLANT 5.0-11 Amendment No. 215 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued)

c. Provisions for SG tube plugging criteria. Tubes found by ins ervice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubes heet weld at the tube inlet to the tube-to-tubesheet weld at the tube out let, and that may satisfy the applicable tube plugging criteria. The tube-to -tubesheet weld is not part of the tube. In addition to meeting the requi rements of d.1, d.2, and d.3 below, the inspection scope, inspection metho ds, and inspection intervals shall be such as to ensure that SG tube in tegrity is maintained until the next SG inspection. A degradation assessm ent shall be performed to determine the type and location of flaws to whi ch the tubes may be susceptible and, based on this assessment, to dete rmine which inspection methods need to be employed and at what locati ons.
1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2. After the first refueling outage following SG installation, i nspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections).* In addition, the minimum number of tubes inspec ted at each scheduled inspection shall be the number of tubes in al l SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging crit eria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection per iod may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of th e inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection per iod
  • As approved by Amendment No. 223, performance of the steam generator inspection scheduled for Refuel Outage 24 (fall 2020) may be deferred to Refuel Outage 25 (spring 2022) on a one-time basis.

(continued)

CALLAWAY PLANT 5.0-12 Amendment No. 223 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued)

after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection out age in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

(a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period;

(b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period;

(c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and

(d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for th e degradation mechanism that caused the crack indication shall no t exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnos tic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then t he indication need not be treated as a crack.
e. Provisions for monitoring operational primary to secondary LEAKAGE.

(continued)

CALLAWAY PLANT 5.0-13 Amendment No. 223 Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.10 Secondary Water Chemistry Program

This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variab les and control points for these variables;
b. Identification of the procedures used to measure the values o f the critical variables;
c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;
d. Procedures for the recording and management of data;
e. Procedures defining corrective actions for all off control po int chemistry conditions; and
f. A procedure identifying the authority responsible for the int erpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.11 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at t he frequencies specified in Regulatory Guide 1.52, Rev. 2, and uses the test procedure guidance in Regulatory Guide 1.52, Revision 2, Positions C.5.a, C.5.c and C.5.d.

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1.0% when tested at the system flowrate specified below.

ESF Ventilation System Flowrate

Control Room Filtration 2000 cfm, +/- 200 cfm Control Room Pressurization 500 cfm, +500, -50 cfm Emergency Exhaust System 9000 cfm, +/- 900 cfm

(continued)

CALLAWAY PLANT 5.0-14 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.11 Ventilation Filter Testing Program (VFTP) (continued)

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0% w hen tested at the system flowrate specified below.

ESF Ventilation System Flowrate

Control Room Filtration 2000 cfm, +/- 200 cfm Control Room Pressurization 500 cfm, +500, -50 cfm Emergency Exhaust System 9000 cfm, +/- 900 cfm

c. Demonstrate for each of the ESF systems within 31 days after r emoval that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the va lue specified below when tested in accordance with ASTM D3803-1989 a t a temperature of 30°C and the relative humidity specified below.

ESF Ventilation System Penetration RH

Control Room Filtration 2.0% 70%

Control Room Pressurization 2.0% 70%

Emergency Exhaust System 2.0% 70%

d. Demonstrate at least once per 18 months for each of the ESF sy stems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below.

ESF Ventilation System Delta P Flowrate

Control Room Filtration 5.4" WG 2000 cfm,

+/- 200 cfm Control Room Pressurization 5.4" WG 500 cfm,

+500,- 50 cfm Emergency Exhaust System 5.4" WG 9000 cfm,

+/- 900 cfm

(continued)

CALLAWAY PLANT 5.0-15 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.11 Ventilation Filter Testing Program (VFTP) (continued)

e. Demonstrate at least once per 18 months that the heaters for e ach of the ESF systems dissipate the value specified below when tested in accordance with ANSI 510-1975 and corrected to design nameplate voltage settings.

ESF Ventilation System Wattage

Control Room Pressurization 5 +/- 1 KW Emergency Exhaust System 37 +/- 3 KW

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFT P test frequencies.

5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring P rogram

This program provides controls for potentially explosive gas mixtures contained in the Gaseous Radwaste System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ET SB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure, Revision 0". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures, Revision 2".

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the G aseous Radwaste System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withsta nd a hydrogen explosion);
b. A surveillance program to ensure that the quantity of radioac tivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of t he tanks' contents; and
c. A surveillance program to ensure that the quantity of radioac tivity contained in the outdoor liquid radwaste tanks listed below tha t are not

(continued)

CALLAWAY PLANT 5.0-16 Amendment No. 219 l Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring P rogram (continued)

surrounded by liners, dikes, or walls, capable of holding the t anks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste System is less than the quantities determined in accordance with the Standard Review Plan, Section 15.7.3:

a. Reactor Makeup Water Storage Tank,
b. Refueling Water Storage Tank,
c. Condensate Storage Tank, and
d. Outside temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.13 Diesel Fuel Oil Testing Program

A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall in clude sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to st orage tanks by determining that the fuel oil has:
1. an API gravity or an absolute specific gravity within limits,
2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
3. a water and sediment content within limits for ASTM 2D fuel oi l.
b. Other properties for ASTM 2D fuel oil are analyzed within 31 da ys following sampling and addition of new fuel oil to storage tank s; and
c. Total particulate concentration of the stored fuel oil is 10 mg/l when tested every 31 days based on applicable ASTM D-2276 standards.
d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

(continued)

CALLAWAY PLANT 5.0-17 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.14 Technical Specifications (TS) Bases Control Program

This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriat e administrative controls and reviews.
b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
1. a change in the TS incorporated in the license; or
2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
d. Proposed changes that meet the criteria of Specification 5.5. 14b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e ).

5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriat e actions may be taken as a result of the support system inoperability and corre sponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis do es not go undetected;
b. Provisions for ensuring the plant is maintained in a safe con dition if a loss of function condition exists;

(continued)

CALLAWAY PLANT 5.0-18 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.15 Safety Function Determination Program (SFDP) (continued)

c. Provisions to ensure that an inoperable supported system's Co mpletion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
d. Other appropriate limitations and remedial or compensatory ac tions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be perf ormed. For the purpose of this program, a loss of safety function may exist wh en a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
b. A required system redundant to the system(s) in turn supporte d by the inoperable supported system is also inoperable; or
c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriat e Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exceptions:
1. The visual examination of containment concrete surfaces inten ded to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

(continued)

CALLAWAY PLANT 5.0-19 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.16 Containment Leakage Rate Testing Program (continued)

2. The visual examination of the st eel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.
3. The unit is excepted from post-modification integrated leakag e rate testing requirements associated with steam generator replacement during the Refuel 14 outage (fall of 2005).
4. The first Type A test performed after the October 26, 1999 Ty pe A test shall be performed no later than October 25, 2014.
b. The peak calculated containment internal pressure for the des ign basis loss of coolant accident, Pa, is 48.1 psig.
c. The maximum allowable containment leakage rate, L a, at Pa, shall be 0.20% of the containment air weight per day.
d. Leakage rate acceptance criteria are:
1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L a for the Type B and C tests and 0.75 La for Type A tests;
2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is 0.05 La when tested at Pa;

b) For each door, leakage rate is 0.005 La when pressurized to 10 psig.

e. The provisions of Technical Specification SR 3.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Program.
f. The provisions of Technical Specification SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

(continued)

CALLAWAY PLANT 5.0-20 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.17 Control Room Envelope Habitability Program

A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal condition s and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radi ation protection is provided to permit access and occupancy of the CRE under des ign basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the follow ing elements:

a. The definition of the CRE, CRE boundary, control building env elope (CBE), and the CBE Boundary.
b. Requirements for maintaining the CRE and CBE boundaries in th eir design condition, including configuration control and preventiv e maintenance.
c. Requirements for (i) determining the unfiltered air inleakage past the CRE and CBE boundaries in accordance with the testing methods and a t the Frequencies specified in Sections C.1 and C.2 of Regulatory Gui de 1.197, "Demonstrating Control Room Envel ope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitab ility at the Frequencies specified in Sections C.1 and C.2 of Regulatory Gui de 1.197, Revision 0.

The following exception is taken to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1. The Tracer Gas Test based on the Brookhaven National Laboratory Atmospheric Tracer Depletion (ATD) Method is used to determine the unfiltered air inleakage past the CRE and CBE boundaries. The ATD Method is described in AmerenUE letters dated December 15, 2004 (ULNRC-05104), June 6, 2006 (ULNRC-05298), July 16, 2007 (ULNRC-05427), and October 30, 2007 (ULNRC-05448).
d. Measurement, at designated locations, of the CRE pressure rel ative to the outside atmosphere during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

(continued)

CALLAWAY PLANT 5.0-21 Amendment No. 215 l Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.17 Control Room Envelope Habitability Program (continued)

e. The quantitative limits on unfiltered air inleakage into CRE and CBE.

These limits shall be stated in a manner to allow direct compar ison to the unfiltered air inleakage measured by the testing described in paragraph c.

The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air leakage limits for hazardous chem icals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE and CBE unfiltered inleakage, and measuring CRE pressure and assessing CRE and CBE as required by paragraphs c and d, respectively.

5.5.18 Surveillance Frequency Control Program

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technica l Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Contr ol Program shall contain a list of Frequencies of those Surveillance Requirements for which the Fr equency is controlled by the program.
b. Changes to the Frequencies listed in the Surveillance Frequen cy Control Program shall be made in accordance with NEI 04-10, Risk-Infor med Method for Control of Surveillance Frequencies, Revision 1.
c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance F requency Control Program.

CALLAWAY PLANT 5.0-22 Amendment No. 215 l Reporting Requirements 5.6

5.0 ADMINISTRATIVE CONTROLS

5.6 Reporting Requirements

The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Not Used.

5.6.2 Annual Radiological Environmental Operating Report

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitte d by May 1 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring pro gram for the reporting period.

The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendi x I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pu rsuant to the locations specified in the table and figures in the ODCM, as we ll as summarized and tabulated results of these analyses and measurements in a f ormat similar to the table in the Radiological Assessment Branch Technical Posit ion, Revision 1, November 1979. In the event that some individual results are no t available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report

The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of ea ch year in accordance with 10 CFR 50.36a. The report shall include a summar y of the quantities of radioactive liquid and gaseous effluents and soli d waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conform ance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

5.6.4 Not used.

(continued)

CALLAWAY PLANT 5.0-23 Amendment No. 215 l Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each relo ad cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1. Moderator Temperature Coefficient limits in Specification 3.1.3,
2. Shutdown Bank Insertion Limit for Specification 3.1.5,
3. Control Bank Insertion Limits for Specification 3.1.6,
4. Axial Flux Difference Limits for Specification 3.2.3,
5. Heat Flux Hot Channel Factor, FQ(Z), FQRTP, K(Z), W(Z) and FQ Penalty Factors for Specification 3.2.1,
6. Nuclear Enthalpy Rise Hot Channel Factor, F H RTPF N H, and Power Factor Multiplier, PF H, limits for Specification 3.2.2,
7. Shutdown Margin Limits for Specifications 3.1.1, 3.1.4, 3.1.5, 3.1.6, and 3.1.8,
8. Reactor Core Safety Limits Figure for Specification 2.1.1,
9. Overtemperature T and Overpower T Setpoint Parameters for Specification 3.3.1, and
10. Reactor Coolant System Pressure and Temperature DNB Limits for Specification 3.4.1.
b. The analytical methods used to determine the core operating l imits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY.
2. WCAP-10216-P-A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL AND FQ SURVEILLANCE TECHNICAL SPECIFICATION.
3. WCAP-10266-P-A, THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE.

(continued)

CALLAWAY PLANT 5.0-24 Amendment No. 215 l Reporting Requirements 5.6

5.6 Reporting Requirements

4. WCAP-12610-P-A, VANTAGE + FUEL ASSEMBLY REFERENCE CORE REPORT.
5. WCAP-11397-P-A, REVISED THERMAL DESIGN PROCEDURE.
6. WCAP-14565-P-A, VIPRE-01 MODELING AND QUALIFICATION FOR PRESSURIZED WATER REACTOR NON-LOCA THERMAL-HYDRAULIC SAFETY ANALYSIS.
7. WCAP-10851-P-A, IMPROVED FUEL PERFORMANCE MODELS FOR WESTINGHOUSE FUEL ROD DESIGN AND SAFETY EVALUATIONS.
8. WCAP-15063-P-A, WESTINGHOUSE IMPROVED PERFORMANCE ANALYSIS AND DESIGN MODEL (PAD 4.0).
9. WCAP-8745-P-A, DESIGN BASES FOR THE THERMAL OVERPOWER DT AND THERMAL OVERTEMPERATURE DT TRIP FUNCTIONS.
10. WCAP-10965-P-A, ANC: A WESTINGHOUSE ADVANCED NODAL COMPUTER CODE.
11. WCAP-10965-P-A Addendum 2-A, Qualification of the New Pin Power Recovery Methodolgy.
12. WCAP-13524-P-A, APOLLO: A ONE DIMENSIONAL NEUTRON DIFFUSION THEORY PROGRAM.
13. WCAP-14565-P-A Addendum 2-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications.
14. WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON.
15. WCAP-16045-P-A Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology.
c. The core operating limits shall be determined such that all a pplicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits suc h as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

(continued)

CALLAWAY PLANT 5.0-25 Amendment No. 217 Reporting Requirements 5.6

5.6 Reporting Requirements

d. The COLR, including any midcycle revisions or supplements, sh all be provided upon issuance for each reload cycle to the NRC.

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing and PORV lift setting as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
1. Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits, and
2. Specification 3.4.12, Cold Overpressure Mitigation System (COMS).
b. The analytical methods used to determine the RCS pressure and temperature and COMS PORV limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressur e Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves".
c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement theret o.

5.6.7 Not used.

5.6.8 PAM Report

When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alte rnate method of monitoring, the cause of the inoperability, and the plans and s chedule for restoring the instrumentation channels of the Function to OPERA BLE status.

5.6.9 Not used.

(continued)

CALLAWAY PLANT 5.0-26 Amendment No. 217 l Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.10 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial e ntry into MODE 4 following completion of an inspection performed in accordance w ith Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG;
b. Degradation mechanisms found;
c. Nondestructive examination techniques utilized for each degra dation mechanism;
d. Location, orientation (if linear), and measured sizes (if available) of service induced indications;
e. Number of tubes plugged during the inspection outage for each degradation mechanism;
f. The number and percentage of tubes plugged to date, and the e ffective plugging percentage in each steam generator; and
g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

CALLAWAY PLANT 5.0-27 Amendment No. 215 High Radiation Area 5.7

5.0 ADMINISTRATIVE CONTROLS

5.7 High Radiation Area

As provided in paragraph 20.1601(c) of 10 CFR Part 20, the follow ing controls shall be applied to high radiation areas in place of the controls required by pa ragraph 20.1601 (a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem /hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation:

a. Each entryway to such an area shall be barricaded and conspic uously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment;
b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiati on protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously displays radi ation dose rates in the area; or
2. A radiation monitoring device that continuously integrates th e radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or
3. A radiation monitoring device that continuously transmits doe s rate and cumulative dose rate information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or
4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and

(continued)

CALLAWAY PLANT 5.0-28 Amendment No. 215 l High Radiation Area 5.7

5.7 High Radiation Area

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/ hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)

(i) Be under the surveillance, as specified in the RWP or equiva lent, while in the area, of an individual qualified in radiation prot ection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or

(ii) Be under the surveillance as specified in the RWP or equiva lent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and w ith the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection proc edures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/h our at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from a ny Surface Penetrated by the Radiation:

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuou sly guarded door or gate that prevents unauthorized entry, and, in addition:
1. All such door and gate keys shall be maintained under the administrative control of the Shift Manager/Operating Superviso r or Radiation Protection Department Supervision, or his or her designee.
2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiatio n dose rates in the immediate work area(s) and other appropriate radiation p rotection equipment and measures.

(Continued)

CALLAWAY PLANT 5.0-29 Amendment No. 215 l High Radiation Area 5.7

5.7 High Radiation Area

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/h our at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from a ny Surface Penetrated by the Radiation: (continued)

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise followin g plant radiation protection procedures for entry to, exit from, and wo rk in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously integrates th e radiation rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or
2. A radiation monitoring device that continuously transmits dos e rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or
3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and

(i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or

(ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area, or

4. In those cases where options (2) and (3), above, are impracti cal or determined to be inconsistent with the As Low As is Reasonably

(Continued)

CALLAWAY PLANT 5.0-30 Amendment No. 215 l High Radiation Area 5.7

5.7 High Radiation Area

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/h our at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from a ny Surface Penetrated by the Radiation: (continued)

Achievable principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

e. Except for individual qualified in radiation protection proce dures or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.
f. Such individual areas that are within a larger area, such as PWR containment, where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the ind ividual area need not be controlled by a locked door or gate nor contin uously guarded, but shall be barricaded, conspicuously posted, and a c learly visible flashing light shall be activated at the area as a warn ing device.

CALLAWAY PLANT 5.0-31 Amendment No. 215 l

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