Text

Informational Copy of Operating License & Appendices. Technical Specification Pages 5.0-1 Through 5.0-34 for Amendment 176
	ML040780507
Person / Time
Site:	Robinson
Issue date:	02/18/2004
From:	; Progress Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML040780507 (34)
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Responsibility 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

The Plant Manager or his designee shall approve, prior to implementation, each proposed test, experiment and modification to systems or equipment that affect nuclear safety.

5.12 The Superintendent-Shift Operations (SSO) shall be responsible for the control room command function. During any absence of the SSO from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.

During any absence of the SSO from the control room while the unit is in MODE5 or 6, an individual with an active SR0 license or Reactor Operator license shall be designated to assume the control room command function.

HBRSEP Unit No. 2 5.0-l Amendment No. 176

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 organizations shall include the positions for activities affecting safety of the nuclear 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. These lines of authority, responsibility, and communication shall be documented in the UFSAR:

b. The Plant Manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant:

c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure 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 radiation control, or perform quality assurance functions may report to the appropriate 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 auxiliary operator shall be assigned to the shift crew when fuel is in the reactor . An additional auxiliary (continued)

HBRSEP Unit No. 2 5.0-2 Amendment No. 176

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) operator shall be assigned to the shift crew while the unit is operating in MODES1, 2, 3, or 4.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

C. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specification 5.2.2.a and 5.2.2.9 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d. An individual qualified as a radiation control technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiation control personnel, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time; (continued)

HBRSEP Unit No. 2 5.0-3 Amendment No. 176

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized in advance by the Plant Manager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

f. The Operations Manager or Superintendent in charge of the operations shift crews shall hold an SR0 license.

9. During MODES1, 2, 3, and 4, the shift technical advisor (STA) shall provide advisory technical support to the SSO with regard to the safe operation of the unit. If an individual that holds an SR0 license also meets the STA requirements, that individual may act in both capacities.

HBRSEP Unit No. 2 5.0-4 Amendment No. 176

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 minimum qualifications of ANSI N18.L1971 for comparable positions, except for the manager of the radiation control function, who shall meet or exceed the minimum qualifications of ANSWANS 3.1-1981, and the STA, who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

HBRSEP Unit No. 2 5.0-5 Amendment No. 176

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 commitments to NUREG-0737 and of NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; C. Quality assurance for effluent and environmental monitoring:

d. Fire Protection Program implementation; and

e. All programs specified in Specification 5.5.

HBRSEP Unit No. 2 5.0-6 Amendment No. 176

Programs and Manuals 5.5.

5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs and manuals shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCMshall 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 monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program: and

b. The ODCMshall 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.

C. Licensee initiated changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

(a) sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b) a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations:

2. Shall become effective after the approval of the Plant Manager: and

3. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCMas a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCMwas made. Each change shall be identified by (continued)

HBRSEP Unit No. 2 5.0-7 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 te Dose Calculation Manual (ODCM) (continued) markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Primary Coolant Sources Outside Containment This program provides controls to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include Residual Heat Removal, Safety Injection, Containment Spray, Post Accident Containment Ventilation; and portions of Chemical and Volume Control, Liquid Waste Disposal, Gaseous Waste Disposal, and Sampling. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

5.5.3 Deleted 5.5.4 ve Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of (continued)

HBRSEP Unit No. 2 5.0-8 Amendment No. 192

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued) the public from radioactive 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 shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2401; C. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents 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 current calendar year in accordance with the methodology and parameters in the ODCMat 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 projected dose commitments due to the release of effluents to unrestricted areas exceed specified limits conforming to 10 CFR 50, Appendix I:

G. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary as follows:

1. For noble gases: < 500 mrem/yr to the whole body, < 3000 mrem/yr to the skin; and (continued)

HBRSEP Unit No. 2 5.0-9 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

2. For I-131, I-133, tritium, and all radionuclides in particulate form (inhalation pathway only) with half lives > 8 days: < 1500 mrem/yr to any organ.;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond 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 areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the UFSAR, Table 3.9.1-1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitorins any tendon degradation in pre-stressed concrete containments in ~cluding effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include inspection frequencies, and acceptance criteria.

The provisions of SR 3.0.2 and SR 3.0.3 are applicabl e to the Tendon Surveillance Program inspection frequencies.

(continued)

HBRSEP Unit No. 2 5.0-10 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program provides controls for the inspection of each reactor coolant pump flywheel in accordance with the Inservice Inspection Program.

5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities:

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

(continued)

HBRSEP Unit No. 2 5.0-11 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Tube Surveillance Proqram This program provides controls for the inservice inspection of SG tubes to assure the continued integrity of the Reactor Coolant System pressure boundary and shall include the following:

a. Tube Inspection Entry from either the hot-leg side or cold-leg side with examination encompassing the area from the hot-leg tube end completely around the U-bend to the top support of the cold leg is considered a tube inspection.

b. Sample Selection and Testing Selection and testing of steam generator tubes shall be made on the following basis:

1. One steam generator shall be inspected during inservice inspection in accordance with the following requirements:

(a) The inservice inspection may be limited to one steam generator on a rotating sequence basis.

This examination shall include at least 9% of the tubes if the results of the first or a prior inspection indicate that all three generators are performing in a comparable manner.

(b) When other steam generators are required to be examined by Table 5.5-l and if the condition of the tubes in one or more generators is found to be more severe than in the other steam generators, the steam generator sampling sequence at the subsequent inservice inspection shall be modified to examine the steam generator or generators with the more severe condition.

2. The minimum sample size, inspection result classification and the associated required action shall be in conformance with the requirements specified in Table 5.5-l. The results of each sampling examination of a steam generator shall be classified into the following three categories:

(continued)

HBRSEP Unit No. 2 5.0-12 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

Category C-l: less than 5% of the total number of tubes examined are degraded but none are defective.

Category C-2: Between 5% and 10% of the total number of tubes examined are degraded, but none are defective or one tube to not more than 1% of the sample is defective.

Category C-3: More than 10% or the total number of tubes examined are degraded, but none are defective or more than 1% of the sample is defective.

In the first sample of a given steam generator during any inservice inspection, degraded tubes not beyond the plugging limit detected by the prior examinations in that steam generator shall be included in the above percentage calculations, only if these tubes are demonstrated to have a further wall penetration of greater than 10% of the nominal tube wall thickness.

3. Tubes shall be selected for examination primarily from those areas of the tube bundle where service experience has shown the most severe tube degradation.

4. The tubes examined in a given steam generator during the first examination of any inservice inspection shall include all non-plugged tubes in that steam generator that from prior examination were degraded, plus additional tubes are required to satisfy the minimum sample size specified in Table 5.5-l. If any selected tube does not permit passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection. This information shall be included in the report required by Specification 5.6.8.b.

5. During the second and third sample examinations of any inservice inspection, the tube inspection may be limited to those sections of the tube lengths where (continued)

HBRSEP Unit No. 2 5.0-13 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued) imperfections were detected during the prior examination.

6. During subsequent inservice inspections, the tube inspection may be limited to certain areas of the tube sheet array and those sections of the tube lengths where imperfections were detected during previous inservice inspections.

C. Examination Method and Requirements Steam generator tubes shall be examined in accordance with the method prescribed in Appendix IV, "Eddy Current Examination of Non-Ferromagnetic Steam Generator Heat Exchanger Tubes," as contained in ASME Boiler and Pressure Vessel Code,Section XI, "Inservice Inspection of Nuclear Power Plant Components."

d. Inspection Intervals

1. Inservice inspections shall not be more than 24 calendar months apart, except that reduced or tightened inspection intervals shall be governed as specified in 5.5.9.d.3 and d.4.

2. The inservice inspections may be scheduled to be coincident with refueling outages or any plant shutdown, provided the inspection intervals of 5.5.9.d.1, d.3, or d.4, as applicable, are not exceeded.

3. If two consecutive inservice inspections covering a time span of at least 12 months yield results that fall in C-l category, the inspection frequency may be extended to 40 month intervals between inspections.

4. If the results of the inservice inspection of steam generator tubing conducted in accordance with Table 5.5-l at 40 month intervals fall in category C-3, the inspection frequency shall be reduced to at least once per 20 months. The increase in inspection frequency shall apply until a subsequent inspection meets the (continued)

HBRSEP Unit No. 2 5.0-14 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued) conditions specified in 5.5.9.d.3 and the interval can be extended to a 40 month period.

5. Unscheduled inspections shall be conducted in accordance with Specification 5.5.9.b on any steam generator with primary-to-secondary tube leakage (not including leaks originating from tube-to-tube sheet welds) exceeding Specification 3.4.13.

All steam generators shall be inspected before returning to power in the event of a seismic occurrence greater than an operating basis earthquake, a LOCA requiring actuation of engineering safeguards, or a main steam line or feedwater line break.

e. Acceptance Limits Definitions:

Imperfection is an exception to the dimension, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

Degradation means a service induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube.

Degraded Tube is a tube that contains imperfections caused by degradation equal to or greater than 20% of the nominal tube wall thickness.

Defect is an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

Plugging Limit is the imperfection depth beyond which a degraded tube must be removed from service by plugging, because the tube may become defective prior to the next scheduled inspection of that tube. The plugging limit is 47% of the nominal tube wall thickness if the next inspection interval of that tube is 12 months, and a 2%

(continued)

HBRSEP Unit No. 2 5.0-15 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued) reduction in the plugging limit for each 12 month period until the next inspection of the inspected steam generator.

f. Corrective Measures All tubes that leak or are determined to have degradation exceeding the plugging limit shall be plugged prior to return to power.

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 critical parameters, their sampling frequency, sampling points, and control band limits;

b. Procedures used to measure the critical parameters:

C. Requirements for the documentation and review of sample results:

d. Procedures which identify the administrative events and corrective actions required to return the secondary chemistry to its normal control band following an out of control band condition; and

e. Identification of the authority responsible for the interpretation of the sample results.

(continued)

HBRSEP Unit No. 2 5.0-16 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP)

This program provides controls for implementation of the following required testing of Engineered Safety Feature (ESF) ventilation filter systems at the frequencies specified in Positions C.5 and C.6 of Regulatory Guide 1.52, Revision 2, March 1978, and conducted in general conformance with ANSI N510-1975 or N510-1980.

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows the specified penetration and system bypass leakage when tested in accordance with the referenced standard at the system flowrate specified below.

ESF Ventilation Penetration System /Bypass Flowrate Reference Std Control Room <0.05% 3300 - 4150 ACFM Regulatory Guide Emergency 1.52, Revision 2, March 1978, C.5.a, C.5.c C.5.d (using ANSI N510-1980)

Spent Fuel <1% 11070- 13530 CFM ANSI N510-1975 Building Containment <1% 31500- 38500 CFM ANSI N510-1975 Purge (continued)

HBRSEP Unit No. 2 5.0-17 Amendment No. 176

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 the specified penetration and system bypass leakage when tested in accordance with the referenced standard at the system flowrate specified below.

ESF Ventilation Penetration System /Bypass Flowrate Reference Std Control <0.05% 3300 - 4150 ACFM Regulatory Guide Room 1.52, Revision 2, Emergency March 1978, C.5.a, C.5.c C.5.d (using ANSI N510-1980)

Spent Fuel -<l% 11070- 13530 CFM ANSI N510-1975 Building Containment -<l% 31500- 38500 CFM ANSI N510-1975 Purge (continued)

HBRSEP Unit No. 2 5.0-18 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (continued)

C. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°) and the relative humidity specified below.

FSF Filter System Penetration RH Control Room <=2.5% 70%

Emergency Spent Fuel <=10% 70%

Building Containment Purge <=10% 95%

(continued)

HBRSEP Unit No. 2 5.0-19 Amendment No. 189

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)

d. Demonstrate for each of the ESF systems 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 Filter System Delta P Flowrate Control Room 13.4 inches 3300 - 4150 ACFM Emergency water gauge Spent Fuel <6 inches 12300 CFM +lO%

Building water gauge Containment <6 inches 35000 CFM +lO%

Purge water gauge

e. Demonstrate that the heaters for the Spent Fuel Building ventilation filter system maintains the filter inlet air at

< 70% relative humidity when tested in accordance with ASME N510-1975.

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

5.5.12 Explosive Gas and Storage Tank Radioactivity Monitorins Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Decay Tanks, the quantity of radioactivity contained in The Waste Gas Decay Tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Decay Tanks and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate (continued)

HBRSEP Unit No. 2 5.0-20 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitorins Program (continued) to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A surveillance program to ensure that the quantity of radioactivity contained in each Waste Gas Decay 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 the tanks' contents:

and C. A surveillance program to ensure that the quantity of radioactivity contained in each outdoor liquid radwaste tank that is not surrounded by liners, dikes, or walls, capable of holding the tank's contents and that does not have tank overflows and surrounding area drains connected to the Liquid Waste Disposal System is less than or equal to ten (10) Curies, excluding tritium and dissolved or entrained noble gases.

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 shall be established requiring testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria. The testing methods shall be in accordance with applicable ASTM Standards. The acceptance criteria shall be in accordance with the diesel engine manufacturer specifications.

The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has not become contaminated with other products during transit, thus altering the quality of the fuel oil.

(continued)

HBRSEP Unit No. 2 5.0-21 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program (continued)

b. Acceptability of fuel oil for use by testing the following parameters at a 31 day frequency:

API or specific gravity, viscosity, water and sediment, and cloud point.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program surveillance frequencies.

5.5.14 Technical Specifications (TS) Bases Control Program This program provides controls for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve 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 involves an unreviewed safety question as defined in 10 CFR 50.59.

C. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

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 provides controls to ensure 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 appropriate actions may be (continued)

HBRSEP Unit No. 2 5.0-22 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued) taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. 'This program implements the requirements of LCO 3.0.6.

a. The SFDP shall contain the following:

1. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected:

2. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists:

3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4. Other appropriate limitations and remedial or compensatory actions.

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

1. A required system redundant to the system(s) supported by the inoperable support system is also inoperable: or

2. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable: or

3. A required system redundant to the support system(s) for the supported systems described in b.1 and b.2 above is also inoperable.

c. The SFDP identifies where a loss of safety function exists.

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

(continued)

HBRSEP Unit No. 2 5.0-23 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program This program provides controls for implementation of 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 for Type A testing. 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 exception:

a. NE1 94 1995, Section 9.2.3: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than May 9, 2004.

Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix 3, Option A.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 40.5 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criteria is I 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 La for the Type B and Type C tests, and 5 0.75 La for Type A tests.

The provisions of SR .3.0.3 are applicable to the Containment Leakage Rate Testing Program.

(continued)

HBRSEP Unit No. 2 5.0-24 Amendment No. a l-W 193

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

TABLE 5.5-l STEAM GENERATORTUBE INSPECTION 1st SAMPLE EXAMINATION 2nd SAMPLE EXAMINATION 3rd SAMPLE EXAMINATION Sample Size Result Action Required Result Action Required Result Action Required A minimum of S C-l Acceptable for N/A N/A N/A N/A tubes per Steam Continued Service Generator (SG)

C-2 Plug tubes C-l Acceptable for N/A N/A exceeding the Continued S=3(N/n)% plugging limit and Service proceed with 2nd sample examination C-2 Plug tubes C-l Acceptable for where: of 2S tubes in exceeding the Continued Service same steam plugging limit generator and proceed with C-2 Plug tubes exc.

N is the number 3rd sample plug limit.

of steam examination of Acceptable for generators in 4s tubes in same continued service the plant = 3 steam generator C-3 Perform action required under n is the number C-3 of 1st sample of steam examination generators inspected during an C-3 Perform action N/A N/A examination required under C-3 of 1st sample examination C-3 Inspect all tubes All other Acceptable for N/A N/A in this SG, plug SGs are C-l Continued tubes exceeding Service the plugging limit and proceed with 2nd sample Some SGs are Perform Action N/A N/A examination of 2S C-2 but no required under tubes in each additional C-2 of 2nd other steam SGs are sample generator not C-3 examination included in the above inservice inspection Additional Inspect all N/A N/A program. Report SG is C-3 tubes in this SG results to NRC. and plug tubes exceeding the plugging limit.

Report results to NRC HBRSEP Unit No. 2 5.0-25 Amendment No. 176

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 Occupational Radiation Exposure Report A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving an annual deep dose equivalent > 100 mrem/yr and their associated collective deep dose equivalent (reported in person-rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, maintenance, waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report shall be submitted by April 30 of each year.

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 submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program 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, Appendix I, Sections IV.B.2, IV.B.3, and 1V.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 pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of Table 3 in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

(continued)

HBRSEP Unit No. 2 5.0-26 Amendment No. 176

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operatins Report (continued)

In the event that some individual results are not 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 shall be submitted in accordance with 10 CFR 50.36a.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCMand Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.l.

5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Shutdown Margin (SDM) for Specification 3.1.1;

2. Moderator Temperature Coefficient limits for Specification 3.1.3:

3. Shutdown Bank Insertion Limits for Specification 3.1.5; (continued)

HBRSEP Unit No. 2 5.0-27 Amendment No. 176

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (continued)

4. Control Bank Insertion Limits for Specification 3.1.6;

5. Heat Flux Hot Channel Factor (FQ(Z)) limit for Specification 32.1;

6. Nuclear Enthalpy Rise Hot Channel Factor(FNDH) limit for Specification 3.22;

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and

8. Boron Concentration limit for Specification 3.9.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR.

These methods are those specifically described in the following documents:

1. XN-75-27(A), "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," approved version as specified in the COLR.

2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.

3. XN-NF-82-21(A), "Application of Exxon Nuclear Company PWRThermal Margin Methodology to Mixed Core Configurations," approved version as specified in the COLR.

4. Steam Line Break Methodology as defined by:

ANF-84-093(P)(A), Steamline Break Methodology for PWRs," approved version as specified in the COLR.

EMF-84-093(P)(A), Steam Line Break Methodology for PWRs," approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-28 Amendment No. 176, 188

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (continued)

5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow," approved version as specified in the COLR.

6. XN-NF-82-49(A), "Exxon Nuclear Corporation Evaluation Model EXEM PWRSmall Break Model," approved version as specified in the COLR.

7. EMF-2087 (P)(A), SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," approved version as specified in the COLR.

8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR.

9. XN-NF-621(A), "XNB Critical Heat Flux Correlation,"

approved version as specified in the COLR.

10. ANF-1224(A), "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel,"

approved version as specified in the COLR.

11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

12. WCAP-10080-A, "NOTRUMP,A Nodal Transient Small Break and General Network Code," approved version as specified in the COLR.

13. WCAP-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMPcode," approved version as specified in the COLR.

14. WCAP-8301 (Proprietary) and WCAP-8305 (Nonproprietary),

"LOCTA-IV Program: Loss of Coolant Transient Analysis,"

approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-29 Amendment No. 176. 188

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 F CORE OPERATING LIMITS REPORT

15. "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 87 to Facility Operating License No. DPR-23, Carolina Power & Light Co.,

H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261," USNRC, Washington, DC 20555, 7 Nov. 84.

16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B.

Robinson Unit 2," approved version as specified in the COLR.

17. ANF-88-133 (P>(A), "Qualification of Advanced Nuclear Fuels' PWRDesign Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR.

18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR.

19. EMF-92-081(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.

20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel,"

approved version as specified in the COLR.

21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs,"

approved version as specified in the COLR.

23 EMF-92-116, Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-30 Amendment No. 176,188

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (continued)

C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

5.6.6 When a report is required by Condition B or H 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 alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

(continued)

HBRSEP Unit No. 2 5.0-31 Amendment No. 176, 178 185, 188 I

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.7 Tendon Surveillance Report

a. Notification of a pending sample tendon test, along with detailed acceptance criteria, shall be submitted to the NRC at least two months prior to the actual test.

b. A report containing the sample tendon test evaluation shall be submitted to the NRC within six months of conducting the test.

5.6.8 Steam Generator Tube Inspection Report

a. A report of the number of tubes plugged in each steam generator shall be submitted to the NRC within 14 days after completion of the tube plugging.

b. A report of the results of the steam generator tube inspection shall be included in the Monthly Operating Report for the period beginning after the final inspection is completed.

Reports shall include:

Number and extent of tubes inspected

2. Location and percent of wall thickness penetration on for each eddy current indication and any leaks.

3. Identification of tubes plugged.

C. A report of examination results falling in Category C-3 of Table 5.5-l shall be submitted to the NRC within 30 days, and prior to resumption of plant operation.

A report of investigations conducted to determine cause(s) of the tube degradation and corrective measures taken to prevent recurrence shall be submitted within 90 days following completion of the startup test program.

HBRSEP Unit No. 2 5.0-32 Amendment No. 176 , 185 I

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area 5.7.1 In lieu of the control device" or alarm signal" required by paragraph 20.1601(a) of 10 CFR 20, each High Radiation Area in which the intensity of radiation is 1000 mRem/hour or less shall be barricaded and conspicuously posted by requiring issuance of a Radiation Work Permit (RWP).

Radiation control personnel or personnel escorted by radiation control personnel shall be exempt from the RWP issuance requirements during the performance of their assigned duties within the RCA, provided they comply with approved radiation protection procedures for entry into High Radiation Areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device provided for each individual that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

C. An individual qualified as a radiation control technician with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the radiation control supervisor in the RWP.

(continued)

HBRSEP Unit No. 2 5.0-33 Amendment No. 176

High Radiation Area 5.7 5.7 High Radiation Area (continued) 5.7.2 The requirements of 57.1 shall apply to each High Radiation Area in which the intensity of radiation is greater than 1000 mRem/hour at 30 centimeters (12 inches) 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 any surface penetrated by the radiation. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the SS on duty and/or the radiation control supervisor.

Entrance thereto shall also be controlled by requiring issuance of an RWP. The exemption from RWP issuance requirements discussed in 5.7.1 is not applicable for any High Radiation Area in which the intensity of radiation is greater than 1000 mRem/hour.

HBRSEP Unit No. 2 5.0-34 Amendment No. 176

ML040780507

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