ML072140130

Technical Specifications, Issuance of Amendment Steam Generator Tube Integrity TS Amendment Using the Consolidated Line Item Improvement Process

ML072140130
Person / Time
Site:	Davis Besse
Issue date:	07/31/2007
From:	NRC/NRR/ADRO/DORL/LPLIII-2
To:
Wengert, Thomas J, NRR/DORL, 415-4037
Shared Package
ML072050089	List: ML072040417 ML072140130
References
TAC MD0077, TAC MD2145
Download: ML072140130 (37)

Text

ATTACHMENT TO LICENSE AMENDMENT NO. 276 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License NPF-3 License NPF-3 Page 4 Page 4 TSs TSs INDEX, V INDEX, V INDEX, VII. INDEX, VII INDEX, IX INDEX, IX INDEX, XII INDEX,.XII INDEX, XV INDEX, XV 1-4 1-4 3/4 1-1 3/4 1-1 3/4 4-6 3/4 4-6 3/4 4-6a 3/4 4-6b 3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-9a 3/4 4-10 3/4 4-10a 3/44-11 3/4 4-12 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 7-38 3/4 7-39 6-1 6-1 6-2 6-2 6-3 6-3 6-4 6-4 6-5 6-5 6-6 6-6 6-7 6-7 6-8 6-8 6-9 6-9 6-10 6-10 6-11 6-1 1 6-12 6-12 6-13 6-13

Remove Insert TSs TSs 6-14 6-14 6-15 6-15 6-16 6-16 6-17 6-17 6-18 6-18 6-19 6-19 6-20 6-20 6-21 6-21 6-22 6-22 6-23 6-24

2.C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level FENOC is authorized to operated the facility at steady state reactor core power levels not in excess of 2772 megawatts (thermal). Prior to attaining the power level, Toledo Edison Company shall comply with the conditions identified in Paragraph (3) (o) below and complete the preoperational tests, startup tests and other items identified in Attachment 2 to this license in the sequence specified.

Attachment 2 is an integral part of this license.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission:

(a) FENNIC shall not operate the reactor in operational Modes 1 and 2 with less than three reactor coolant pumps in operation.

(b) Deleted per Amendment 6

© Deleted per Amendment 5 Amendment No. 276

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.4 PRESSURIZER ............................................. 3/4 4-5 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY ................. 3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection System s .................................... 3/4 4-13 Operational Leakage .......................................... 3/4 4-15 3/4.4.7 Deleted .................................................... 3/4 4-17 3/4.4.8 SPECIFIC ACTIVITY ........................................ 3/4 4-20 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System ....................................... 3/4 4-24 D eleted .................................................... 3/4 4-29 3/4.4.10 STRUCTURAL INTEGRITY ................................. 3/4 4-30 3/4.4.11 Deleted ............................................... 3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 CORE FLOODING TANKS ................................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg >280°F ............................. 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 280°F ............................ 3/4 5-6 3/4.5.4 BORATED WATER STORAGE TANK .......................... 3/4 5-7 DAVIS-BESSE, UNIT I V Amendment No. 43-572047-2345 5 276

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety V alves ............................................... 3/4 7-1 Auxiliary Feedwater System ................................... 3/4 7-4 Condensate Storage Tanks .................................... 3/4 7-6 A ctivity .............................. 3/4 7-7 M ain Steam Line Isolation Valves .............................. 3/4 7-9 Motor Driven Feedwater Pump System .......................... 3/4 7-12a Main Feedwater Control Valves and Startup Feedwater Control Valves. 3/4 7-12d Turbine Stop Valves ......................................... 3/4 7-12e 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION. 3/4 7-13 3/4.7.3 COMPONENT COOLING WATER SYSTEM .................... 3/4 7-14 3/4.7.4 SERVICE WATER SYSTEM ................................. 3/4 7-15 3/4.7.5 ULTIMATE HEAT SINK ..................................... 3/4 7-16 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ....... 3/4 7-17 3/4.7.7 SNUBBERS ............................................... 3/4 7-20 3/4.7.8 SEALED SOURCE CONTAMINATION ........................ 3/4 7-36 3/4.7.9 STEAM GENERATOR LEVEL ................................ 3/4 7-38 3/4.7.10 Deleted 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES O perating .................................................. 3/4 8-1 Shutdow n .................................................. 3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating .................................. 3/4 8-6 A.C. Distribution - Shutdown ................................. 3/4 8-7 D.C. Distribution - Operating .................................. 3/4 8-8 D.C. Distribution - Shutdown .................................. 3/4 8-11 DAVIS-BESSE, UNIT I VII Amendment No. 84;

+06.-1350

-164-,-1-74-2-46, 276

INDEX BASES SECTION PAGE 3/4.0 A PPLICABILITY ........................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL ..... .......................... B 3/4 1-1 3/4.1.2 BORATION SYSTEM S ..................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ........................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS ............................. B 3/4 2-1 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEMS INSTRUMENTATION ............................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION .................... B 3/4 3-2 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS .............................. B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES ............................... B 3/4 4-1 3/4.4.4 PRESSURIZER ........................................... B 3/4 4-2 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY .................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ................... B 3/4 4-4 3/4 .4.7 D eleted .................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY ...................................... B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ........................ B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY ............................... B 3/4 4-13 3/4 .4.11 D eleted .................................................. B 3/4 4-13 DAVIS-BESSE, UNIT I IX Amendment No. i35, 20-1 234, 276

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ......................................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION B 3/4 7-4 3/4.7.3 COMPONENT COOLING WATER SYSTEM ................... B 3/4 7-4 3/4.7.4 SERVICE WATER SYSTEM ................................. B 3/4 7-4 3/4.7.5 ULTIM ATE HEAT SINK .................................... B 3/4 7-4a 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ...... B 3/4 7-4a 3/4.7.7 SN U B B ER S ............................................... B 3/4 7-5 3/4.7.8 SEALED SOURCE CONTAMINATION ........................ B 3/4 7-6 3/4.7.9 STEAM GENERATOR LEVEL ............................... B 3/4 7-6 I 3/4.7. 10 Deleted 3/4.8 ELECTRICAL POWER SYSTEMS ............................. B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................ B 3/4 9-1 3/4.9.2 INSTRUM ENTATION ...................................... B 3/4 9-1 3/4.9.3 D ECA Y TIM E ............................................. B 3/4 9-1 3/4.9.4 CONTAINMENT PENETRATIONS ........................... B 3/4 9-1 3/4.9.5 DELETED DAVIS-BESSE, UNIT 1 XII Amendment No. 38,- f06.-1-35,

- 1 74;-224P-2461-276

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPO N SIBILITY ................................................. 6-1 6.2 ORGANIZATION Offsite and Onsite Organizations .................................... 6-1 F acility Staff .................................................. 6-1 Facility Staff O vertim e ........................................... 6-2 6.3 FACILITY STAFF QUALIFICATIONS ................................ 6-3 6.4 DE LET E D ........................................................ 6-3 6.5 REVIEW AND AUDIT 6.5.1 Deleted 6.5.2 Deleted 6.5.3 Technical Review and Control ...................................... 6-4 6.6 D EL ET E D ..................................................... 6-5 6.7 D E LET E D ..................................................... 6-5 6.8 PROCEDURES AND PROGRAMS ................................. 6-5 6.9 REPORTING REQUIREMENTS 6.9.1 Routine R eports ................................................. 6-13 6.9.2 Special R eports .................................................. 6-16 6.10 RECORD RETENTION .......................................... 6-16 6.11 DELETED 6.12 HIGH RADIATION AREA ........................................ 6-16 6.13 ENVIRONMENTAL QUALIFICATION ............................. 6-19 6.14 D EL ET ED ..................................................... 6-19 6.15 OFFSITE DOSE CALCULATION MANUAL (ODCM) ................. 6-20 6.16 CONTAINMENT LEAKAGE TESTING PROGRAM ................... 6-21 6.17 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM... 6-22 DAVIS-BESSE, UNIT 1 XV Amendment No. 3 8,-1-3-5170,-1.8.,23-t,.-235,

-236,-240g-,2-44,-248-,-249-,-2-7-2-, 276

DEFINITIONS

c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

QUADRANT POWER TILT 1.18 QUADRANT POWER TILT is defined by the following equation and is expressed in percent.

QUADRANT POWER TILT =

100 ( Power in any core quadrant Average power of all quadrants -1)

DOSE EQUIVALENT 1-131 1.19 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (pCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.20 E-AVERAGE DISINTEGRATION ENERGY shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies DAVIS-BESSE, UNIT I 1-4 Amendment No. 276

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 1% Ak/k.

APPLICABILITY: MODES 1, 2*, 3**, 4 and 5.

ACTION:

With the SHUTDOWN MARGIN < 1% Ak/k, immediately initiate and continue boration at > 25 gpm of 7875 ppm boron or its equivalent, until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 1% Ak/k:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODES 1 or 2, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, by verifying that regulating rod groups withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 24" within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading by consideration of the factors of e. below, with the regulating rod groups at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.4 See LCO 3.7.9, Steam Generator Level, for additional SHUTDOWN MARGIN requirements.

'With kff > 1.0 "With kff < 1.0 DAVIS-BESSE, UNIT I 3/4 1-1 Amendment No. + -4-9-2;276

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 a. SG tube integrity shall be maintained, and

b. All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION:

Note: These ACTIONS may be entered separately for each SG tube.

a. With one or more SG tubes satisfying the tube repair criteria and not plugged or repaired in accordance with the Steam Generator Program,

1. Within 7 days, verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and

2. Plug or repair the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

b. With SG tube integrity not maintained, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged or repaired in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

DAVIS-BESSE, UNIT 1 3/4 4-6 Amendment No.-g,-2 l-,-27,-62.,-

(next page is 3/4 4-13) -- H-I-,-1-3 7-1;-t.4 -192. 20; 2-24-2-5:2;2

• 276

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. I GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 10 GPM CONTROLLED LEAKAGE, and

f. 5 GPM leakage from any Reactor Coolant System Pressure Isolation Valve as specified in Table 3.4-2.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE or primary to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> except as permitted by paragraph c below.

C. In the event that integrity of any pressure isolation valve specified in Table 3.4-2 cannot be demonstrated, POWER OPERATION may continue, provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.(a

d. The provisions of Section 3.0.4 are not applicable for entry into MODES 3 and 4 for the purpose of testing the isolation valves in Table 3.4-2.

(')Motor operated valves shall be placed in the closed position and power supplies deenergized.

DAVIS-BESSE, UNIT 1 3/4 4-15 E~rdei-dtd-4920AH--

Amendment No. 4 8*2-O 276

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each.

of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Monitoring the containment sump level and flow indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals to the makeup system when the Reactor Coolant System pressure is 2185 +/- 20 psig at least once per 31 days.

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation. (1)(2)

e. Verifying that primary to secondary leakage is _<150 gallons per day through any one steam generator, at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. (2) 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:

a. After each refueling outage,

b. Whenever the plant has been in COLD SHUTDOWN for 7 days, or more, and if leakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.

4.4.6.2.3 Whenever the integrity of a pressure isolation valve listed in Table 3.4-2 cannot be demonstrated, determine and record the integrity of the high pressure flowpath on a daily basis.

Integrity shall be determined by performing either a leakage test of the remaining pressure isolation valve, or a combined leakage test of the remaining pressure isolation valve in a series with the closed motor-operated containment isolation valve. In addition, record the position of the closed motor-operated containment isolation valve located in the high pressure piping on a daily basis.

SNot applicable to primary to secondary leakage.

(2) Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

DAVIS-BESSE, UNIT 1 3/4 4-16 -Order-d-ated 412 6/-8-1 Amendment No. -44.1QI620,276

3/4.7 PLANT SYSTEMS 3/4.7.9 STEAM GENERATOR LEVEL LIMITING CONDITION FOR OPERATION 3.7.9 Each Steam Generator shall have a minimum water level of 18 inches and the maximum specified below as applicable:

MODES I and 2:

a. The acceptable operating region of Figure 3.7-1.

MODE 3

b. 50 inches Startup Range with the SFRCS Low Pressure Trip bypassed and one or both Main Feedwater Pump(s) capable of supplying Feedwater to any Steam Generator.

c. 96 percent Operate Range with:

1. The SFRCS Low Pressure Trip active, or

2. The SFRCS Low Pressure Trip bypassed and both Main Feedwater Pumps incapable of supplying Feedwater to the Steam Generators.

MODE 4:

d. 625 inches Full Range Level APPLICABILITY: MODES 1,2, 3, and 4, as above.

ACTION:

With one or more steam generator's water level outside the limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.9 The steam generator shall be demonstrated OPERABLE by verifying steam generator level to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• Establish adequate SHUTDOWN MARGIN to ensure the reactor will stay subcritical during a MODE 3 Main Steam Line Break.

DAVIS-BESSE, UNIT I 3/4 7-38 Amendment No.-2-l,-1-7-1---9-, 276

Figure 3.7-1 Maximum Allowable Steam Generator Level in MODES I and 2 100 -

(43,96)

P,n

• 80 -

W= Unacceptable Ope ratiLng 4,1 W

I..

~.70 - Reio 0

• " 60-I' Acceptable Operating Region 50 (0,43) 40 10 Main Steam Superheat (OF)

DAVIS-BESSE, UNIT I 3/4 7-39 Amendment No. 192, 276

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his/her absence.

6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for facility operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels up to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Updated Safety Analysis Report.

b. 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.

c. The plant manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 FACILITY STAFF

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be in the control panel area when fuel is in the reactor.

DAVIS-BESSE, UNIT 1 6-1 Amendment No. 9 3_,-1-137,-272-,- 276

6.0 ADMINISTRATIVE CONTROLS 6.2.2 (Continued)

c. At least two licensed Operators, one of which has a Senior Reactor Operator license, shall be present in the control room while in MODES 1, 2, 3, or 4.

d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactoro.

e. All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f. Deleted

g. The operations manager shall either hold or have held a senior reactor operator's license on a pressurized water reactor. The assistant operations manager shall hold a senior reactor operator license for the Davis-Besse Nuclear Power Station.

6.2.3 FACILITY STAFF OVERTIME Administrative controls shall be developed and implemented to limit the working hours of personnel who perform safety-related functions (e.g., senior reactor operators, reactor operators, auxiliary operators, health physicists, and key maintenance personnel). The controls shall include guidelines on working hours that ensure that adequate shift coverage is maintained without routine heavy use of overtime for individuals.

Any deviation from the working hour guidelines shall be authorized in advance by the plant manager or his/her designees, in accordance with approved administrative procedures, and with documentation of the basis for granting the deviation. Routine deviation from the above guidelines shall not be authorized.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the plant manager or his/her designee(s) to ensure that excessive hours have not been assigned.

The individual qualified in radiation protection procedures may be less than the minimum requirements for a period of time not to 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, provided immediate action is taken to fill the required position.

DAVIS-BESSE, UNIT 1 6-2 Amendment No. 9 -- 8;- 8,-8 9 8 1,- --,--1 ,-

3-, ~ , 1-2; -72q* 2 76

6.0 ADMINISTRATIVE CONTROLS TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION#

LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 SOL 1*

OL 2 1 Non-Licensed 2 Shift Technical Advisor 1** None Required

1. Shift crew composition may be less than the minimum requirements for a period of time not to 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 inunediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

• Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling supervising CORE ALTERATIONS.

• • One of the two required individuals filling the SOL positions may also assume the STA function provided the individual meets the qualifications for the combined SRO/STA position specified for Option I of the Commission's Policy Statement on Engineering.

Expertise on Shift. If this option is used for a shift, then the separate STA position may be eliminated for that shift.

6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N 18.1-1971 for comparable positions, except for (1) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor 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, and (3) the operations manager whose requirement for a senior reactor operator license is as stated in Specification 6.2.2.g.

6.4 Deleted 6.5 REVIEW AND AUDIT 6.5.1 Deleted 6.5.2 Deleted DAVIS-BESSE, UNIT I 6-3 Amendment No. 9T-2T 27, 32. 74,. 76.

-86,-84,-3-,-98-,4 - 3 4-3 7,

-£3il.235;-2_36.-272. 276

6.0 ADMINISTRATIVE CONTROLS

.6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:

a. Plant procedures required by Section 6.8.1 and changes thereto shall be prepared, reviewed and approved. Each such procedure or procedure change shall be reviewed by an individual/group other than the individual/group which prepared the procedure or procedure change, but who may be from the same organization as the individual/group which prepared the procedure or procedure change. Plant procedures, (including plant administrative procedures), Physical Security Plan Implementing Procedures and Davis-Besse Emergency Plan Implementing Procedures will be approved by procedurally authorized individuals.

b. Temporary approval of changes to plant procedures cited in Section 6.8.1 which clearly do not change the intent of the approved procedures, can be made by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License.

For changes to plant procedures, which may involve a change in intent of the approved procedures, the person authorized in Section 6.5.3.1 a to approve the procedure shall approve the change

c. Proposed changes or modifications to plant structures, systems and components shall be reviewed as designated by procedurally authorized individuals. Each such modification shall be reviewed by an individual/group other than the individual/group which designed the modification, but who may be from the same organization as the individual/group which designed the modifications. Implementation of modifications to plant structures, systems and components shall be approved by proecdurally authorized individuals.

d. Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Safety Analysis Report shall be reviewed by an individual/group other than the individual/group which prepared the proposed test or experiment and shall be approved by procedurally authorized individuals.

e. Individuals responsible for reviews performed in accordance with Section 6.5.3.1 a, b, c and d above shall meet or exceed the appropriate qualification requirements of Section 4.2, 4.3.1, 4.4 or 4.6 of ANSI 18.1, 1971, and be previously designated by procedurally authorized individuals. Each such review shall include a determination of whether an additional, cross disciplinary, review is necessary. If deemed necessary, such review shall be performed by the review personnel of the appropriate discipline.

f. Each review will include a determination of whether prior NRC approval is required pursuant to 10 CFR 50.59.

DAVIS-BESSE, UNIT I 6-4 DAVISBESSE Amendment UNIT1No.6-4 4.09, 139,.248,.272,276

6.0 ADMINISTRATIVE CONTROLS 6.6 Deleted 6.7 Deleted 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, February, 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d, Physical Security Plan implementation.

e. Davis-Besse Emergency Plan implementation.

f. Fire Protection Program implementation.

g. The radiological environmental monitoring program.

h. Deleted.

i. Offsite Dose Calculation Manual implementation.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed and approved prior to implementation as set forth in 6.5.3 above.

6.8.3 Deleted 6.8.4 The following programs shall be established, implemented and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include makeup, letdown, seal injection, seal return, low pressure injection, containment spray, high pressure injection, waste gas, primary sampling and reactor coolant drain systems. The program shall include the following:

(i) Preventive maintenance and/or periodic visual inspection requirements, and DAVIS-BESSE, UNIT 1 6-5 Amendment No. 9-27,-5l -,6,-93,-9g, "09;-1-3"9, ,23-5) 248)260;272r-276

6.0 ADMINISTRATIVE CONTROLS 6.8.4.a (Continued)

(ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1) Training of personnel,

2) Procedures for monitoring, and

3) Provisions for maintenance of sampling and analysis equipment.

c. Deleted

d. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

I) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM.

2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

3) 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.

4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, DAVIS-BESSE, UNIT 1 6-6 Amendment No. 5t-84,--1-31-,-264, 276

6.0 ADMINISTRATIVE CONTROLS 6.8.4.d (Continued)

5) 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 ODCM at least every 31 days.

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1,

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) 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 greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) 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 Part 190.

e. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and DAVIS-BESSE, UNIT 1 6-7 Amendment No. 1-70., 276

6.0 ADMINISTRATIVE CONTROLS 6.8.4.e (Continued)

3) Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

f. Ventilation Filter Testing Program (VFTP):

A program shall be established to implement the following required testing of safety related filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2*,

ANSI/ASME N510-1980, and ASTM D 3803-1989.

1) Demonstrate for each of the safety related systems that an in-place test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1%

when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI/ASME N510-1980 at the system flowrate specified below, +/- 10%.

Safety Related Ventilation System Flowrate Shield Building Emergency Ventilation System 8000 cfm Control Room Emergency Ventilation System 3300 cfm

2) Demonstrate for each of the safety related systems that an in-place test of the charcoal adsorber shows a penetration and system bypass < I% when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSL'ASME N510-1980 at the system flowrate specified below, +/-10%.

Safety Related Ventilation System Flowrate Shield Building Emergency Ventilation System 8000 cfm Control Room Emergency Ventilation System 3300 cfm

• The periodic testing for the Shield Building Emergency Ventilation System and the Control Room Emergency Ventilation System are performed once each REFUELING INTERVAL.

The need for testing following painting, a fire, or a chemical release in any ventilation zone communicating with the Shield Building Emergency Ventilation System or the Control Room Emergency Ventilation System is as specified by the VFTP. The method of testing is based on Regulatory Guide 1.52, Revision 2, except for charcoal laboratory testing which will be performed in accordance with ASTM D 3803-1989.

DAVIS-BESSE, UNIT I 6-8 Amendment No. t70,244,-265,- 276

6.0 ADMINISTRATIVE CONTROLS 6.8.4.f (Continued)

3) Demonstrate for each of the safety related 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 D 3803-1989 at a temperature of 30' C and the relative humidity (RH) specified below.

Safety Related Ventilation System Penetration RH Shield Building Emergency Ventilation System < 2.5% 95%

Control Room Emergency Ventilation System < 2.5% 70%

4) Demonstrate for each of the safety related systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI/ASME N510-1980 at the system flowrate specified below,

+/- 10%.

Safety Related Ventilation System Delta P Flowrate Shield Building Emergency Ventilation System 6 inches Water Gauge 8000 cfm Control Room Emergency Ventilation System 4.4 inches Water Gauge 3300 cfm The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

DAVIS-BESSE, UNIT I 6-9 Amendment No.-2"4"4;2657 276

6.0 ADMINISTRATIVE CONTROLS 6.8.4 (Continued)

g. 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 Generator Program shall include the following provisions:

I1) Provisions for condition monitoring assessments: Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during a SG inspection outage, as determined from the inservice inspection results or by oth er means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.

2) Performance criteria for SG tube integrity: SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

a. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the fall range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against 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.

b. Accident induced leakage performance criterion: The primary to 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 I gpm per SG, except during a SG tube rupture.

c. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."

DAVIS-BESSE, UNIT I 61 6-10 mnmn No.

Amendment o 7 276

6.0 ADMINISTRATIVE CONTROLS 6.8.4.g (Continued)

3) Provisions for SG tube repair criteria

a. Tubes found by inservice inspection to contain flaws, in a region of the tube that contains no repair, with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired.

b. Sleeves found by inservice inspection to contain flaws, in a region of the sleeve that contains no sleeve joint, with a depth equal to or exceeding 40% of the nominal sleeve wall thickness shall be plugged.

c. Tubes with a flaw, in either the parent tube or the sleeve, within a sleeve-to-tube joint shall be plugged.

d. Tubes with a flaw in a repair roll shall be plugged.

4) 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-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. For tubes that have undergone repair rolling, the tube and tube roll, outboard of the new roll area in the tube sheet, can be excluded from inspections because it is no longer part of the pressure boundary once the repair roll is installed. For tubes that have undergone sleeving repairs; the segment of the parent tube between the bottom of the upper-most sleeve roll and the top of the middle sleeve roll can be excluded from inspection because it is no longer part of the pressureboundary orinethe sleeve is installed. In addition to meeting the requirements of 4.a through 4.e below, the inspection scope, inspection methods and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

a. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

b. Inspect 100% of the tubes at sequential periods of 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one interval between refueling outages (whichever is less) without being inspected.

DAVIS-BESSE, UNIT 1 6-11 Amendment No. 276

6.0 ADMINISTRATIVE CONTROLS 6.8.4.g.4 (Continued)

c. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one interval between refueling outages (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

d. During each periodic SG tube inspection, inspect 100% of the tubes that have been repaired by the repair roll process. This special inspection shall be limited to the repair roll joint and the roll transitions of the repair roll.

e. Inspect peripheral tubes in the vicinity of the secured internal auxiliary feedwater header between the upper tube sheet and the 15th tube support plate during each periodic SG tube inspection. The tubes selected for inspection shall represent the entire circumference of the steam generator and shall total at least 150 peripheral tubes.

5) Provisions for monitoring operational primary to secondary leakage.

6) Provisions for SG tube repair methods: Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

a. Sleeving in accordance with Topical Report BAW-2120P.

b. Repair rolling in accordance with Topical Report BAW-2303P, Revision 4. The new roll area must be free of flaws in order for the repair to be considered acceptable.

7) Special visual inspections: Visual inspections of the secured internal auxiliary feedwater header, header to shroud attachment welds, and the external header thermal sleeves shall be performed on each SG through the auxiliary feedwater injection penetrations. These inspections shall be performed during the third period of each ten-year Inservice Inspection Interval (ISI).

DAVIS-BESSE, UNIT I 6-12 Amendment No. 276

6.0 ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the appropriate Regional Office unless.

otherwise noted.

STARTUP REPORT 6.9.1.1 Deleted.

6.9.1.2 Deleted.

6.9.1.3 Deleted.

ANNUAL OPERATING REPORT 6.9.1.4 Annual reports covering the activities of the unit during the previous calendar year shall be submitted prior to March 31 of each year.

6.9.1.5 Reports required on an annual basis shall include:

a. Deleted

b. Deleted

c. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2)

Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

MONTHLY OPERATING. REPORT 6.9.1.6 Deleted DAVIS-BESSE, UNIT 1 6-13 Amendment No.-8;1-04

-I-3-5,-2-5&,-267-,- 276

6.0 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle and any remaining part of a reload cycle for the following:

2.1.2 AXIAL POWER IMBALANCE Protective Limits for Reactor Core Specification 2.1.2 2.2.1 Trip Setpoint for Flux -- AFlux/Flow for Reactor Protection System Setpoints Specification 2.2.1 3.1.1.3c Negative Moderator Temperature Coefficient Limit 3.1.3.6 Regulating Rod Insertion Limits 3.1.3.7 Rod Program 3.1.3.8 Xenon Reactivity 3.1.3.9 Axial Power Shaping Rod Insertion Limits 3.2.1 AXIAL POWER IMBALANCE 3.2.2 Nuclear Heat Flux Hot Channel Factor, FQ N

3.2.3 Nuclear Enthalpy Rise Hot Channel Factor, F AH 3.2.4 QUADRANT POWER TILT The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be: those previously reviewed and approved by the NRC, as described in BAW-101 79P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses", or any other new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A. The applicable approved revision number for BAW- 101 79P-A at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT. The CORE OPERATING LIMITS REPORT shall also list any new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW- 101 79P-A.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

DAVIS-BESSE, UNIT I 6-14 Amendment No. 144.+54)-189, 276

6.0 ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.10 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May I of each year. The report shall include summaries, interpretations, and analysis 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 (1) the ODCM, and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.11 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 (1) consistent with the objectives outlined in the ODCM and the Process Control Program, and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.g, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

DAVIS-BESSE, UNIT I 6-15 Amendment No. 86-,-1-70;,-1-8-,-2-7-2;, 276

6.0 ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission in accordance with 10 CFR 50.4 within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specifications:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b. Deleted

c. Deleted

d. Deleted

e. Deleted

f. Deleted

g. Inoperable Remote Shutdown System control circuit(s) or transfer switch(es) required for a serious control room or cable spreading room fire, Specification 3.3.3.5.2.

6.10 RECORD RETENTION Records of facility activities shall be retained as described in the USAR Chapter 17 Quality Assurance Program.

6.11 Deleted 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601 (c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

6.12.1 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 entry way to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

DAVIS-BESSE, UNIT 1 6-16 Amendment No. 9--5-,r6,3-,-94-, +06,

-F35-170--7A8-7O+/-~34;235276

6.0 ADMINISTRATIVE CONTROLS 6.12.1 (Continued)

b. Access to, and activities in, each such area shall be controlled by means of a 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 (e.g., health physics personnel) and personnel continuously escorted by such individuals may be exempted from the requirement for a RWP or equivalent while performing their assigned duties provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual (whether alone or in a group) entering such an area shall possess:

1) A radiation monitoring device that continuously displays radiation dose rates in the area; or

2) A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

3) A radiation monitoring device that continuously transmits dose rate and cumulative dose 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 be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, by personnel qualified in radiation protection procedures responsible for controlling personnel radiation exposure in the area.

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

6.12.2 Locked high radiation areas with dose rates greater than 1.0 rem/hour 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 any 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 door, gate, or other barrier that prevents unauthorized entry, and, in addition:

DAVIS-BESSE, UNIT I 6-17 DAVI-BESEUNI Amendment I -17No. 1, 2 76.

6.0 ADMINISTRATIVE CONTROLS 6.12.2.a (Continued)

1) All keys to such doors, gates, or other barriers shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee.

2) Doors, gates, or other barriers 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 radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for a RWP or equivalent while performing radiation surveys in such areas provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual (whether alone or in a group) entering such an area shall possess:

1) A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

2) A radiation monitoring device that continuously transmits dose 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, by 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, by 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 76 DAVIS-BESSE, UNIT I 6-18 Amendment No. 23-1,2

6.0 ADMINISTRATIVE CONTROLS 6.12.2.d (Continued)

4) In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

e. Except for an individual qualified in radiation protection procedures, 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 that is controlled as a high radiation area, where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area,.need not be controlled by a locked door or gate, but shall be barricaded and conspicuous, and a clearly visible flashing light shall be activated at the area as a warning device.

6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License NPF-3 dated October 24, 1980.

6.13.2 By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.14 Deleted DAVIS-BESSE, UNIT 1 6-19 Order dated 10/24/80 Amendment No. 863-170j-234,.235,

-2605-272;- 276

6.0 ADMINISTRATIVE CONTROLS 6.15 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained as required by the USAR Chapter 17 Quality Assurance Program.- This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s), and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose.

or setpoint calculations.

b. Shall become effective after the approval of the plant manager.

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

DAVIS-BESSE, UNIT I 6-20 Amendment No. 86;7--184T-231t,

-26G,-2-72-, 276

6.0 ADMINISTRATIVE CONTROLS 6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM

a. A program shall establish 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) A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

2) The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, As-found testing will not be required.

b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 38 psig.

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.50% of 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 < 0.75 La, for Type A tests, < 0.60 La for all penetrations and valves subject to Type B and Type C tests, and < 0.03 La for all penetrations that are secondary containment bypass leakage paths;

2) A single penetration leakage rate of< 0.15 La for each containment purge penetration;

3) Air lock acceptance criteria are:

a) Overall air lock leakage rate is < 0.015 La when tested at> Pa, b) For each door, seal leakage rate is < 0.01 La when the volume between the door seals is pressurized to > 10 psig.

e. The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

f. The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

DAVIS-BESSE, UNIT I 6-21 Amendment No. 240, 276

6.0 ADMINISTRATIVE CONTROLS 6.17 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 appropriate 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 USAR 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 USAR.

d. Proposed changes that meet the criteria of 6.17b. I and 6.17b.2 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).

DAVIS-BESSE, UNIT I 6-22 Amendment No. -249-,276