ML032820393

Tech Spec Pages for License Amendment 93, Relocates Technical Specifications Associated with Boration Systems and Chemistry and Revises Selected Technical Specifications Associated with Reactor Coolant System

ML032820393
Person / Time
Site:	Seabrook
Issue date:	10/03/2003
From:	NRC/NRR/DLPM
To:
References
TAC MB6614
Download: ML032820393 (26)

Text

--LiLA LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.1.2 BORATION SYSTEMS Isolation of Unborated Water Sources - Shutdown .................................. 3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height ............................................................ 3/4 1-15 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD .............. ................. 3/4 1-17 Position Indication Systems - Operating ........................................ ........... 3/4 1-18 Position Indication System - Shutdown .............................................. ...... 3/4 1-19 Rod Drop Time ............................................................ 3/4 1-20 Shutdown Rod Insertion Limit .... 3/4 1-21 Control Rod Insertion Limits .. . . 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE ..................... ............................... 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fa(Z) ........................................ 3/4 2-4 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ......................... 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATIO .................................................... 3/4 2-9 3/4.2.5 DNB PARAMETERS .................................................... 3/4 2-10 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION .................................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION .3/4 3-2 SEABROOK - UNIT 1 iii Amendment No. 8, 93

114LJCA LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 4.3-5 (THIS TABLE NUMBER IS NOT USED) .................. ...................... 3/4 3-58 Explosive Gas Monitoring Instrumentation ....................................... 3/4 3-60 TABLE 3.3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION ....................................... 3/4 3-61 TABLE 4.3-6 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ....................................... 3/4 3-64 3/4.3.4 (THISSPECIFICATION NUMBER IS NOT USED) .. .. 3/4 3-67 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation ... 3/4 4-1 Hot Standby .. . 3/4 4-2 Hot Shutdown ... 3/4 4-4 Cold Shutdown - Loops Filled ....................... 3/4 4-6 Cold Shutdown - Loops Not Filled....................... 3/4 4-7 3/4.4.2 SAFETY VALVES Shutdown ......................... 3/4 4-8 Operating .......................... 3/4 4-9 3/4.4.3 PRESSURIZER ......................... 3/4 4-10 3/4.4.4 RELIEF VALVES ......................... 3/4 4-11 3/4.4.5 STEAM GENERATORS.................................................................................................. 34 4-13 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION .. . . 3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION .. . . 3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.... 3/4 4-20 Operational Leakage .... 3/4 4-21 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES .... 3/4 4-24 3/4.4.7 (THIS SPECIFICATION NUMBER NOT USED) .... 3/4 4-25 TABLE 3.4-2 (THIS TABLE NUMBER IS NOT USED) .. . . 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY .............................................................. 3/4 4-27 SEABROOK - UNIT v Amendment No. 59, 6 93

INULA LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.12.2 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-3 3/4.12.3 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY .............................................. 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 SYSTEMS .............................................. B 3/4 1-3 3/4.1.3 MOVABLE CONTROL ASSEMBLIES .................................. .. ......................... B 3 4 14 3/4.2 POWER DISTRIBUTION LIMITS............................................................................ B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE .............................................. B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ................................... B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO ................... ............................ B 3/4 2-3 3/4.2.5 DNB PARAMETERS .............................................. B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .............B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ........................................................ B 3/4 3-3 3/4.3.4 (THIS SPECIFICATION NUMBER IS NOT USED) ....................... ................. B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION. B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................................................................ B 3/4 4-1 3/4.4.3 PRESSURIZER. B 3/4 4-2 3/4.4.4 RELIEF VALVES .B 3/4 4-2 3/4.4.5 STEAM GENERATORS.................................................................................. B 3/4 4-2 314.4.6 REACTOR COOLANT SYSTEM LEAKAGE................................................... B 3/4 4-3 3/4.4.7 (THIS SPECIFICATION NUMBER IS NOT USED) .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 4-7 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E > I MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................................................................ B 3/4 4-9 FIGURE B 3/4.4-2 (This figure number not used) .................................. B 3/4 4-10 SEABROOK - UNIT 1 x Amendment No. 50,66,74, 93

PAGE INTENTIONALLY BLANK SEABROOK - UNIT 1 3/4 1-7 Amendment No. 93

PAGE INTENTIONALLY BLANK SEABROOK- UNIT I 3/4 1-8 Amendment No.0, 93

PAGE INTENTIONALLY BLANK SEABROOK - UNIT I 3/4 1-9 Amendment No.74, 83, 93

PAGE INTENTIONALLY BLANK SEABROOK- UNIT I 3/4 1-10 Amendment No. 0, 74, 83, 93

PAGE INTENTIONALLY BLANK SEABROOK - UNIT 1 3/4 1-11 Amendment No. 42 93

PAGE INTENTIONALLY BLANK SEABROOK - UNIT 1 3/4 1-12 Amendment No. 42 93

PAGE INTENTIONALLY BLANK SEABROOK - UNIT 1 314 1 -13 Amendment No.: 93

REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS ISOLATION OF UNBORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 Provisions to isolate the Reactor Coolant System from unborated water sources shall be OPERABLE with:

a. The Boron Thermal Regeneration System (BTRS) isolated from the Reactor Coolant System, and

b. The Reactor Makeup Systems inoperable except for the capability of delivering up to the capacity of one Reactor Makeup Water pump to the Reactor Coolant System.

APPLICABILITY: MODES 4, 5, and 6 ACTION:

With the requirements of the above specification not satisfied immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and, if within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the required SHUTDOWN MARGIN is not verified, initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored and the isolation provisions are restored to OPERABLE.

SURVEILLANCE REQUIREMENTS 4.1.2.7 The provisions to isolate the Reactor Coolant System from unborated water sources shall be determined to be OPERABLE at least once per 31 days by:

a. Verifying that at least the BTRS outlet valve, CS-V-302, or the BTRS moderating heat exchanger outlet valve, CS-V-305, or the manual outlet isolation valve for each demineralizer not saturated with boron, CS-V-284, CS-V-295, CS-V-288, CS-V-290, CS-V-291, is closed and locked closed, and

b. Verifying that power is removed from at least one of the Reactor Makeup Water pumps, RMW-P-1 6A or RMW-P-16B.

A demineralizer may be unisolated to saturate a bed with boron provided the effluent is not directed back to the Reactor Coolant System.

SEABROOK - UNIT /1-4AedetN.9 3/4 1-14 Amendment No. 9 3

REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two reactor coolant loops shall be OPERABLE with two reactor coolant loops in operation when the Control Rod Drive System is capable of rod withdrawal and one reactor coolant loop in operation when the Control Rod Drive System is not capable of rod withdrawal.*

APPLICABILITY: MODE 3.

ACTION:

a. With less than two reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With only one reactor coolant loop in operation and the Control Rod Drive System is capable of rod withdrawal, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> return the required reactor coolant loop to operation or place the Control Rod Drive System in a condition incapable of rod withdrawal.

c. With no reactor coolant loop in operation, place the Control Rod Drive System in a condition incapable of rod withdrawal, and suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.

• AII reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10oF below saturation temperature.

SEABROOK - UNIT 1 3/4 4-2 Amendment No. 34, 93

REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION HOT STANDBY SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps shall be determined I OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability*. I 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 14% at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.

SEABROOK- UNIT 1 3/4 4-3 Amendment No. 9 3

REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 92%

of pressurizer level (1656 cubic feet), and at least two groups of pressurizer heaters each having a capacity of at least 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1,2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the pressurizer otherwise inoperable, fully insert all rods, place the Control l Rod Drive System in a condition incapable of rod withdrawal, and 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within Us limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters from the emergency power supply and measuring circuit current at least once each refueling interval.

SEABROOK - UNIT 1 3/-A 41 0 Amendment No. 34 93

REACTOR COOLANT SYSTEM 3/4.4.7 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT 1 3/4hd 4-25 Amendment No. 46 93

TABLE 3.4-2 (THIS TABLE NUMBER IS NOT USED)

SEABROOK - UNIT 1 3/4 4-26 Amendment No. 46,93

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6.

ACTION:

a. With one of the above required monitors Inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b. With both of thebabove required monitors Inoperable or not operating, immediately initiate corrective action to restore one source range neutron flux monitor to OPERABLE status and determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK 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. A CHANNEL CALIBRATION* at least once per 18 months. I I

• Neutron detectors may be excluded from CHANNEL CALIBRATION.

SEABROOK - UNIT I 3/4 9-2 Amendment No. 9 3

REACTIVITY CONTROL SYSTEMS BASES BORATION CONTROL 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued)

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

The cycle specific upper MTC limit in the COLR is determined during the design of each cycle. The upper MTC limit provides assurance of compliance with the ATWS Rule and the basis for the Rule by limiting core damage frequency from an ATWS event below the target of 1.0 x 10-5 per reactor year established in SECY-83-293. The COLR limit will also assure that the core will have an MTC less positive than -8 PCM/DEG F for at least 95% of the cycle time at full power.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, the MTC is measured as required by Surveillance Requirement 4.1.1.3.a. A measurement bias is derived from the difference between test measurement and test prediction. All predicted values of MTC for the cycle are conservatively corrected based on measurement bais. The corrected predications are then compared to the maximum upper limit of Technical Specification 3.1.1.3. Control rod withdrawal limits are established, if required, to assure all corrected values of predicted MTC will be less positive than the limit specified in the COLR, and the maximum upper limit required by Technical Specification 3.1.1.3.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 5510 F. This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RTNDT temperature.

SEABROOK - UNIT B 314 1-2 Amendment No.41 93

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS The limitations on OPERABILITY of isolation provisions for the Boron Thermal Regeneration System and the Reactor Water Makeup System in Modes 4, 5, and 6 ensure that the boron dilution flow rates cannot exceed the value assumed in the transient analysis.

A resin bed is considered saturated with boron when the effluent boron concentration is within 5% or 5 ppm, whichever is greater, of the Reactor Coolant System boron concentration at the time the resin bed was saturated. Saturation ensures that no further boron may be removed by the resin bed regardless of the current boron concentration.

SEABROOK - UNIT B 34 1-3 Amendment No. 9, 31442,4 4 93

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION A reactor coolant loop is comprised of its associated steam generator and reactor coolant pump. An OPERABLE reactor coolant system loop consists of an OPERABLE reactor coolant pump and an OPERABLE steam generator in accordance with the Steam Generator Tube Surveillance Program.

The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant 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 />.

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by placing the Control Rod Drive System in a condition incapable of rod withdrawal. Single failure considerations require that two loops be OPERABLE at all times.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.

A Reactor Coolant "loops filled" condition is defined as follows: (1) Having pressurizer level greater than or equal to 50% if the pressurizer does not have a bubble, and greater than or equal to 17% when there is a bubble in the pressurizer. (2) Having the air and non-condensables evacuated from the Reactor Coolant System by either operating each reactor coolant pump for-a short duration to sweep air from the Steam Generator U-tubes into the upper head area of the reactor vessel, or removing the air from the Reactor Coolant System via an RCS evacuation skid, and (3) Having vented the upper head area of the reactor vessel if the pressurizer does not have a bubble. (4)

Having the Reactor Coolant System not vented, or if vented capable of isolating the vent paths within the time to boil.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting an RCP in MODES 4 and 5 are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 500 F above each of the RCS cold-leg temperatures.

SEABROOK - UNIT 1 B 3/4 4-1 Amendment No. 93

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

SEABROOK - UNIT 1 B 3/4 4-1 a Amendment No. 931

REACTOR COOLANT SYSTEM BASES 3/4.4.7 (THIS SPECIFICATION NUMBER IS NOT USED) I 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident i conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Seabrook site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

SEABROOK - UNIT 1 B 34 4-5 Amendment No. 34 T9

BASES 3/4.9.2 INSTRUMENTATION BACKGROUND The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed source range neutron flux monitors are part of the Nuclear Instrumentation System (NIS). These detectors are located external to the reactor vessel and detect neutrons leaking from the core.

The installed source range neutron flux monitors are BF3 detectors operating in the proportional region of the gas filled detector characteristic curve. The detectors monitor the neutron flux in counts per second. The. instrument range covers six decades of neutron flux (1 E+6 cps) with an 11.5% instrument accuracy. The detectors also provide continuous visual indication in the control room and an audible count rate in the control room and containment to alert operators to a possible dilution accident. The source range circuitry provides a signal to the Shutdown Margin Monitor (SMM). The SMM provides an audible alarm in the control room to alert operators to a possible dilution accident. The NIS is designed in accordance with the criteria presented in Reference 1.

APPLICABLE SAFETY ANALYSES Two OPERABLE source range neutron flux monitors are required to provide a signal to alert the operator to unexpected changes in core reactivity such as with a boron dilution accident (Reference 2) or an improperly loaded fuel assembly. The audible count rate from the source range neutron flux monitors provides prompt and definite indication of any boron dilution. The count rate increase is proportional to the subcritical multiplication factor and allows operators to promptly recognize the initiation of a boron dilution event.

Prompt recognition of the initiation of a boron dilution event is consistent with the assumptions of the safety analyses and is necessary to assure sufficient time is available for isolation of the primary makeup water source before SHUTDOWN MARGIN is lost (Reference 2).

The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LIMITING CONDITION FOR OPERATION This LCO requires that two source range neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE, each monitor must provide visual indication in the control room. In addition, at least one of the two monitors must provide an OPERABLE audible count function to alert the operators to the initiation of a boron dilution event.

SEABROOK - UNIT I B 34 9-2a Amendment No. 93

614.Y t tLItI'LN Utrv- I BASES 3/4.9.2 INSTRUMENTATION (Continued)

APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, these same installed source range detectors and circuitry are also required to be OPERABLE by LCO 4.3.1.1, uReactor Trip System Instrumentation."

ACTIONS A

With only one source range neutron flux monitor OPERABLE, redundancy has been lost.

Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and any operation that would add positive reactivity must be suspended immediately. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation.

Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum refueling boron concentration.

This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Performance of Required Action A shall not preclude completion of movement of a component to a safe position.

B With no source range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately. Once initiated, action shall be continued until a source range neutron flux monitor is restored to OPERABLE status.

With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition is determined by performing SR 4.9.1.2 to ensure that the required boron concentration exists.

The Completion Time of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.

SEABROOK - UNIT 1 B 3/4 9-2b Amendment No. 93

3/4.9 REFUELING OPERATIONS BASES 3/4.9.2 INSTRUMENTATION (Continued)

SURVEILLANCE REQUIREMENTS SR 4.9.2.a SR 4.9.2.a is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to- a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.

The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 4.3.1.1.

SR4.9.2.b SR 4.9.2.b is the performance of a CHANNEL CALIBRATION every 18 months. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. The CHANNEL CALIBRATION also includes verification of the audible count rate function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 13, GDCP 26, GDC 28, and GDC 29.

2. FSAR, Section 15.4.6.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses.

SEABROOK - UNIT 1 B 314 9-2c Amendment No. 3

3/4.9 REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The Limiting Condition for Operation (LCO) limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetrations, the approved alternate closure methods and the containment personnel airlock.

For the approved alternate closure methods, the LCO requires that a designated individual must be available to close or direct the remote closure of the penetration in the event of a fuel handling accident. "Available" means stationed at the penetration or performing activities controlled by a procedure on equipment associated with the penetration.

For the personnel airlocks (containment or equipment hatch), the LCO ensures that the airlock can be closed after containment evacuation in the event of a fuel handling accident. The requirement that the airlock door is capable of being closed requires that the door can be closed and is not blocked by objects that cannot be easily and quickly removed. As an example, the use of removable protective covers for the door seals and sealing surfaces is permitted. The requirement for a designated individual located outside of the airlock area available to lose the door following evacuation of the containment will minimize the release of radioactive material.

The fuel handling accident analysis inside containment assumes both of the personnel airlock doors are open and an additional 12" diameter penetration (or equivalent area) is open. The analysis is bournded by these assumptions since all of the available activity is assumed to be released instantaneously from the containment to the atmosphere.

SEABROOK - UNIT 1 B 3/4 9-2d Amendment No. 3