GNRO-2010/00042, Technical Specification Bases & Technical Requirements Manual Update

Technical Specification Bases & Technical Requirements Manual Update

ML101480899
Person / Time
Site:	Grand Gulf
Issue date:	05/27/2010
From:	Perino C Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2010/00042
Download: ML101480899 (19)

Text

~Entergy Entergy Operations, Inc.

7003 Bald Hill Road P.O. Box 756 Port Gibson, MS 39150 Tel 601 437 6299 Christina L. Perino Manager Licensing GNRO-2010/00042 May 27,2010 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

SUBJECT:

Technical Specification Bases and Technical Requirements Manual Update to the NRC Dated May 27,2010 Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29

Dear Sir or Madam:

Pursuant to Grand Gulf Nuclear Station (GGNS) Technical Requirements Manual Section 1.04, Entergy Operations, Inc. hereby submits an update of all changes made to the GGNS Technical Requirements Manual since the last submittal (GNRO-2009/00061 dated October 15,2009). Additionally, Technical Specification Bases are submitted for all changes made since the last submittal (GNRO-2010/00009 dated January 20, 2010), in accordance with GGNS Technical Specification 5.5.11. These updates are consistent with update frequency listed in 10CFR50.71 (e).

This letter does not contain any commitments.

Should you have any questions, please contact Michael Larson at (601) 437-6685.

CLP\MJL

Attachment:

GGNS Technical Requirements Manual and Technical Specification Bases Revised Pages cc: (See Next Page)

GNRO-2010100042 Page 2 of 2 cc:

NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U.S. Nuclear Regulatory Commission ATTN: Mr. Elmo E. Collins, Jr. (w/2)

Region Administrator, Region IV 612 East Lamar Blvd, Suite 400 Arlington, TX 76011-4125 U. S. Nuclear Regulatory Commission ATTN: Mr. Carl F. Lyon, NRRlADRO/DORL (w/2)

Mail Stop OWFN/B B1 Washington, DC 20555-0001

ATTACHMENT to GNRO-2010/00042 Gran d G UIf Tec h* nlcaIS peci Ica Ion Bases ReVlsedPages LBDCR# BASES PAGES AFFECTED TOPIC of CHANGE 10010 B 3.1-38a, B 3.8-79 Editorial changes to selected pages to correct header and connectors.

10002 B 3.6-78, B 3.6-82 Revision to clarify drywell purge compressors are not credited for drywall source term dilution.

10006 B 3.4-20a, B 3.4-21, B 3.5..13, Changes as a result of adoption of mandatory B 3.5-14a, B 3.6..34, B 3.6..35 Appendix I, paragraph 1-3410(A) and (0) of the 2004 edition of the ASME OM code for Safety Relief Valve inservice testing.

09037 B 2.0-3, B 2.0-6, B 3.2-5, B 3.2-6, Changes resulting from implementation of B 3.2-8 Technical Specification Amendment 184.

Gran d GU If Tech* nlcaI RequlrementsM anuaIReVlsedP ag e LBDCR# TRM PAGES AFFECTED TOPIC of CHANGE 10013 6.9-10 Extension of Horizontal Fuel Transfer System surveillance freQuency from 7 to 31 days.

Reactor Core SLs B 2.1.1 BASES APPLICABLE Fuel Cladding InteQ.[it.Y (continued)

SAFETY ANALYSES ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER> 50% RTP. Thus a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.

Because of the design thermal hydraulic compatibility of the reload fuel designs with the cycle 1 fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.

2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an ADO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. Reference 6 describes the methodology used in determining the MCPR SL.

The calculated MCPR safety limit is reported to the customary three significant digits (i.e., X.XX); the MCPR operating limit is developed based on the calculated MCPR safety limit to ensure that at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The fuel vendor's critical power correlations are based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the correlations, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the (continued)

GRAND GULF B 2.0-3 LBDCR 09037

Reactor Core Sls B 2.1.1 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. ANF-524(P)(A) , Revision 2, Supplements 1 and 2, November 1990.

3. EMF-2209(P)(A) , Revision 2, September 2003.

4. 10 CFR 50.67, "Acci dent Source Term."

5. NEDC-33383-P, Revision 1, "GEXL97 Correlation Applicable to ATRIUM-IO Fuel," June, 2008.

6. NEDE-24011-P-A, GESTAR-II.

GRAND GULF B 2.0-6 LBDeR 09037

SLC System B 3.1.7 BASES ACTIONS A.l and A.2 when the boron concentration is in the Limited operation region (between 15.2 weight percent and 28.5 weight percent), the SBlC system contains sufficient boron to perform its design basis functions. But the associated solution temperatures required to prevent precipitation of the boron from solution is potentially greater than the primary containment's ambient temperature. AS a result, the non safety tank heaters may be required to maintain the tank (continued)

GRAND GULF B 3.1-38a LBDCR 10010

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Minimum Critical Power Ratio (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly pcwer that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref. 1),

the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the AOOs to establish the operating limit MCPR are presented in the UFSAR, Chapters 4, 6, and 15, and References 2, 3, 4, and 5. To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (~CPR). When the largest ~CPR is added to the MCPR SL, the required operating limit MCPR ;s obtained.

(continued)

GRAND GULF B 3.2-5 LBDCR 09037

MCPR B 3.2.2 BASES APPLICABLE The MCPR operating limits derived from the transient SAFETY ANALYSES analysis are dependent on the operating core flow and power (continued) state (MCPRf and MCPR p , respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 3, 4, and 5). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref.

6) and the steady state thermal hydraulic code (Ref. 2).

MCPRf curves are provided based on the maximum credible flow runout transient for Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation ;s the runout of only one loop because both recirculation loops are under independent control.

Power dependent MCPR limits (MCPR p ) are determined by the three dimensional BWR simulator code and the one dimensional transient code (Ref. 7). The MCPR p limits are established for a set of exposure intervals. The limiting transients are analyzed at the limiting exposure for each interval.

Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPR p operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement.

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPRf and MCPR p limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP ;s unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

(continued)

GRAND GULF B 3.2-6 LBDCR 09037

MCPR B 3.2.2 BASES (continued)

SURVEILLANCE REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is ~ 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER reaches ~ 25% RTP is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1. NUREG-0562, "Fuel Failures As A Consequence of Nucleate Boiling or Dry Out," June 1979.

2. NEDE-24011-P-A General Electric Standard Application for Reactor Fuel (GESTAR II).

3. UFSAR, Chapter 15, Appendix 15B.

4. UFSAR, Chapter 15, Appendix 15C.

5. UFSAR, Chapter IS, Appendix 150.

6. NEDE-30130-P-A, Steady-State Nuclear Methods.

7. NEDO-24154, Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors.

GRAND GULF B 3.2-8 LBDCR 09037

S/RVS B 3.4.4 BASES SURVEILLANCE SR 3.4.4.3 (continued)

REQUIREMENTS verify that the valve is functioning properly. This SR can be demonstrated by one of two methods. If performed by method 1), plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements (Ref. 6), prior to valve installation. Therefore, this SR is modified by a Note that states the surveillance is 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 reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If performed by method 2), valve OPERABILITY has been demonstrated for all installed S/RVS based upon the successful operation of a test sample of S/RVs.

1. Manual actuation of the S/RV, with verification of the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow (e.g.,

tailpipe temperature or pressure). Adequate reactor steam pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVS divert steam flow upon opening. sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is consistent with the pressure recommended by the valve manufacturer.

2. The sample population of S/RVS tested each refueling outage to satisfy SR 3.4.4.1 will be stroked in the relief mode during uas-found" testing to verify proper operation of the S/RV. Just prior to installation of the to be newly-installed S/RVs to satisfy 3.4.4.1 the valve will be stroked in the relief mode during certification testing to verify proper operation of the S/RV. The successful performance of the test sample of S/RVS will perform in a similar fashion.

After the S/RVS are replaced, the electrical and pneumatic connections shall be verified either through (continued)

GRAND GULF B 3.4-20a LBDCR 10006

S/RVS B 3.4.4 BASES SURVEILLANCE SR 3.4.4.3 (continued)

REQUIREMENTS mechanical/electrical inspection or test prior to the resumption of electric power generation to ensure that no damage has occurred to the S/RV during transportation and installation.

This verifies that each replaced S/RV will properly perform its intended function.

If the valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

The STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV relief-mode actuator is alternately tested. The Freguency of the required relief-mode actuator testing was developed based on the S/RV tests required by the ASME Boiler and Pressure vessel code, section XI (Ref.

1) as implemented by the Inservice Testing program of specification 5.5.6. The testing Frequency required by the Inservice Testing program is based on operating experience and valve performance. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(Reference 5)

REFERENCES 1. ASME, Boiler and Pressure vessel code, Sections III and XI.

2. UFSAR, section 5.2.2.2.3.

3. UFSAR, section 15.

4. GNRI-96/00134, Amendment 123 to the operating License.

5. GNRI-96/00229, Amendment 130 to the operating License.

6. ASME code of operation and Maintenance of Nuclear power plants, part 1.

GRAND GULF B 3.4-21 LBDCR 10006

ECCS - operat; ng B 3.5.1 BASES SURVEILLANCE SR 3.5.1.7 (continued)

REQUIREMENTS provides a reasonable time to complete the SR. If performed by method 2), valve OPERABILITY has been demonstrated for all installed ADS valves based upon the successful operation of a test sample of S/RVs.

1. Manual actuation of the ADS valve, with verification of the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow (e.g., tailpipe temperature or pressure). Adequate reactor steam pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passlng through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is consistent with the pressure recommended by the valve manufacturer.

2. The sample population of S/RVs tested each refueling outage to satisfy SR 3.4.4.1 will be stroked in the relief mode during "as-found" testing to verify proper operation of the S/RV. Just prior to installation of the to be newly-installed S/RVS to satisfy SR 3.4.4.1 the valve will be stroked in the relief mode during certification testing to verify proper operation of the S/RV. The successful performance of the test sample of S/RVS provides reasonable assurance that the remaining installed S/RVS will perform in a similar fashion. After the S/RVS are replaced, the electrical and pneumatic connections shall be verified either through mechanical/electrical inspection or test prior to the resumption of electrical power generation to ensure that no damage has occurred to the S/RV during transportation and lnstallation. This verifies that each replaced S/RV will properly perform its intended function.

SR 3.5.1.6 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

The STAGGERED TEST BASIS Frequency ensures that both solenoids for each ADS valve relief-mode actuator are (continued)

GRAND GULF B 3.5-13 LBDCR 10006

ECCS - operati ng B 3.5.1 BASES REFERENCES 17. GNRI-97/00181, Amendment 133 to the operating License.

(continued)

18. ASME/ANSI OM-1987, operation and Maintenance of Nuclear Power plants, oMa-1988 Addenda part 6, Inservice Testing of pumps in Light water Reactor power plants.

19. ASME code of operation and Maintenance of Nuclear power plants, part 1.

GRAND GULF B 3.5-14a LBDCR 10006

LLS valves B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 (continued)

REQUIREMENTS method 1), plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements (Ref. 3), prior to valve installation. Therefore, this SR is modified by a Note that states the surveillance is 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 reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If performed by method 2), valve OPERABILITY has been demonstrated for all installed LLS valves based upon the successful operation of a test sample of S/RVS.

1. Manual actuation of the LLS valve, with verification of the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow (e.g., tailpipe temperature or pressure). Adequate reactor steam pressure must be available to perform this test to avoid dama9ing the valve. Also, adequate steam flow must be passlng through the main turbine or turbine bypass valves to continue to control reactor pressure when the LLS valves divert steam flow upon opening. sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at whic~ this test is to be performed is consistent with the pressure recommended by the valve manufacturer.

2. The sample population of S/RVS tested each refueling outage to satisfy SR 3.4.4.1 will be stroked in the relief mode during "as-found" testing to verify proper operation of the S/RV. Just prior to installation of the to be newly-installed S/RVs to satisfy SR 3.4.4.1 the valve will be stroked in the relief mode during certification testing to verify proper operation of the S/RV. The successful performance of the test sample of S/RVs provides reasonable assurance that the remaining installed S/RVS will be perform in a similar fashion. After the S/RVS are replaced, the electrical and pneumatic connections shall be verified either through mechanical/electrical inspection or test prior and pneumatic connections shall be verified either through mechanical/electrical inspection or test prior to the resumption of electric power generation to ensure that no damage has occurred to the S/RV during transportation and lnstallation. This verifies that each replaced S/RV will properly perform its intended function.

(continued)

GRAND GULF B 3.6-34 LBDeR 10006

LLS valves B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 (continued)

REQUIREMENTS The STAGGERED TEST BASIS Frequency ensures that both solenoids for each LLS valve relief-mode actuator are alternatively tested. The Frequency of the required relief-mode actuator testing is based on the tests required by ASME OM part 1 (Ref. 3), as implemented by the Inservice Testing program of specification 5.5.6. The testing Frequency required by the Inservice Testing program is based on operating experience and valve performance.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (Reference 4)

SR 3.6.1.6.2 The LLS designed S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the automatic LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.5.4 overlaps this SR to provide complete testing of the safety function.

The 18 month Frequency is based on the need to perform this surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the surveillance were performed with the reactor at power.

operating experience has shown these components usually pass the surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation. This prevents a reactor pressure vessel pressure blowdown.

REFERENCES 1. GESSAR-II, Appendix 3B, Attachment A, section 3BA.8.

2. UFSAR, Section 5.2.2.2.3.3.

3. ASME code of operation and Maintenance of Nuclear Power plants, Part 1.

4. GNRI-96/00229, Amendment 130 to the operating License.

GRAND GULF B 3.6-35 LBDCR 10006

Drywell purge system B 3.6.3.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.3 Drywell Purge system BASES BACKGROUND The Drywell purge System ensures a uniformly mixed post accident containment atmosphere, thereby minimizing the potential for local hydrogen burns due to a pocket of hydrogen above the flammable concentration.

The drywell purge compressor also performs the function diluting the drywel1 source term with the containment and suppression pool environment by pressurizing the drywell and discharging the drywe11 source term through the drywel1 suppression pool vents with the implementation of the Alternative Source Term (Reference 3), this dilution of drywe11 source term is no longer credited in the Equipment Qualification analysis.

The Drywel1 purge system is an En9ineered safety Feature and is designed to operate followlng a loss of coolant accident (LOCA) in post accident environments without loss of function. The system has two independent subsystems, each consisting of a compressor and associated valves, controls, and piping. Each subsystem is sized to pump 1000 scfm. Each subsystem is powered from a separate emergency power supply. since each subsystem can provide 100% of the mixin9 requirements, the system will provide its design functl0n with a worst case single active failure.

Following a LOCA, the drywell is immediately pressurized due to the release of steam into the drywe11 environment.

This pressure is relieved by the lowering of the water level within the weir wall, clearing the drywel1 vents and allowing the mixture of steam and noncondensibles to flow into the primary containment through the suppression pool, removing much of the heat from the steam. The remaining steam in the drywell begins to condense. AS steam flow from the reactor pressure vessel ceases, the drywell pressure falls rapidly. Both drywell purge compressors start automatically 30 seconds after a LOCA signal is received from the Emergency Core cooling System instrumentation, but only when drywell pressure has decreased to within approximately 0.87 psi above primary containment pressure.

(continued)

GRAND GULF B 3.6-78 LBDCR 10002

Drywell purge System B 3.6.3.3 BASES SURVEILLANCE SR 3.6.3.3.2 (continued)

REQUIREMENTS that all associated controls are functioning properly. It also ensures that blockage, compressor failure, or excessive vibration can be detected for corrective action.

The 92 day Frequency is consistent with Inservice Testing pr09ram Frequencies, operating experience, the known rel,ability of the compressor and controls, and the two redundant subsystems available.

SR 3.6.3.3.3 operating each drywell purge subsystem for ~ 15 minutes and verifying that each drywell purge subsystem flow rate is

~ 1000 scfm ensures that each subsystem is capable of maintaining drywell hydrogen concentrations below the flammability limit. The 18 month Frequency is based on the need to perform this surveillance under the conditions that apply during a plant outage when the drywell boundary is not required. operating experience has shown that these components usually pass the surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.3.3.4 This SR verifies that the pressure differential required to open the vacuum breakers is ~ 1.0 psid and that the isolation valve differential pressure actuation instrumentation opens the valve at 0.0 to 1.0 psid (drywell minus containment). This SR includes a CHANNEL CALIBRATION of the isolation valve differential pressure actuation instrumentation. operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. Regulatory Guide 1.7, Revision 1.

2. UFSAR, section 6.2.5.

3. Technical specification Amendment 145 to GGNS operating License.

GRAND GULF B 3.6-82 LBDCR 10002

Di stri buti on systems - Operati ng 8 3.8.7 Table 8 3.8.7-1 (page 1 of 1)

AC and DC Electrical Power Dlstribution Systems TYPE NOMINAL DIVISION 1* DIVISION 2"1r DIVISION 3*

VOLTAGE AC Electric 4160 V lSAA 16A8 17AC Power Distribution 480 V LCCS lS8A1, lS8A2, 16881, 16882, ---

System lS8A3, 158A4, 16883, 16884, lS8AS, 158A6 16885, 16886 480 V MCCS 15811, 15821, 16811, 16821, 17801, 17811 15831, 15841, 16831, 16841, 15851, 15861 16851, 16861 120 V Dist. 15P11, 15P21, 16p11, 16p21, 17p11 panels 15P31, 15P41, 16P31, 16p41, 15P51, 15p61 16P51, 16P61 DC Electric 125 V Bus I1DA Bus 11DB Bus 11DC Power Distribution Dist. 1DA1, 1DA2 1081, 1082 locI System panels

• Each division of the AC and DC electrical power distribution systems is a subsystem.

GRAND GULF B 3.8-79 LBDCR 10010

6.9 REFUELING OPERATIONS 6.9.7 HORIZONTAL FUEL T~SFER SYSTEM LCO 6.9.7 The horizontal fuel transfer system (HFTS) shall be operable.

APPLICABILITY: MODE 4 with the reactor mode switch in the Refuel position (as controlled under LCOs 3.10.2 or 3.10.4)

MODE 5 with the reactor mode switch in the Refuel position.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The HFTS inoperable. A.1 Suspend HFTS operation with Immediately the HFTS at either the Spent Fuel Building pool or the Reactor Containment Building pool terminal point.

SURVEILLANCE RE UIREMENTS SURVEILLANCE FREQUENCY SR 6.9.7.1 Verify: Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to operation

a. Room 1A525, Auxiliary Building, of HFTS.

elevation 182', the room through which the transfer system penetrates, is sealed.

31 days

b. ----------------------NOTE-------------------

Not required to be met for HFTS operations when equivalent administrative controls are in effect.

All interlocks with the refueling and fuel handling platforms are OPERABLE.

c. All HFTS primary carriage position indicators are OPERABLE.

TRM 6.9-10 LBDeR 10013