ML030030480

Technical Specification Bases Pages

ML030030480
Person / Time
Site:	Monticello
Issue date:	12/19/2002
From:	Forbes J Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML030030480 (25)

Text

NMC*

Monticello Nuclear Generating Plant Committed to Nudear Elxcenle*

Operated by Nuclear Management Company, LLC December 19, 2002 Technical Specification 6.8.K U. S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Technical Specification Bases Pages Using the Monticello Technical Specification Bases Control Program, Monticello Technical Specification Bases pages have been changed. The affected pages are designated with the amendment applicable at the time and the suffix "a". The changes are summarized in Attachment A. Marked up pages applicable at the time the changes were made are provided in Attachment B. A final typed copy of the changed pages that are applicable, for entry into the NRC authority copy, are provided in Attachment C. The current copy of our list of effective pages and record of revision is attached for your information, as attachment D.

Please contact John Fields, Senior Licensing Engineer at (763) 295-1663 with any questions or comments.

Jeffrey S. Forbes Site Vice President Monticello Nuclear Generating Plant Attachment A - Summary of Technical Specification Bases Changes (TSBC)

Attachment B - Monticello Technical Specification Bases Pages Marked Up with Changes Attachment C - Revised Monticello Technical Specification Bases Pages Attachment D - Monticello Technical Specification List of Effective Pages and Record of Revision cc:

Regional Administrator - Ill, NRC NRR Project Manager, NRC Sr. Resident Inspector, NRC Minnesota Department of Commerce 2807 West County Road 75

• Monticello, Minnesota 55362-9637 Telephone: 763.295.5151

• Fax: 763.295.1454

Attachment A Summary of Technical Specification Bases Changes (TSBC)

Following is a summary of the bases changes forwarded herein. The changes have been processes in accordance with the Monticello Technical Specification Bases Control Program described in Technical Specification (TS) 6.8.K.

TSBC-130a Technical Specification Involved - 3.5.A and 3.5.D Pages affected - 111 and 113 (pages 112, 114 and 115 contained repaginated text only)

Summary of Change: Changes in TS Bases section 3.5.A consist of the following: When HPCI is inoperable; TS require verification that RCIC is operable. The change substitutes a verification of operability of the RCIC system and the redundant and diversified low pressure core cooling systems (administrative check) for a demonstration of RCIC system operability (surveillance test). The change was an oversight from TS amendment 77. TS amendment 77 removed the requirement to perform a surveillance test of RCIC for HPCI inoperability.

Changes in TS Bases section 3.5.D consist of the following: When RCIC is inoperable; TS require verification that HPCI is operable. The change substitutes a verification of operability of the HPCI system and the redundant and diversified low pressure core cooling systems (administrative check) for a demonstration of HPCI system operability (surveillance test). The change was an oversight from TS amendment 77. TS amendment 77 removed the requirement to perform a surveillance test of HPCI for RCIC inoperability.

Attachment B Monticello Technical Specification Bases Pages Marked Up with Changes This attachment consists of Monticello Technical Specification bases pages marked up with changes. The pages included are listed below:

Paqe 111 112 113 114 115

Bases 3.5/4.5 (Continued),:

The surveillance requirements provide. adequate assurance that the LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clod temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCI system continues to operate until reactor vessel pressure is below the pressure at which Core Spray system operation or LPCI mode of the RHR system operation maintains core cooling.

The flow tests for the HPCI System are performed at two different pressure ranges such that the system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be > 950 psig to perform SR 4.5.A.3.a and _5 165 psig to perform SR 4.5.A.3.b. Adequate steam flow is represented by total steam flow > 106 lb/hr. Reactor startup, and pressure increase to < 165 psig, is allowed prior to performing the low pressure surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the surveillance test is short. Therefore, pressure may be raised above 150 psig, but 2165 psig to perform this surveillance without entering an LCO for the HPCI System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or reason to believe that HPCI is inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.A.3.a and SR 4.5.A.3:b are modified by a note which states that the surveillances are 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 the reactor steam pressure and flow are adequate to perform the test.

With the HPCI system inoperable, adequate core cooling is assured by the operability of the redundant and diversified automatic depressurization system and both the Core Spray and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period-of 14 days is based on the d-m-o-nstrated operability of redundant and die4.rsifid low

.p-essui Gore

"_cooling systoms and the.RI, system.verified operability of the RCIC system and the redundant and diversified low pressure core cooling systems. Verification of RCIC operability may be oerformed as an administrative check by examining logs or other information to determine if RCIC Is out of service for maintenance or other reasons, It does not mean to perform the surveillance-needed to demonstrate the operability of the RCIC system,

[moved last paragraph to page 112 ]

3.5/4.5 BASES 111 08/0/401 Amendment No. 37, 63, 77, 79, 00a,*-,22

Bases 3.5/4.5 (Continued.:

[The surveillance requirements provide adequate assurance that the HPCI system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.]

Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 22000F. ADS is conservatively required to be operable whenever reactor vessel pressure exceeds 150 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.

ADS automatically controls three selected safety-relief valves although the safety analysis only takes credit for two valves. It is therefore appropriate to permit one valve to be out-of-service for up to 14 days Without materially reducing system reliability.

B.

RHR Intertie Line An intertie line is provided to connect the RHR suction line with the two RHR loop return lines. This four-inch line is equipped with three isolation valves. The purpose of this line is to reduce the potential for water hammer in the recirculation and RHR systems. The isolation valves are opened during a cooldown to establish recirculation flow through the RHR suction line and return lines, thereby ensuring a uniform cooldown of this piping. The RHR loop return line isolation valves receive a closure signal on LPCI initiation. In the event of an inoperable return line Isolation valve, there is a potential for some of the LPCI flow to be diverted to the broken lo6p during a loss of coolant accident. Surveillance requirements have been established to periodically cycle the RHR intertie line isolation valves. In the event of an inoperable RHR loop return line isolation valve, either the inoperable valve is closed. or the other two isolation valves are closed to prevent diversion of LPCI flow. The RHR intertie line flow Is not permitted in the Run Mode to eliminate 1) the need to compensate for the small change in jet pump drive flow or 2) a reduction in core flow during a loss of coolant accident.

C.

Containment Spray/Cooling Systems Two containment spray/cooling subsystems of the RHR system are provided to remove heat energy from the containment and control torus and drywell pressure in the event of a loss of coolant accident. A containment spray/cooling subsystem consists of 1 RHR service water pump, a RHR heat exchanger, 1 RHR pump, and valves and piping necessary for Torus Cooling and Drywell Spray. Torus Spray, is not considered part of a containment spray/cooling subsystem. Placing a containment spray/cooling subsystem into operation following a loss of coolant accident is a manual operation.

[moved last paragraph to page 113]

3.5/4.5 BASES 112 12/05/01.

Amendment No. 3:7, 77, 79, 8, O

-0a, 102, 122,124.

Bases 3.5/4.5 (Continued):

[ The most degraded condition for long term containment heat removal following the design basis loss of coolant accident results from the loss of one diesel generator. Under these conditions, only one RHR pump and one RHR service water pump in the redundant division can, be used for containment spray/cooling. The containment temperature and pressure have been analyzed under these conditions assuming service water and initial suppression pool temperature are both 900F.. Acceptable margins to containment design conditions have been demonstrated. Therefore the containment spray/cooling system is more than ample to provide the required heat removal capability. Refer to USAR Sections 5.2.3.3, 6.2.3.2.3, and 8.4.1.3. ]

During normal plant operation, the containment spray/cooling system provides cooling of the suppression pool water to maintain temperature within the limits specified in Specification 3.7.A.1.

The surveillance requirements and testing in accordance with the IST Program provide adequate assurance that the containrrfent spray/cooling system will be operable when required.

D.

RCIC The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the feedwater system is not available. The pumping capacity of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts. The system may also be manually initiated at any time.

With the RCIC system inoperable. tThe HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefore, the. 6pecific-ationM c-all8t for an operability check of the,4PGI system shou,.tld the R.IG system be*f.d4*to he inoperable.Therefore. the RCIC out-of-service period of 14 days is based on the verified operability of the HPCI system and the redundant and diversified low pressure cooling systems. Verification of HPCI operability may be performed as an administrative check by examining logs or other information to determine if HPCI is out of service for maintenance or other reasons., It does not mean to perform the surveillance needed to demonstrate the operability of the HPCI system.

The flow tests for the RCIC System are performed at two different pressure ranges such that the system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be ->950 psig to perform SR 4.5.D.1.a and *5165 psig to perform SR 4.5.D.1.b. Adequate steam flow is represented by

[moved 3 paragraphs to page 114]

3.5/4.5 BASES 113 08/0-110!

Amendment No. 77, 79, 93, 98, 1Q0aj-=--.

Bases 3.5/4.5 (Continued):

[ total steam flow 2 106 lb/hr. Reactor startup, and pressure increase to ! 165 psig, isallowed prior to performing the low pressure surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the surveillance test is short. Therefore, pressure may be raised above 150 psig, but _* 165 psig to perform this surveillance without entering an LCO for the RCIC System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or reason to believe that RCIC is inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.D.1.a and SR 4.5.D.1.b are modified by a note which states that the surveillances are not required to be performed until 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the reactor steam pressure and flow are adequate" to perform the test.

The surveillance requirements provide adequate assurance that the RC IC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge' piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.]

E.

Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times. It is during refueling outages that major maintenance is performed and during such time that air core and containment spray/cooling subsystems may be out of service.

This specification allows all core and containment spray/cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.1. In this situation, a sufficient inventory of water-is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal Integrity.

[ moved Section F to page 115]

3.5/4.5 BASES

.NE...T.P.A GE iS ! 21 114 98/01110:

Amendment No. 24, 47, 79, 97, no 2,122

Bases 3.5/4.5 (Continued):

[F. Recirculation System The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal-hydraulic instability, a power-flow exclusion region, to be avoided during normal operation, is calculated using the approved methodology as stated in specification 6 7.A.7. Since the exclusion region may change each fuel cycle the limits are contained in the Core Operating Limits Report. Specific directions are provided to avoid operation in this region and to immediately exit upon an entry. Entries into the exclusion region are not part of normal operation. An entry may occur as the result of an abnormal event such as a single recirculation pump trip. In these events, operation in the exclusion region may be needed to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow biased scram function will suppress oscillations prior to exceeding the fuel safety limit.

Power distribution controls are established to ensure the reactor is operated within the bounds of the stability analysis. With these controls in place, there is confidence that an oscillation will not occur outside of the stability exclusion region. Without these controls, it is theoretically possible to operate the reactor in such, a manner as to cause an oscillation outside of the exclusion region. A nominal 5%.

power-flow buffer region outside of the exclusion region is provided to establish a stability margin to the analytically defined exclusion region. The buffer region may be entered only when the power distribution controls are in place.

Continuous operation with one recirculation loop was analyzed and the adjustments specified in specification 3.5.F.3 were determined by NEDO-24271, June 1980, "Monticello Nuclear Generating Plant Single Loop.Operation;" NEDC-30492, April 1984, "Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant;"

and NEDC-32456P, July 1996. Specification 3.6.A.2 governs the restart of the pump in an idle recirculation loop. Adherence to. this specification limits the probability of excessive flux transients and/or thermal stresses.]

3.5/4.5 BASES NEXT PAGE IS 121 115 Amendment No.

Attachment C Revised Monticello Technical Specification Bases Pages This attachment consists of the revised Monticello Technical Specification Bases pages that incorporate the change. These pages should be entered into the NRC Authority copies of the Monticello Technical Specifications. The pages included are listed below:

Page 111 112 113 114 115

Bases 3.5/4.5 (Continued):

The surveillance requirements provide adequate assurance that the LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The puhnp discharge piping is maintained full to prevent water hammer damage to piping an'd to start cooling at the earliest moment.

The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCI system continues to operate until reactor vessel pressure is below the pressure at which Core Spray system operation or LPCI mode of the RHR system operation maintains core cooling.

The flow tests for the HPCI. System are performed at two different pressure ranges such that the system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be _> 950 psig to perform SR 4.5.A.3.a and < 165 psig to perform SR 4.5.A.3.b. Adequate steam flow is represented by total steam flow > 106 lb/hr. Reactor startup, and pressure increase to <5165 psig, is allowed prior to performing the low pressure.

surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the surveillance test is short. Therefore, pressure may be raised above 150 psig, but *<165 psig to perform this surveillance without entering an LCO for the HPCI System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfacto*'ily completed since there would be no indication or reason to believe that HPCI [s inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.A.3.a and SR 4.5.A.3.b are modified by a note which states that the surveillances are 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 the reactor steam pressure and flow are adequate to perform the test.

With the HPCI system inoperable, adequate core cooling is assured by the operability of the redundant and diversified automatic depressurizatlon system and both the Core Spray and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period of 14 days is based on the verified operability of the RCIC system and the redundant and diversified low pressure core cooling. systems. Verification of RCIC operability may be performed as an administrative check by examining logs or other information to determine if.RCIC is out of service for maintenance or other reasons. It does, not mean to perform the surveillance needed to demonstrate the operability of the RCIC system.

3.5/4.5 BASES 111 Amendment No. 37, 63, 77, 70, Q00, 122

Bases 3.5/4.5 (Continued):

The surveillance requirements provide adequate assurance that the HPCI system will be operable when required. All active components are testable and full flow can be demonstrated by rec*rculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 22000F. ADS is conservatively required to be operable whenever reactor vessel pressure exceeds 150 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.

ADS automatically controls three selected safety-relief valves although the safety analysis only takes credit for two valves. It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability.

B.

RHR Intertie Line An intertie line Is provided to connect the RHR suction line with the two RHR loop return lines. This four-inch line is equipped with three isolation valves. The purpose of this line is to reduce the potential for water hammer in the recirculation and RHR systems. The isolation valves are opened during a cooldown to establish recirculation flow through the RHR suction line and return lines, thereby ensuring a uniform cooldown of this piping. The RHR loop returh line isolation valves receive a closure signal on LPCI initiation. In the event of an inoperable return line isolation valve, there is a potential for some of the LPCI flow to be diverted to the broken loop during a loss of coolant accident. Surveillance requirements have been established to periodically cycle the RHR intertie line isolation valves. In the event of an inoperable RHR loop return line isolation valve, either the inoperable valve is closed or the other two isolation valves are closed to prevent diversion of LPCI flow. The RH.R intertie line flow is not permitted in the Run Mode, to eliminate 1) the need to compensate for the small change in jet pump drive flow or 2) a reduction in core flow during a loss of coolant accident.

C.

Containment Spray/Cooling Systems Two containment spray/cooling subsystems of the RHR system are provided to remove heat energy from the containment and control torus and drywell pressure in the event of a loss of coolant accident. A containment spray/cooling subsystem consists of 1 RHR service water pump, a RHR heat exchanger, 1 RHR pump, and valves and piping necessary for Torus Cooling and Drywell Spray. Torus Spray is not considered part of a containment spray/cooling subsystem. Placing a containment spray/cooling subsystem into operation following a loss of coolant accident is a manual operation.

3.5/4.5 BASES 112 Amendment No. 37, 77-, 79, Q9 "nGn, 102, 422, 124a

Bases 3.5/4.5 (Continued):

The most degraded conditi6n for long term containment heat removal following the design basis loss of coolant accident results from the loss of one diesel generator. Under these conditions% only one RHR pump and one RHR service water pump in the redundant division can be used for containment spray/cooling. The containment temperature and pressure have been analyzed under these conditions assuming service water and initial suppression pool temperature are both 90°F. Acceptable margins to containment design conditions have been demonstrated. Therefore the containment spray/cooling system is more than ample to provide the required heat removal capability. Refer to USAR Sections 5.2.3.3, 6.2.3.2.3, and 8.4.1.3.

During normal plant operation, the containment spray/cooling system provides cooling of the suppression pool water to maintain temperature within the limits specified in Specification 3.7.A.1.

The surveillance requirements and testing in accordance with the IST1,rogram provide adequate assurance that the containment spray/cooling system will be operable when required.

D.

RCIC The RCIC system Is provided to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the feedwater system is not available. The pumping capacity of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation. -if the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts. The system may also be manually initiated at any time.

With the RCIC system inoperable, the HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefore, the RCIC out-of-service period of 14 days is based on the verified operability of the HPCI system and the redundant and diversified low pressure cooling systems. Verification of HPCI operability may be performed as an administrative check by examining logs or other Information to determine if HPCI is out of service for maintenance or other.reasons. It does not mean to perform the surveillance needed to demonstrate the operability of the HPCI system.

The flow tests for the RCIC System are performed at two different pressure ranges such that the system capability to provide rated flowis tested at both thehigherand lower operating ranges of the system. Additionally, adequate steam flow must be passingthrough the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be > 950 psig to perform SR 4.5.D.1.a and s 165 psig to perform SR 4.5.D.1.b. Adequate steam flow is represented by 3.5/4.5 BASES 113 Amendment No. 77, -79, 93, 9", 00a, 22

Bases 3.5/4.5 (Continued):

total steam flow a 106 lb/hr. Reactor startup, and pressure increase to -5165 psig, is allowed prior to performing the low pressure surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the surveillance test is short. Therefore, pressure may be raised above 150 psig, but _ 165-psig to perform this surveillance without entering an LCO for the RCIC System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or.reason to believe that RCIC is inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.D.1.a and SR 4.5.D.1.b are modified by a note which states that the surveillances are 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 the reactor steam pressure and flow are adequate to perform the test.

The surveillance requirements provide adequate assurance that the RCIC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

E.

Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times. It is during refueling outages that major maintenance is performed and during such time that all core and containment spray/cooling subsystems may be out of service.

This specification allows all core and containment spray/cooling subsystems to be Inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the Inspection required by Specification 4.7.A.1. In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.

3.5/4.5 BASES 114 Amendment No. 24,7,79, 977,lOna,-no 2,-422

Bases 3.5/4.5 (Continued):

F.

Recirculation System The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a therrfial-hydraulic instability, a power-flow exclusion region, to be avoided during normal operation, is calculated using the approved methodology as stated in specification 6.7.A.7. Since the exclusion region may change each fuel cycle the limits are contained in the Core Operating Limits Report. Specific directions are provided to avoid operation in this region and to Immediately exit upon an entry. Entries into the exclusion region are not part of normal operation. An entry may occur as the result of an abnormal event such as a single recirculation pump trip. In these events, operation in the exclusion region may be needed to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow biased scram function will suppress oscillations prior to exceeding the fuel safety limit.

Power distribution controls are established to ensure the reactor is operated within the bounds of the stability analysis. With these.controls in place, there is confidence that an oscillation will not occur outside of the stability exclusion region. Without these controls, it is theoretically possible to operate the reactor in such a manner as to cause an oscillation outside of the exclusion region. A nominal 5%

power-flow buffer region outside of the exclusion region is provided to establish a stability margin to the analytically defined exclusion region. The buffer region may. be entered only when the power distribution controls are in place.

Continuous operation with one recirculation loop was analyzed and the adjustments specified in specification 3.5.F.3 were determined by NEDO-24271, June 1980, "Monticello Nuclear Generating Plant Single Loop Operation;"' NEDC-30492, April 1984, "Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant;"

and NEDC-32456P, July 1996. Specification 3.6.A.2 governs the restart of the pump in an idle recirculation loop. Adherence to this specification limits the probability of excessive flux transients and/or thermal stresses.

3.5/4.5 BASES NEXT PAGE IS 121 115 Amendment No.

I

Attachment D Monticello Technical Specification List of Effective Pages and Record of Revision This attachment consists of the current Monticello Technical Specification List of Effective Pages and Record of Revision. The pages included are listed below:

Page A

B C

D E

F G

H J

MONTICELLO NUCLEAR GENERATING PLANT APPENDIX A TECHNICAL SPECIFICATIONS RECORD'OF REVISIONS Amend Paae No.

"A 131

.B

-131 C

115 D

115 E

115 F

115

-G 115 H -

119 i

130 J

131 i

128 ii, 104 iii 120 "iv 128 v

120 vi 121 vii 122 1

119 2

70 3

21 4

102 5

120 5a 120 6

128 7

128 8

128 9

128 10 128 11 128 12 128,.

25a 127

• 25b 127 25c 127 25d 127 26 5

27 81 27a 81 28 128 29 128 30" 103 31 104 32 103 33 103 34 83 35 100a Amend Page No.

36.

128 37 128 38 128 39 129b 40 129b

.42 103 45, 0

46 70 46a 37 47 40 48 89 49 128 50 128 50a 117 51 117 51a 117 52 128' S53 128 54 128 55 103 56 102

.57 70 58-84 58a 29 59 128 59a 103

.60 128 60a 31 60b 62 60c 30 60d.

128 60e 89 61 104 62 117

.63 117 63a 117 64 128 65 117 66 119a 67 117 68 129b 69 129b

.69a 129b 70 117 71 100a Amend Paae No.

71 a 129b 72 104 76 0

77 86 78 0

79 0

80 29 81 3

82 123 82a 63 83 24 83a 24 84 100a 85 100a 86 1 00a 87 100a 88 100a 89 104 90 100a 91 123 92 100a 93 122 94 106 95 77 96 77 97 57 98 56, 99 104 O100 10a 101 122 102 1*2 103 122 104 122 105 122 106 79 107 97 108 128 109 100a 110 lOa 111 130a 112 130a 113 130a 114 130a 115 130a 121 0

Amend Paae No.

122.

.10.6 123 117 124 121 125 104 126 104.

126a 87 127 128 128 42 129 122 130 82 131 122 132 39 132a 122 133 106 134 106 135 106 136 106 137 0

138 100a 145 118a 146 106 147 107 148 117 149 100a 150 128 151 128 153 100a 154 129a 155 122 156 93 157.

130 158 107 159 95 160 95 163 130.

164 104 165 130 166 130 167 112 168 94 169 94 170 130.

171 130.

171a 130 172 71 A

Amendment No. 131.

10/02/02

MONTICELLO NUCLEAR GENERATING PLANt APPENDIX A TECHNICAL SPECIFICATIONS RECORD OF REVISIONS Amend Paae No..

175 107

-175a 117 176 100a 177 130 178 100a 179 123a 180 128a 181 130 182 130 182a 130 183 117 184 100a 185 100a 188 104 189 130 190 130 191 0

192 121 193 121 196 126a 197 121 198 121 199 51 200 129 201 129 202 129 203 41 204 129 204a 129 205 129 206 0

207 123 208 63 209 123 209a 100a 210 1ooa 211 131 212 109 213 99 216 looa 217 128*

218 120

.223 119 224 119 Amend Paae No.

.225 119 226 119 229a 63 229b 104 229c 104 229d 63 229e 122 229u 104 229v 112 229v v 112 229w 112 229ww 112 229x 112 229y 115a 229z 112 230 54 231 34 232 119 233 124 234 119 235 115 236 115 243 128.

244 124 248 59 249 120 250 128 251 124 252 120 253 120

• 254 120 255 120 256 122 257 122 258 122 259 120 260 120 261 120 262 120 B

Amendment No. 131 10/02/02

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NSP Page Revision (REV) No.

Original 1

1/19/71 Note 2 Note 2 Note 2 2

2/20/73 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 3

4 6/17/74 6

8/20/74 Note 3 5

7 8

9 10 7/8/75 12 11 13 14 15 1/22/76 16 2/3/76 17 3/16/76 NOTE 3 18 4/14/76 21 5/20/76 AEC Tech Spec Change Issuance No. and date Note 1 2

1/14/72 3

10/31/72 4

12/8/72 Note 1 5

3/2/73 1

4/28/71&

6 4/3/73 7

5/4/73 8

7/2/73 9

8/24/73 10 10/2/73 11 11/27/73&

12 11/15/73 13 3/30/74 14 5/14/74 Note 1 Note 1 Note 3 10/24/74 15 1/15/75 16 2/3/75 17 2/26/75 18 4/10/75 Note 1 20 9/15/75 19 9/17/75 21 10/6/75 22 10/30/75 C

Amendment No. 115 12/21/00 Major Subject Appendix A Technical Specifications incorporated in DPR-22 on 9/8/70 Removed 5 MWt restriction MOGS Technical Specification changes issued by AEC but never distributed or put into effect, superseded by TS Change 12 11/15/73 RHR service water pump capability change Temporary surveillance test waiver Increase in U-235 allowed In fission chambers Miscellaneous Technical Specification changes, Respiratory Protection, & Administrative Control Changes Respiratory Protection Changes Relief Valve and CRD Scram Time Changes Fuel Densification Umits Safety Valve Setpoint Change Offgas Holdup System, RWM, and Miscellaneous Changes 8x8 Fuel Load Authorization 8x8 Full Power authorization Changed byproduct material allowance Changed byproduct material allowance Inverted Tube (CRD) Limits REMP Changes Reactor Vessel Surveillance Program Changes Vacuum Breaker Test Changes Corrects Errors & Provides Clarification Increased allowed quantity of U-235 Snubber Requirements Removed byproduct material allowance Suppression Pool Temperature Limits Appendix K and GETAB Limits Reporting Requirements CRD Collet Failure Surveillance NSP Organization Changes Adoption of GETAB Containment Isolation Valve Testing Interim Appendix B, Section,,.4 Tech. Specs.

NOTE 4 4/13/76 I

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Revision (REV) No.

26 27 28 29 30

31.

32 33 34 35 36 37 38 39 NOTE 1 40 41 "42 43 44 45 46 47 48 49 50 NOTE 1 NOTE 1 51 52 NOTE 1 License DPR-22 Amend No. & Date 19 5/27/76 20 6/18/76 22 7/13/76 23 9/27/76 24 10/15/76 25 10/27/76 26 4/1/77 27 5/24/77 28 6/10/77 29 9/16/77 30 9/28/77 31 10/14/77 32 12/9/77 33 1/25/78 34 4/14/78 35 9/15/78 36 10/30/78 37 11/6/78 NOTE 3 11/24/78 38 3/15/79 39 5/15/79 40 6/5/79 41 8/29/79 42 12/28/79 43 2/12/80 44 2/29/80 8/29/80 9/19/80 10/24/80 1/9/81 1/9/81 I

D Amendment No. 115 12/21/00 Major Subject Low Steamline Pressure Setpoint and MCPR Changes.

APLHGR, LHGR, MCPR Limits Correction of Errors and Environmental Reporting Standby Gas Treatment System Surveillance CRD Test Frequency Snubber Testing Changes APRS Test Method MAPLHGR Clamp at Reduced Flow Radiation Protection Supervisor Qualification REMP Changes More Restrictive MCPR Inservice Inspection Changes Reporting Requirements Fire Protection Requirements Increase in spent fuel storage capacity RPT Requirements Suppression Pool Surveillance 8x8RAuthorization, MCPR Limits &SRV Setpoints Corrected Downcomer Submergence Incorporation of Physical Security Plan into License Revised LPCI Flow Capability Respiratory Protection Program Changes Fire Protection Safety.Evaluation Report MAPLHGR vs. Exposure Table MCPR & MAPLHGR Changes for Cycle 8 and Extended Core Burnup ILRT Requirements Order for Modification of License-Environmental Qualification Revised Order for Modification of License-Environmental Qualification Order for Modification.of License-Environmental Qualification Records Issuance of Facility Operating License (FTOL)

Order for Modification of License Concerning BWR Scram Discharge Systems (License conditions removed per Amendment No. 11 dated 10/8/82)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 1/13/81 1

2/12/81 2

3/2/81 "NSP Revision (REM No.

NOTE 1 53 54 NOTE 1 55 56 57 58 59 NOTE 1 3/27/81 3/27/81 5

5/4/81 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 6/3/81 6/30/81 11/5/81 12/28/81 1/19/82 5/20/82 10/8/82 11/30/82 12/6/82 12/10/82 12/17/82 4/18/83 4/17/83 11/28/83 12/30/83 1/16/84 1/23/84 2/2/84 4/3/84 5/1/84 8/15/84 9/24/84 10/31/84 11/2/84 11/16/84 11/16/84 11/27/84 3

4 E

Amendment No. 115 12/21/00 Major Subject Order for Modification Mark I Containment Revision of License Conditions Relating to Fire Protection Modifications TMI Lessons Learned & Safety Related Hydraulic Snubber Additions Low voltage protection, organization and miscellaneous Incorporation of Safeguards Contingency Plan and Security Force Qualification and Training Plan into License Cycle 9 - ODYN Changes, New MAPLHGR's, RPS Response time change Inservice Inspection Program Fire Protection Technical Specification Changes Mark I Containment Modifications Inservice Surveillance Requirements for Snubbers Revised Order for Modification Mark I Containment Scram Discharge Volume New Scram Discharge Volumes RPS Power Monitor Cycle;10 Recirc Piping and Coolant Leak Detection Appendix I Technical Specifications (removed App. B)

Organizational Changes Miscellaneous Changes Steam Line Temperature Switch Setpoint Radiatibn Protection Program SRM Count Rate Definition of Operability Miscellaneous Technical Specification Changes RPS Electrical Protection Assembly Time Delay Scram. Discharge Volume Vent and Drain Valves Miscellaneous Technical Specification Changes Cycle 11 RHR Intertie Une Addition Hybrid I Control Rod Assembly ARTS Low Low Set Logic Degraded Voltage Protection Logic 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Revision (RE No.

82 83 84 85 86 87 88 89 9o 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 114 115 116 License DPR-22 Amend No. & Date 32 5/28/85 33 10/7/85 34 10/8/85 35 12/3/85 36 12/23/85 37 1/22/86 38 2/12/86 39 3/13/86 40 3/18/86 41 3/24/86 42 3/27/86 43 4/8/86

.44 5/27/86 45 7/1/86 46 7/1/86 47 10/22/86 48 12/1/86 49 8/26/87 50 8/26/87 51 10/16/87 52 11/18/87 53 11/19/87 54 11/25/87 55 11/25/87 56 12/11/87 57 9/23/88 58 12/13/88 59 2/16/89 60 2/28/89 61 3/29/89 62 3/31/89 63 4/18/89 64 5/10/89 65 5/30/89 66 5/30/89 Major Subject Surveillance Requirements Screen Wash/Fire Pump (Partial)

Fuel Enrichment Limits Combustible Gas Control System Vacuum Breaker Cycling NUREG-0737 Technical Specifications Environmental Technical Specifications Administrative Changes Clarification of Radiation Monitor Requirements 250 Volt Battery Jet Pump Surveillance Simmer Margin Improvement Cycle 12 Operation Miscellaneous Changes LER Reporting and Miscellaneous Changes Single Loop Operation Offgas System Trip Rod Block Monitor APRM and IRM Scram Requirements 2R Transformer Surveillance Intervals - ILRT Schedule Extension of Operating License Cycle 13 and Misc Changes Appendix J Testing.

ATWS - Enriched Boron Increased Boron Enrichment Physical Security Plan Miscellaneous Administrative Changes Miscellaneous Administrative Changes Fire Protection and Detection System ADS Logic and S/RV Discharge Pipe Pressure Miscellaneous Technical Specification Improvements Containment Vent and Purge Valves NUREG-0737 - Generic Letter 83-36 Reactor Vessel Level Instrumentation I

F Amendment No. 115 12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS "N NSP Revision (REVM No.

11.7.

118 119 120 121 122 123 124 125 126 127 128 125 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 146 147 License DPR-22 Amend No. & Date 67 6/19/89

.68 7/14/89 69 9/12/89 70 9/28/89 71 10/19/89 72 11/2/89 73 5/1/90 74 6/5/90 75 10/12/90 76 12/20/90 77 2/15/91 78 3/28/91 79 4/9/91 80 8/12/91 81 4/16/92 82 7/15/92 83 8/18/92 84 1/27/93 85 6/29/93 86 7/12/93 87 4/15/94 88 6/30/94 89 8/25/94 90 91 92 93 94 95 96 97 9/7/94 9/9/94 9/15/94 7/12/95 10/2/95 4/3/96 4/9/96 9/17/96 Major Subiect Extension of MAPLHGR. Exposure for One Fuel Type SRO Requirements & Organization Chart Removal Operations Committee, Quorum Requirements Relocation of Cycle-Specific Thermal-Hydraulic Limits Deletion of Primary Containment Isolation Valve Table RG 1.99, Rev 2, ISI & ILRT Combined STNLSO Position Removal of WRGM Automatic ESF Actuation Diesel Fuel Oil Storage Miscellaneous Administrative Changes Redundant and IST Testing Alarming Dosimetry SAFER/GESTR Torus Vacuum Breaker Test Switch/EDG Fuel Oil Tank Level Surveillance Test Interval Extension - Part I Alternate Snubber Visual Inspection Intervals Revisions to Reactor Protection System Tech Specs MELLIA and Increase Core Flow Revision to Diesel Fire Pump Fuel Oil Sampling Requirements Revisions to Control Rod Drive Testing Requirements Revised Coolant Leakage Monitoring Frequency Average Planar Linear Heat Generation Rate (APLHGR)

Specification & Minimum Critical Power Ratio Bases Revisions Removal of Chlorine Detection Requirements and Changes to Control Room Ventilation. System Requirements Revisions to Radiological Effluent Specifications Secondary Containment System and Standby Gas Treatment System Water Level Setpoint Change Change in Safety Relief Valves Testing Requirements Revised Core Spray Pump Flow Standby Gas Treatment and Secondary Containment Systems MSIV Combined Leakrate, and Appendix J, Option B Purge and Vent Valve Seal Replacement Interval.

Implementation of BRWOG Option I-D core Stability Solution and re-issue of pages 11, 12, 82 and 231 to reflect pages issued by NRC amendments.

I G

Amendment No. 115 12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 98 7/25/97 99 10/29/97 11/25/97 NSP Revision (REV) No.

148 149 NOTE 5 150 NOTE 6 4/20/98 4/30/98 8/28/98 102 9/16/98 103 12/23/98 104 105 106 107 108 109 110 111 12/24/98 03/19/99 10/12/99 11/24/99 12/8/99 2/16/00 8/7/00 8/18/00 112 8/18/00 113 10/2/00 114 11/30/00 115 11 5a 12/21/00 02/13/01 116 03/01/01 117 03/07/01 118 03/09/01 118a 05/10/01 Major Subject Bases changes on containment overpressure and number of RHR pumps required to be operable. Reissue pages 207, 209, 219, 229k, 229p, 230, 245 to reflect pages issued by NRC amendments.

SLMCPR for Cycle 18 and reissue pages vi, 155, 202, 207, 219, 229u Reissue pages a, b, g, iii, vi, 14, 25a, 155, 198y, 198z, 202, 207, 209, 219, 229k, 229p, 229r, 229u, 230, 245 SLMCPR for Cycle 19 Reissue all pages.

Reactor Coolant Equivalent Radioiodine Concentration and Control Room Habitability Monticello Power Rerate Surveillance Test Interval/Allowed Outage Time Extension Program - Part 2 Revision of Statement on Shift Length & other Misc Changes CST Low Level HPCI/RCIC Suction Transfer Revised RPV-PT Curves & remove SBLC RV setpoint Reactor Pressure Vessel Hydrostatic and Leakage Testing Testing Requirements for Control Room EFT Filters Safety Limit Minimum Critical Power Ratio for Cycle 20 Transfer of Operating Authority from NSP to NMC Transfer of Operating License from NSP to a New Utility Operating Company Emergency Filtration Train Testing Exceptions and Technical Specification Revisions Alternate Shutdown System Operability Requirements Safety/Relief Valve Bellows Leak Detection System Test Frequency Administrative Controls and Other Miscellaneous Changes Bases Change to Reflect Modification 98Q145 Installed Control Room Toxic Gas Air Supply Relocation of Inservice Inspection Requirements to a Licensee Program Reactor Water Cleanup (RWCU) System Automatic Isolation and Miscellaneous Instrumentation System Changes Revision of Standby Liquid Control System Surveillance Requirements Bases Change - 50'F Loop Temperature, Bus Transfer &

Rerate Correction H

Amendment No. 119 04/05/01 100 1 00a 101

MONTICELLO NUCLEAR GENERATING PLANT.

RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Revision (REV1 No.

125 126 126a 12/06/01 01/18/02 02/15/02 127 05/31/02 128 06/11/02 128a 07/11/02 129 08/27/02 129a 129b 130 130a 09/12/02 09/12/02 09/23/02 09/26/02 License DPR-22 Amend No. & Date 119 04/05/01 119a 06/28/01 120 07/24/01 121 07/25/01 122 08/01/01 122a 10/22/01 123 10/26/01 123a 10/25/01 124 10/30/01 124a 12/05/01 Amendment No. 130a 09/26/02 Major Subject Fire Protection Technical Specification Changes Bases Change - Added information on cooldown rate Relocation of Radiological Effluent Technical Specifications to a Licensee-Controlled Program Clarify air ejector offgas activity sample point and operability requirements Relocation of Inservice Testing Requirements to a Licensee-Controlled Program Bases Change - Remove scram setpoints sentence and correct typo Control Rod Drive and Core Monitoring Technical Specification Changes Bases Change - Change to reflect new operation of drywell to suppression chamber vacuum breaker valve position indicating lights Relocation of Technical Specification Administrative Controls Related to Quality Assurance Plan Bases Change - Change to reflect revised Technical Specification definition of a containment spray/cooling subsysýem Safety Limit Minimum Critical Power Ratio for Cycle 21 Elimination of Local Suppression Pool Temperature Limits Bases Change - Change reflects relocation of sample point for the offgas radiation monitor Missed Surveillance Requirement Technical Specification Changes Changes to the Technical Specifications Revised Reference Point for Reactor Vessel Level Setpoints, Simplification of Safety Limits, and Improvement to the Bases Bases Change - Correct Drywell to Suppression Chamber Vacuum Breaker Indicating Light Description Revise Technical Specifications and Surveillance Requirements Relating to Standby Diesel Generators Bases Change - Change to Snubber Operability Description Bases Change - Remove Language That Implies Trip Settings Can Be Modified By Deviation Values Containment Systems Technical Specification Changes Bases Change - HPCI - Change Wording / HPCI & RCIC Enhance with Wording Consistent with NUREG-1 433-Rev 1 I

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Revision (REV) No.

License DPR-22 Amend No. & Date 131 10/02/02 Maior-Subiect Update the Multiplier Values for Single Loop.Operation Average:l Planar Unear Heat Generation Rate (APLHGR) 0

1.

License Amendment or Order for Modification of License not affecting Technical Specifications.

2.

Technical Specification change issued prior to 10 CFR revisions which require issuance of Technical Specification changes as License Amendments.

3.

Modification to Bases. No Technical Specification change or License Amendment issued.

4.

Technical Specification change numbers no longer assigned beginning with Amendment 15.

5.

Pages reissued 11/25/97 to conform with NRC version. Revision number of effected pages not changed.

6.

All pages reissued using INTERLEAF in different font. Using NRC Amendment Nos. and issue date. For Bases and Table of Contents, spelling errors corrected and editorial corrections made and all Amendment Nos. changed to 1 00a. For remaining Tech Spec pages, no other changes made and current Amendment Nos. used.

J Amendment No. 131 10/02/02