ML030030480

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Technical Specification Bases Pages
ML030030480
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 12/19/2002
From: Forbes J
Nuclear Management Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML030030480 (25)
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NMC* Monticello Nuclear Generating Plant Committed to NudearElxcenle* 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 _5165 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 beoerformed 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 22000 F. 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 90 0F.. 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-all8tfor anoperability 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 ->950psig 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 steamtest 2 106 lb/hr. Reactor startup, and pressure increase to ! 165 psig, isallowed prior to performing the low pressure flowbecause surveillance 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.AGE 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 2200 0F. 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 -5165psig, 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,-no2,-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

  • Amend .- Amend Amend Paae No. Page No. Paae No. Paae No.

"A 131 36. 128 71 a 129b 122. .10.6

.B -131 37 128 72 104 123 117 C 115 38 128 76 0 124 121 D 115 39 129b 77 86 125 104 E 115 40 129b 78 0 126 104.

F 115 .42 103 79 0 126a 87

-G 115 45, 0 80 29 127 128 H- 119 46 70 81 3 128 42 i 130 46a 37 82 123 129 122 J 131 47 40 82a 63 130 82 i 128 48 89 83 24 131 122 ii, 104 49 128 83a 24 132 39 iii 120 50 128 84 100a 132a 122 "iv 128 50a 117 85 100a 133 106 v 120 51 117 86 1 00a 134 106 vi 121 51a 117 87 100a 135 106 vii 122 52 128' 88 100a 136 106 1 119 S53 128 89 104 137 0 2 70 54 128 90 100a 138 100a 3 21 55 103 91 123 145 118a 4 102 56 102 92 100a 146 106 5 120 .57 70 93 122 147 107 5a 120 58- 84 94 106 148 117 6 128 58a 29 95 77 149 100a 7 128 59 128 96 77 150 128 8 128 59a 103 97 57 151 128 9 128 .60 128 98 56, 153 100a 10 128 60a 31 99 104 154 129a 11 128 60b 62 O100 10a 155 122 12 128,. 60c 30 101 122 156 93 25a 127 60d. 128 102 1*2 157. 130

  • 25b 127 60e 89 103 122 158 107 25c 127 61 104 104 122 159 95 25d 127 62 117 105 122 160 95 26 5 .63 117 106 79 163 130.

27 81 63a 117 107 97 164 104 27a 81 64 128 108 128 165 130 28 128 65 117 109 100a 166 130 29 128 66 119a 110 lOa 167 112 30" 103 67 117 111 130a 168 94 31 104 68 129b 112 130a 169 94 32 103 69 129b 113 130a 170 130.

33 103 .69a 129b 114 130a 171 130.

34 83 70 117 115 130a 171a 130 35 100a 71 100a 121 0 172 71 A

Amendment No. 131.

10/02/02

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

175 107 .225 119

-175a 117 226 119 176 100a 229a 63 177 130 229b 104 178 100a 229c 104 179 123a 229d 63 180 128a 229e 122 181 130 229u 104 182 130 229v 112 182a 130 229v v 112 183 117 229w 112 184 100a 229ww 112 185 100a 229x 112 188 104 229y 115a 189 130 229z 112 190 130 230 54 191 0 231 34 192 121 232 119 193 121 233 124 196 126a 234 119 197 121 235 115 198 121 236 115 199 51 243 128.

200 129 244 124 201 129 248 59 202 129 249 120 203 41 250 128 204 129 251 124 204a 129 252 120 205 129 253 120 206 0 *254 120 207 123 255 120 208 63 256 122 209 123 257 122 209a 100a 258 122 210 1ooa 259 120 211 131 260 120 212 109 261 120 213 99 262 120 216 looa 217 128*

218 120

.223 119 224 119 B

Amendment No. 131 10/02/02

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License AEC Tech Spec Revision DPR-22 Change Issuance (REV) No. Amend No. & Date No. and date Major Subject Original Appendix A Technical Specifications incorporated in DPR-22 on 9/8/70 1 1/19/71 Note 1 Removed 5 MWt restriction Note 2 2 1/14/72 MOGS Technical Specification changes issued by AEC but never distributed or put into effect, superseded by TS Change 12 11/15/73 1 Note 2 3 10/31/72 RHR service water pump capability change Note 2 4 12/8/72 Temporary surveillance test waiver 2 2/20/73 Note 1 Increase in U-235 allowed In fission chambers 2 Note 2 5 3/2/73 Miscellaneous Technical Specification changes, 3 Note 2 1 4/28/71& Respiratory Protection, & Administrative Control 6 4/3/73 Changes 4 Note 2 7 5/4/73 Respiratory Protection Changes 5 Note 2 8 7/2/73 Relief Valve and CRD Scram Time Changes 6 Note 2 9 8/24/73 Fuel Densification Umits 7 Note 2 10 10/2/73 Safety Valve Setpoint Change 8 Note 2 11 11/27/73& Offgas Holdup System, RWM, and 12 11/15/73 Miscellaneous Changes 9 Note 2 13 3/30/74 8x8 Fuel Load Authorization 10 3 14 5/14/74 8x8 Full Power authorization 4 6/17/74 Note 1 Changed byproduct material allowance 6 8/20/74 Note 1 Changed byproduct material allowance 11 Note 3 Note 3 10/24/74 Inverted Tube (CRD) Limits 12 5 15 1/15/75 REMP Changes 13 7 16 2/3/75 Reactor Vessel Surveillance Program Changes 14 8 17 2/26/75 Vacuum Breaker Test Changes 15 9 18 4/10/75 Corrects Errors & Provides Clarification 10 7/8/75 Note 1 Increased allowed quantity of U-235 16 12

  • 20 9/15/75 Snubber Requirements 17 11 19 9/17/75 Removed byproduct material allowance 18 13 21 10/6/75 Suppression Pool Temperature Limits 19 14 22 10/30/75 Appendix K and GETAB Limits 20 15 1/22/76 NOTE 4 Reporting Requirements 21 16 2/3/76 CRD Collet Failure Surveillance 22 17 3/16/76 NSP Organization Changes 23 NOTE 3 4/13/76 Adoption of GETAB 24 18 4/14/76 Containment Isolation Valve Testing 25 21 5/20/76 Interim Appendix B, Section,,.4 Tech. Specs.

C Amendment No. 115 I

12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (REV) No. Amend No. & Date Major Subject 26 19 5/27/76 Low Steamline Pressure Setpoint and MCPR Changes.

27 20 6/18/76 APLHGR, LHGR, MCPR Limits 28 22 7/13/76 Correction of Errors and Environmental Reporting 29 23 9/27/76 Standby Gas Treatment System Surveillance 30 24 10/15/76 CRD Test Frequency

31. 25 10/27/76 Snubber Testing Changes 32 26 4/1/77 APRS Test Method 33 27 5/24/77 MAPLHGR Clamp at Reduced Flow 34 28 6/10/77 Radiation Protection Supervisor Qualification 35 29 9/16/77 REMP Changes 36 30 9/28/77 More Restrictive MCPR 37 31 10/14/77 Inservice Inspection Changes 38 32 12/9/77 Reporting Requirements 39 33 1/25/78 Fire Protection Requirements NOTE 1 34 4/14/78 Increase in spent fuel storage capacity 40 35 9/15/78 RPT Requirements 41 36 10/30/78 Suppression Pool Surveillance "42 37 11/6/78 8x8RAuthorization, MCPR Limits &SRV Setpoints 43 NOTE 3 11/24/78 Corrected Downcomer Submergence 44 38 3/15/79 Incorporation of Physical Security Plan into License 45 39 5/15/79 Revised LPCI Flow Capability 46 40 6/5/79 Respiratory Protection Program Changes 47 41 8/29/79 Fire Protection Safety .Evaluation Report 48 42 12/28/79 MAPLHGR vs. Exposure Table 49 43 2/12/80 MCPR & MAPLHGR Changes for Cycle 8 and Extended Core Burnup 50 44 2/29/80 ILRT Requirements NOTE 1 - 8/29/80 Order for Modification of License-Environmental Qualification NOTE 1 - 9/19/80 Revised Order for Modification of License-Environmental Qualification 51 - 10/24/80 Order for Modification .of License-Environmental Qualification Records 52 - 1/9/81 Issuance of Facility Operating License (FTOL)

NOTE 1 - 1/9/81 Order for Modification of License Concerning BWR Scram Discharge Systems (License conditions removed per Amendment No. 11 dated 10/8/82)

D I Amendment No. 115 12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS "NSP License Revision DPR-22 (REM No. Amend No. & Date Major Subject NOTE 1 - 1/13/81 Order for Modification Mark I Containment 1 2/12/81 Revision of License Conditions Relating to Fire Protection Modifications 53 2 3/2/81 TMI Lessons Learned & Safety Related Hydraulic Snubber Additions 54 3 3/27/81 Low voltage protection, organization and miscellaneous NOTE 1 4 3/27/81 Incorporation of Safeguards Contingency Plan and Security Force Qualification and Training Plan into License 55 5 5/4/81 Cycle 9 - ODYN Changes, New MAPLHGR's, RPS Response time change 56 6 6/3/81 Inservice Inspection Program 57 7 6/30/81 Fire Protection Technical Specification Changes 58 8 11/5/81 Mark I Containment Modifications 59 9 12/28/81 Inservice Surveillance Requirements for Snubbers NOTE 1 1/19/82 Revised Order for Modification Mark I Containment 60 10 5/20/82 Scram Discharge Volume 61 11 10/8/82 New Scram Discharge Volumes 62 12 11/30/82 RPS Power Monitor 63 13 12/6/82 Cycle;10 64 14 12/10/82 Recirc Piping and Coolant Leak Detection 65 15 12/17/82 Appendix I Technical Specifications (removed App. B) 66 16 4/18/83 Organizational Changes 67 17 4/17/83 Miscellaneous Changes 68 18 11/28/83 Steam Line Temperature Switch Setpoint 69 19 12/30/83 Radiatibn Protection Program 70 20 1/16/84 SRM Count Rate 71 21 1/23/84 Definition of Operability 72 22 2/2/84 Miscellaneous Technical Specification Changes 73 23 4/3/84 RPS Electrical Protection Assembly Time Delay 74 24 5/1/84 Scram. Discharge Volume Vent and Drain Valves 75 25 8/15/84 Miscellaneous Technical Specification Changes 76 26 9/24/84 Cycle 11 77 27 10/31/84 RHR Intertie Une Addition 78 28 11/2/84 Hybrid I Control Rod Assembly 79 29 11/16/84 ARTS 80 30 11/16/84 Low Low Set Logic 81 31 11/27/84 Degraded Voltage Protection Logic E

Amendment No. 115 12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (RE No. Amend No. & Date Major Subject 82 32 5/28/85 Surveillance Requirements 83 33 10/7/85 Screen Wash/Fire Pump (Partial) 84 34 10/8/85 Fuel Enrichment Limits 85 35 12/3/85 Combustible Gas Control System 86 36 12/23/85 Vacuum Breaker Cycling 87 37 1/22/86 NUREG-0737 Technical Specifications 88 38 2/12/86 Environmental Technical Specifications 89 39 3/13/86 Administrative Changes 9o 40 3/18/86 Clarification of Radiation Monitor Requirements 91 41 3/24/86 250 Volt Battery 92 42 3/27/86 Jet Pump Surveillance 93 43 4/8/86 Simmer Margin Improvement 94 .44 5/27/86 Cycle 12 Operation 95 45 7/1/86 Miscellaneous Changes 96 46 7/1/86 LER Reporting and Miscellaneous Changes 97 47 10/22/86 Single Loop Operation

" 98 48 12/1/86 Offgas System Trip 99 49 8/26/87 Rod Block Monitor 100 50 8/26/87 APRM and IRM Scram Requirements 101 51 10/16/87 2R Transformer 102 52 11/18/87 Surveillance Intervals - ILRT Schedule 103 53 11/19/87 Extension of Operating License 104 54 11/25/87 Cycle 13 and Misc Changes 105 55 11/25/87 Appendix J Testing.

106 56 12/11/87 ATWS - Enriched Boron 107 57 9/23/88 Increased Boron Enrichment 108 58 12/13/88 Physical Security Plan 109 59 2/16/89 Miscellaneous Administrative Changes 110 60 2/28/89 Miscellaneous Administrative Changes 111 61 3/29/89 Fire Protection and Detection System 112 62 3/31/89 ADS Logic and S/RV Discharge Pipe Pressure 113 63 4/18/89 Miscellaneous Technical Specification Improvements 114 64 5/10/89 Containment Vent and Purge Valves 115 65 5/30/89 NUREG-0737 - Generic Letter 83-36 116 66 5/30/89 Reactor Vessel Level Instrumentation F I Amendment No. 115 12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS "N NSP License Revision DPR-22 (REVM No. Amend No. & Date Major Subiect 11.7. 67 6/19/89 Extension of MAPLHGR. Exposure for One Fuel Type 118 .68 7/14/89 SRO Requirements & Organization Chart Removal 119 69 9/12/89 Operations Committee, Quorum Requirements 120 70 9/28/89 Relocation of Cycle-Specific Thermal-Hydraulic Limits 121 71 10/19/89 Deletion of Primary Containment Isolation Valve Table 122 72 11/2/89 RG 1.99, Rev 2, ISI & ILRT 123 73 5/1/90 Combined STNLSO Position 124 74 6/5/90 Removal of WRGM Automatic ESF Actuation 125 75 10/12/90 Diesel Fuel Oil Storage 126 76 12/20/90 Miscellaneous Administrative Changes 127 77 2/15/91 Redundant and IST Testing 128 78 3/28/91 Alarming Dosimetry 125 79 4/9/91 SAFER/GESTR 130 80 8/12/91 Torus Vacuum Breaker Test Switch/EDG Fuel Oil Tank Level 131 81 4/16/92 Surveillance Test Interval Extension - Part I 132 82 7/15/92 Alternate Snubber Visual Inspection Intervals 133 83 8/18/92 Revisions to Reactor Protection System Tech Specs 134 84 1/27/93 MELLIA and Increase Core Flow 135 85 6/29/93 Revision to Diesel Fire Pump Fuel Oil Sampling Requirements 136 86 7/12/93 Revisions to Control Rod Drive Testing Requirements 137 87 4/15/94 Revised Coolant Leakage Monitoring Frequency 138 88 6/30/94 Average Planar Linear Heat Generation Rate (APLHGR)

Specification & Minimum Critical Power Ratio Bases Revisions 139 89 8/25/94 Removal of Chlorine Detection Requirements and Changes to Control Room Ventilation. System Requirements 140 90 9/7/94 Revisions to Radiological Effluent Specifications 141 91 9/9/94 Secondary Containment System and Standby Gas Treatment System Water Level Setpoint Change 142 92 9/15/94 Change in Safety Relief Valves Testing Requirements 143 93 7/12/95 Revised Core Spray Pump Flow 144 94 10/2/95 Standby Gas Treatment and Secondary Containment Systems 145 95 4/3/96 MSIV Combined Leakrate, and Appendix J, Option B 146 96 4/9/96 Purge and Vent Valve Seal Replacement Interval.

147 97 9/17/96 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.

G I Amendment No. 115 12/21/00

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (REV) No. Amend No. & Date Major Subject 148 98 7/25/97 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.

149 99 10/29/97 SLMCPR for Cycle 18 and reissue pages vi, 155, 202, 207, 219, 229u NOTE 5 11/25/97 Reissue pages a, b, g, iii, vi, 14, 25a, 155, 198y, 198z, 202, 207, 209, 219, 229k, 229p, 229r, 229u, 230, 245 150 100 4/20/98 SLMCPR for Cycle 19 NOTE 6 100a 4/30/98 Reissue all pages.

101 8/28/98 Reactor Coolant Equivalent Radioiodine Concentration and Control Room Habitability 102 9/16/98 Monticello Power Rerate 103 12/23/98 Surveillance Test Interval/Allowed Outage Time Extension Program - Part 2 104 12/24/98 Revision of Statement on Shift Length & other Misc Changes 105 03/19/99 CST Low Level HPCI/RCIC Suction Transfer 106 10/12/99 Revised RPV-PT Curves & remove SBLC RV setpoint 107 11/24/99 Reactor Pressure Vessel Hydrostatic and Leakage Testing 108 12/8/99 Testing Requirements for Control Room EFT Filters 109 2/16/00 Safety Limit Minimum Critical Power Ratio for Cycle 20 110 8/7/00 Transfer of Operating Authority from NSP to NMC 111 8/18/00 Transfer of Operating License from NSP to a New Utility Operating Company 112 8/18/00 Emergency Filtration Train Testing Exceptions and Technical Specification Revisions 113 10/2/00 Alternate Shutdown System Operability Requirements 114 11/30/00 Safety/Relief Valve Bellows Leak Detection System Test Frequency 115 12/21/00 Administrative Controls and Other Miscellaneous Changes 11 5a 02/13/01 Bases Change to Reflect Modification 98Q145 Installed Control Room Toxic Gas Air Supply 116 03/01/01 Relocation of Inservice Inspection Requirements to a Licensee Program 117 03/07/01 Reactor Water Cleanup (RWCU) System Automatic Isolation and Miscellaneous Instrumentation System Changes 118 03/09/01 Revision of Standby Liquid Control System Surveillance Requirements 118a 05/10/01 Bases Change - 50'F Loop Temperature, Bus Transfer &

Rerate Correction H

Amendment No. 119 04/05/01

MONTICELLO NUCLEAR GENERATING PLANT.

RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (REV1 No. Amend No. & Date Major Subject 119 04/05/01 Fire Protection Technical Specification Changes 119a 06/28/01 Bases Change - Added information on cooldown rate 120 07/24/01 Relocation of Radiological Effluent Technical Specifications to a Licensee-Controlled Program 121 07/25/01 Clarify air ejector offgas activity sample point and operability requirements 122 08/01/01 Relocation of Inservice Testing Requirements to a Licensee-Controlled Program 122a 10/22/01 Bases Change - Remove scram setpoints sentence and correct typo 123 10/26/01 Control Rod Drive and Core Monitoring Technical Specification Changes 123a 10/25/01 Bases Change - Change to reflect new operation of drywell to suppression chamber vacuum breaker valve position indicating lights 124 10/30/01 Relocation of Technical Specification Administrative Controls Related to Quality Assurance Plan 124a 12/05/01 Bases Change - Change to reflect revised Technical Specification definition of a containment spray/cooling subsysýem 125 12/06/01 Safety Limit Minimum Critical Power Ratio for Cycle 21 126 01/18/02 Elimination of Local Suppression Pool Temperature Limits 126a 02/15/02 Bases Change - Change reflects relocation of sample point for the offgas radiation monitor 127 05/31/02 Missed Surveillance Requirement Technical Specification Changes 128 06/11/02 Changes to the Technical Specifications Revised Reference Point for Reactor Vessel Level Setpoints, Simplification of Safety Limits, and Improvement to the Bases 128a 07/11/02 Bases Change - Correct Drywell to Suppression Chamber Vacuum Breaker Indicating Light Description 129 08/27/02 Revise Technical Specifications and Surveillance Requirements Relating to Standby Diesel Generators 129a 09/12/02 Bases Change - Change to Snubber Operability Description 129b 09/12/02 Bases Change - Remove Language That Implies Trip Settings Can Be Modified By Deviation Values I

130 09/23/02 Containment Systems Technical Specification Changes 130a 09/26/02 Bases Change - HPCI - Change Wording / HPCI & RCIC Enhance with Wording Consistent with NUREG-1 433-Rev 1 Amendment No. 130a 09/26/02

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS

  • Revision DPR-22 (REV) No. Amend No. & Date Maior-Subiect 131 10/02/02 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

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