L-MT-03-085, Technical Specification Bases Pages Changes

Technical Specification Bases Pages Changes

ML033320403
Person / Time
Site:	Monticello
Issue date:	11/11/2003
From:	Thomas J. Palmisano Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-03-085
Download: ML033320403 (28)

Text

Monticello Nuclear Generating Plant Committed to Nuclear Exlce Operated by Nuclear Management Company, LLC November 11, 2003 L-MT-03-085 Technical Specification 6.8.K U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket No. 50-263 License No. DPR-22 Technical Specification Bases Pacies 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" or "b". The changes are summarized in Enclosure 1. Marked up pages applicable at the time the changes were made are provided in Enclosure 2. A final typed copy of the changed pages that are applicable, for entry into the NRC authority copy, are provided in . The current copy of our list of effective pages and record of revision is attached for your information, as Enclosure 4.

Please contact John Fields at 763-295-1663 with any questions.

Thomas J. Palmisano Site Vice President, Monticello Nuclear Generating Plant Nuclear Management Company, LLC - Summary of Technical Specification Bases Changes (TSBC) - Monticello Technical Specification Bases Pages Marked Up With Changes - Revised Monticello Technical Specification Bases Pages - Monticello Technical Specification List of Effective Pages and Record of Revision cc: Administrator, Region ll, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce I

2807 West County Road 75

• Monticello, Minnesota 55362-9637 Telephone: 763.295.5151

• Fax: 763.295.1454

-No

ENCLOSURE 1

SUMMARY

OF TECHNICAL SPECIFICATION BASES CHANGE (TSBC)

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

TSBC-1 36a Technical Specification Involved - 3.7 Page affected - 182a Summary of Change: This TSBC adds language to the Technical Specification Bases that defines the abbreviation EFCV in the bases page as Excess Flow Check Valve.

TSBC-1 37a Technical Specification Involved - 3.6.4. D Pages affected - 150, 151, 152, 152a, 152b Summary of Change: This TSBC modifies the Technical Specification bases for changes made under License Amendment 137 which revised the technical specifications to be similar to alter Reactor Coolant System (RCS) leakage detection requirements and to increase operational flexibility due to the failure of RCS leakage detection equipment.

TSBC-137b Technical Specification Involved - 3.13 Page affected - 225 Summary of Change: This TSBC adds language to the Technical Specification Bases that states that the "System controls on the Alternate Shutdown System (ASDS) panel" refers to portions of control circuits and instrumentation necessary solely to support ASDS functions, including equipment not located within the ASDS panel. The ASDS Limiting Condition for Operation is entered for any of the following conditions: (1) 12 Residual Heat Removal Service Water pump inoperable and (2) System controls on the ASDS panel inoperable.

Page 1 of 1

i 7 ENCLOSURE 2 MONTICELLO TECHNICAL SPECIFICATION BASES PAGES MARKED UP WITH CHANGES This attachment consists of Monticello Technical Specification bases page marked up with changes. The pages included are listed below:

Page 150 151 152 152a 152b 182a 225 6 pages follow

I-Bases 3.6/4.6 (Continued):

D. GeelapA-Leakage Reactor Coolant System (RCS)

1. RCS Operational Leakage The allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed'behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be Pressure Boundary Leakage and they cannot be reduced within the allowed times, the reactor will be shutdown to allow further investigation and corrective action.

The low limit on increase in unidentified leakage assumes a failure mechanism of Intergranular Stress Corrosion Cracking (IGSCC) that produces tight cracks. This flow increase limit is capable of providing an early warning of such deterioration.

No applicable safety analysis assumes the total leakage limit. The total leakage limit considers RCS inventory makeup capability and drywell sump capacity. Drywell Equipment Drain Sump instrumentation is required to support verification of the Total Leakage limit.

With RCS unidentified or total leakage greater than the limits, actions must be taken to reduce the leak. Because the leakage limits are conservatively below the leakage that would constitute a critical crack size, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed to reduce the leakage rates before the reactor must be shut down. If unidentified leakage has been identified and quantified, it may be reclassified and considered as identified leakage; however, the total leakage limit would remain unchanged.

An unidentified leakage increase of > 2 gpm within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is an indication of a potential flaw in the Reactor Coolant Pressure Boundary (RCPB) and must be quickly evaluated. The increase does not necessarily violate the absolute unidentified leakage limit, therefore, an option exists to allow continued reactor operation if certain susceptible components are determined not to be the source of the leakage increase within the required completion time. For an unidentified leakage increase greater than required limits, an alternative to reducing leakage increase to within limits (i.e., reducing the leakage rate such that the current rate is less than the "2'gpm increase in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" limit; either by isolating the source or other possible methods) is to evaluate service sensitive type 304 and type 316 austentic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased leakage. This type of piping is very susceptible to IGSCC. Note also that once leakage is attributed to a specific source, that leakage can be considered to be identified and can be applied against the identified limit, rather than the unidentified limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is reasonable to properly reduce the unidentified leakage increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety.

The Surveillance Requirement (SR) associated with RCS leakage is acceptable because RCS leakage is monitored by a variety of instruments designed to provide alarms when leakage is indicated and to quantify the various types of leakage. Sump level and flow rate are typically monitored to determine actual leakage rates; however, other methods may be used to verify leakage. It is permissible to use pre-existing information, in conjunction with secondary measurements (e.g., Drywell pressure and temperature), to verify that leakage remains within limits by looking for step changes in conditions or to perform calculations to estimate leakage. The complete failure to demonstrate that RCS leakage is within limits, on the required frequency, constitutes a failure to meet this SR, notwithstanding entrance into conditions and required actions of TS 3.6.D.2.

3.6/4.6 BASES 150 06/11/02 Amendment No. 11, 30, 100, 102,4128

Bases 3.6/4.6 (Continued):

2. RCS Leakage Detection Instrumentation Two leakage collection sumps are provided inside primary containment. Identified leakage is piped from the recirculation pump seals, valve stem leak-offs, reactor vessel flange leak-off, bulkhead and bellows drains, and vent cooler drains to the drywell equipment drain sump. All other leakage is collected in the drywell floor drain sump. Both sumps are equipped with level and flow transmitters connected to recorders in the control room. The Drywell Floor Drain Sump Monitoring System instrumentation consists of one floor drain sump flow integrator, one sump level recorder and one sump fill rate computer point (rate of change). The Drywell Floor Drain Sump Monitoring System is operable when any one of these three channels is operable. An annunciator and computer alarm are provided in the control room to alert operators when allowable leak rates are approached.

Drywell airborne particulate radioactivity is continuously monitored as well as drywell atmospheric temperature and pressure. The drywell particulate radioactivity monitoring system monitors the drywell for airborne particulate radioactivity. A sudden increase in radioactivity may be attributed to RCPB steam or reactor water leakage. The drywell particulate radioactivity monitoring system is not capable of quantifying leakage rates, but is sensitive enough to indicate increased leakage rates. The drywell particulate radioactivity monitoring system provides a backup to the Drywell Floor Drain Sump Monitoring System and is capable of monitoring leakage at least as low as 10-9 iCi/cc radioactivity for air particulate monitoring. Systems connected to the reactor coolant systems boundary are also monitored for leakage by the Process Liquid Radiation Monitoring System.

The Drywell Floor Drain Sump Monitoring System is required to quantify the unidentified leakage from the RCS. Thus, for the system to be considered operable, either the flow monitoring or the sump level monitoring portion of the system must be operable. Any failure of a sump monitoring subsystem should be evaluated for its impact on the ability of the associated instrumentation to measure leakage.

Since the flow integrator for each sump is not directly tied to the sump for its input signals, they are not affected in the same way as other instrumentation. However, the loss of flow through the flow integrator prevents the flow integrator from performing its intended safety function of measuring leakage, and even though ifs associated SRs continue to be met, it should be declared inoperable.

It should be noted that system isolation in response to Required Actions of LCO 3.7.D.2, would not render these instruments inoperable, provided the system could be unisolated as allowed by the footnote of LCO 3.7.0.2, as manual operation is allowed.

The total loss of the Drywell Floor Drain Sump Monitoring System results from the loss of all flow and level instrumentation (either directly or indirectly).

An alternate to the Drywell Floor Drain Sump Monitoring System is the drywell equipment drain sump system. Because of the physical size of the sumps, it is possible through detection or calculation to verify the required leakage limit (5 gpm) and rate limit (2 gpm/24 hrs) during the period of time it takes to actually overflow from one sump to the other. Once the drywell floor drain sump is overflowing to the drywell equipment drain sump, the drywell equipment drain sump system can be used to quantify leakage. However, the alarm settings for the equipment drain sump instruments must be reset to detect the lower limit for unidentified leakage. In this condition, all additional leakage measured by the drywell equipment drain sump system is assumed to be unidentified leakage unless the leakage has been identified and 3.6/4.6 BASES 151 6/614/40-Amendment No. 30, 76,93,10-a+428

quantified. The opposite situation is also allowed, where the equipment drain sump is allowed to overflow into the floor drain sump. In this configuration, the alarm settings need not be reset, as they would conservatively quantify all additional leakage as unidentified, unless the leakage has been identified and quantified, and alarm at the appropriate limit. The other monitoring systems provide additional indication to the operators so closer examination of other detection systems will be made to determine the extent of any corrective action that may be required. With the leakage detection systems inoperable, monitoring for leakage is degraded.

With the Drywell Floor Drain Sump Monitoring System inoperable, no other form of sampling can provide the equivalent information to quantify unidentified leakage. However, the drywell particulate radioactivity monitoring system will provide indication of changes in leakage.

With the Drywell Floor Drain Sump Monitoring System inoperable, operation may continue for 30 days. The 30 days is acceptable, based on operating experience, considering other methods of detecting leakage are available. The action requirements are modified by a footnote that allows a Mode change when the Drywell Floor Drain Sump Monitoring System is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

With the drywell particulate radioactivity monitoring system inoperable, operation may continue as long as grab samples are taken every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to analyze the drywell atmosphere. The action requirements are modified by a footnote that allows a Mode change when the drywell particulate radioactivity monitoring system is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

With the required leakage detection instrumentation inoperable, no means of detecting leakage is available. This condition does not provide the required means of leakage detection. The required action is to restore one channel of the inoperable monitoring systems (Drywell Floor Drain Sump Monitoring System or drywell particulate radioactivity monitoring system) to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to regain the intended leakage detection capability. The 1-hour completion time ensures that the plant will not be operated in a degraded configuration for a lengthy time period.

The sensitivity of the sump leakage detection systems for detection of leak rate changes is better than one gpm in a one hour period.

Other leakage detection methods provide warning of abnormal leakage and are not directly calibrated to provide leak rate measurements.

E. Safety/Relief Valves The reactor coolant system safety/relief valves assure that the reactor coolant system pressure safety limit is never reached. In compliance with Section IlIl of the ASME Boiler and Pressure Vessel Code, 1965 Edition, the safety/relief valves must be set to open at a pressure no higher than 105 percent of design pressure, with at least one safety/relief valve set to open at a pressure no greater than design pressure, and they must limit the reactor pressure to no more than 110 percent of design pressure. The safety/relief valves are sized according to the Code for a condition of MSIV closure while operating at 1775 MWt, followed by no MSIV closure scram but scram from an indirect (high flux) means. With the safety/relief valves set as specified herein, the maximum vessel pressure remains below the 1375 psig ASME Code limit. Only five of the eight valves are assumed to be operable in this analysis and the valves are assumed to open at 3% above their setpoint of 1109 psig with a 0.4 second delay. The upper limit on safety/relief valve 3.6/4.6 BASES 152 Amendment No.

Bases 3.6/4.6 (Continued):

setpoint is established by the operating limit of the HPCI and RCIC systems of 1120 psig. The design capability of the HPCI and RCIC systems has been conservatively demonstrated to be acceptable at pressures 3% greater than the safety/relief valve setpoint of 1109 psig.

HPCI and RCIC pressures required for system operation are limited by the Low-Low Set SRV System to well below these values.

The safety/relief valves have two functions; 1) over-pressure relief (self-actuation by high pressure), and 2) Depressurization/ Pressure Control (using air actuators to open the valves via ADS, Low-Low Set system, or manual operation).

The safety function is performed by the same safety/relief valve with self-actuated integral bellows and pilot valve causing main valve operation. Article 9, Section N-911.4(a)(4) of the ASME Pressure Vessel Code Section III Nuclear Vessels (1965 and 1968 editions) requires that these bellows be monitored for failure since this would defeat the safety function of the safety/relief valve.

Low-Low Set Logic has been provided on three non-Automatic Pressure Relief System valves. This logic is discussed in detail in the Section 3.2 Bases.

This logic, through pressure sensing instrumentation, reducesthe opening setpoint and increases the blowdown range of the three selected valves following a scram to eliminate the discharge line water leg clearing loads resulting from multiple valve openings.

Testing of the safety/relief valves in accordance with ANSI/ASME OM-1 -1981 each refueling outage ensures that any valve deterioration is detected. An as-found tolerance value of 3% for safety/relief valve setpoints is specified in ANSI/ASME OM-1-1981. Analyses have been performed with the valves assumed to open at 3% above their setpoint of 1109 psig. The 1375 psig Code limit is not exceeded in any case. When the setpoint is being bench checked, it is prudent to disassemble one of the safety/relief valves to examine for crud buildup, bending of certain actuator members or other signs of possible deterioration.

Provision also has been made to detect failure of the bellows monitoring system. Testing of this system once per cycle provides assurance of bellows integrity.

F. Deleted 3.6/4.6 BASES 152a Amendment No.

-Fm With one or more penetration flow paths with one PCIV inoperableIhe affected penetration must be returned to operable status or isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for MSIVs and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> fords'Q). The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time for MSIVs allows a period of time to restore the MVISIVs to operable status given the fact that MSIV closure will result in a potential for plant shutdown. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time for EFCVs is reasonable considering the instrument and the small diameter of the penetration piping combined with the ability of the penetration to act as an isolation boundary. With one or more penetrations with two PCIVs inoperable, either the inoperable PCIVs must be returned to operable status or the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Specification 3.7.D.3 requires the containment to be purged and vented through the standby gas treatment system except during inerting and deinerting operations. This provides for iodine and particulate removal from the containment atmosphere.

Use of the 2-inch flow path prevents damage to the standby gas treatment system in the event of a loss of coolant accident during purging or venting. Use of the reactor building plenum and vent flow path for inerting and deinerting operations permits the control room operators to monitor the activity level of the resulting effluent by use of the Reactor Building Vent Wide Range Gas Monitors.

E. Combustible Gas Control System The function of the Combustible Gas Control System (CGCS) is to maintain oxygen concentrations in the post-accident containment atmosphere below combustible concentrations. Oxygen may be generated in the hours following a loss of coolant accident from radiolysis of reactor coolant.

The Technical Specifications limit oxygen concentrations during operation to less than four percent by volume during operation.

The maintenance of an inert atmosphere during operation precludes the build-up of a combustible mixture due to a fuel metal-water reaction. The other potential mechanism for generation of combustible mixtures is radiolysis of coolant which has been found to be small.

A special report is required to be submitted to the Commission to outline CGCS equipment failures and corrective actions to be taken if inoperability of one train exceeds thirty days. In addition, if both trains are inoperable for more than 30 days, the plant is required to shutdown until repairs can be made.

3.7 BASES 182a Amendment No. 130 09/23/02 I

Bases 3.13: I The alternate shutdown system panel is provided to assure the capability of achieving cold shutdown, external to the control room, in the unlikely event the control room becomes uninhabitable or safe shutdown equipment in the control room or cable spreading room is damaged by fire. Control of those systems on the alternate shutdown system panel is taken when the locking master transfer switch is moved from the normal to the transfer position and each system's individual transfer switch is put in the transfer mode. When control is established at the alternate shutdown system panel no control of those systems is available from the control room and all automatic initiation signals have been disabled. The master transfer switch shall remain in the locked position at all times when not in use, being tested or being maintained. If the master transfer switch is moved to the transfer position there is an alarm in the control room.

APbU AS A AUce1 "System controls on the ASDS panel" refers to portions of control circuits and instrumentation necessary solely to support ASDS functions, incling equipment not located within the ASDS panel. The ASDS LCO is entered for any of the following conditions:

1. 12 RERSW pump inoperable.

2. System controls on the ASDS panel inoperable.

3.13 BASES 225 34tIe51t, I Amendment No.7-33,77, 85,19-

ENCLOSURE 3 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 Technical Specifications. The pages included are listed below:

Paae 150 151 152 152a 152b 182a 225 7 pages follow

Bases 3.6/4.6 (Continued):

D. Reactor Coolant System (RCS)

1. RCS Operational Leakage The allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for unidentified leakage,.the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be Pressure Boundary Leakage and they cannot be reduced within the allowed times, the reactor will be shutdown to allow further investigation and corrective action.

The low limit on increase in Unidentified Leakage assumes a failure mechanism of Intergranular Stress Corrosion Cracking (IGSCC) that produces tight cracks. This flow increase limit is capable of providing an early warning of such deterioration.

No applicable safety analysis assumes the Total Leakage limit. The Total Leakage limit considers RCS inventory makeup capability and drywell sump capacity. Drywell Equipment Drain Sump instrumentation is required to support verification of the Total Leakage limit.

With RCS Unidentified or Total Leakage greater than the limits, actions must be taken to reduce the leak. Because the leakage limits are conservatively below the leakage that would constitute a critical crack size, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed to reduce the leakage rates before the reactor must be shut down. If Unidentified Leakage has been identified and quantified, it may be reclassified and considered as Identified Leakage; however, the Total Leakage limit would remain unchanged.

An Unidentified Leakage increase of > 2 gpm within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is an indication of a potential flaw in the Reactor Coolant Pressure Boundary (RCPB) and must be quickly evaluated. The increase does not necessarily violate the absolute Unidentified Leakage limit, therefore, an option exists to allow continued reactor operation if certain susceptible components are determined not to be the source of the leakage increase within the required completion time. For an Unidentified Leakage increase greater than required limits, an alternative to reducing leakage increase to within limits (i.e., reducing the leakage rate such that the current rate is less than the "2 gpm increase in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" limit; either by isolating the source or other possible methods) is to evaluate service sensitive type 304 and type 316 austenitic stainless steel piping that is subject to high stress or 3.6/4.6 BASES 150 10/09/03 Amendment No. 14, 30,100,102,128, 137a

Bases 3.6/4.6 (Continued):

that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased leakage. This type of piping is very susceptible to IGSCC. Note also that once leakage is attributed to a specific source, that leakage can be considered to be identified and can be applied against the identified limit, rather than the unidentified limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is reasonable to properly reduce the Unidentified Leakage increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety.

The Surveillance Requirement (SR) associated with RCS leakage is acceptable because RCS leakage is monitored by a variety of instruments designed to provide alarms when leakage is indicated and to quantify the various types of leakage. Sump level and flow rate are typically monitored to determine actual leakage rates; however, other methods may be used to verify leakage.

It is permissible to use pre-existing information, in conjunction with secondary measurements (e.g., drywell pressure and temperature), to verify that leakage remains within limits by looking for step changes in conditions or to perform calculations to estimate leakage. The complete failure to demonstrate that RCS leakage is within limits, on the required frequency, constitutes a failure to meet this SR, notwithstanding entrance into conditions and required actions of TS 3.6.D.2.

2. RCS Leakage Detection Instrumentation Two leakage collection sumps are provided inside primary containment. Identified leakage is piped from the recirculation pump seals, valve stem leak-offs, reactor vessel flange leak-off, bulkhead and bellows drains, and vent cooler drains to the drywell equipment drain sump. All other leakage is collected in the drywell floor drain sump. Both sumps are equipped with level and flow transmitters connected to recorders in the control room. The Drywell Floor Drain Sump Monitoring System instrumentation consists of one floor drain sump flow integrator, one sump level recorder and one sump fill rate computer point (rate of change).

The Drywell Floor Drain Sump Monitoring System is operable when any one of these three channels is operable. An annunciator and computer alarm are provided in the control room to alert operators when allowable leak rates are approached.

Drywell airborne particulate radioactivity is continuously monitored as well as drywell atmospheric temperature and pressure.

The drywell particulate radioactivity monitoring system monitors the drywell for airborne particulate radioactivity. A sudden increase in radioactivity may be attributed to RCPB steam or reactor water leakage. The drywell particulate radioactivity monitoring system is not capable of quantifying leakage rates, but is sensitive enough to indicate increased leakage rates. The drywell particulate radioactivity monitoring system provides a backup to the Drywell Floor Drain Sump Monitoring System and is capable of monitoring leakage at least as low as 10-9oCi/cc radioactivity for air particulate monitoring. Systems connected to the reactor coolant systems boundary are also monitored for leakage by the Process Liquid Radiation Monitoring System.

3.6/4.6 BASES 151 10/09/03 Amendment No. 30, 76, 93, 100a, 114, 128, 137a

Bases 3.6/4.6 (Continued):

The Drywell Floor Drain Sump Monitoring System is required to quantify the unidentified leakage from the RCS. Thus, for the system to be considered operable, either the flow monitoring or the sump level monitoring portion of the system must be operable. Any failure of a sump monitoring system should be evaluated for its impact on the ability of the associated instrumentation to measure leakage.

Since the flow integrator for each sump is not directly tied to the sump for its input signals, they are not affected in the same way as other instrumentation. However, the loss of flow through the flow integrator prevents the flow integrator from performing its intended safety function of measuring leakage, and even though its associated SRs continue to be met, it should be declared inoperable.

It should be noted that system isolation in response to Required Actions of LCO 3.7.D.2, would not render these instruments inoperable, provided the system could be unisolated as allowed by the footnote of LCO 3.7.D.2, as manual operation is allowed.

The total loss of the Drywell Floor Drain Sump Monitoring System results from the loss of all flow and level instrumentation (either directly or indirectly).

An alternate to the Drywell Floor Drain Sump Monitoring System is the drywell equipment drain sump system. Because of the physical size of the sumps, it is possible through detection or calculation to verify the required leakage limit (5 gpm) and rate limit (2 gpm/24 hours) during the period of time it takes to actually overflow from one sump to the other. Once the drywell floor drain sump is overflowing to the drywell equipment drain sump, the drywell equipment drain sump system can be used to quantify leakage. However, the alarm settings for the equipment drain sump instruments must be reset to detect the lower limit for unidentified leakage. In this condition, all additional leakage measured by the drywell equipment drain sump system is assumed to be Unidentified Leakage unless the leakage has been identified and quantified. The opposite situation is also allowed, where the equipment drain sump is allowed to overflow into the floor drain sump. In this configuration, the alarm settings need not be reset, as they would conservatively quantify all additional leakage as unidentified, unless the leakage has been identified and quantified, and alarm at the appropriate limit. The other monitoring systems provide additional indication to the operators so closer examination of other detection systems will be made to determine the extent of any corrective action that may be required. The drywell particulate radioactivity monitoring system provides a backup system to the Drywell Floor Drain Sump Monitoring System. With the leakage detection systems inoperable, monitoring for leakage is degraded.

With the Drywell Floor Drain Sump Monitoring System inoperable, no other form of sampling can provide the equivalent information to quantify Unidentified Leakage. However, the drywell particulate radioactivity monitoring system will provide indication of changes in leakage.

3.6/4.6 BASES 152 10/09/03 Amendment No. 137a

Bases 3.6/4.6 (Continued):

With the Drywell Floor Drain Sump Monitoring System inoperable, operation may continue for 30 days. The 30 days is acceptable, based on operating experience, considering other methods of detecting leakage are available. The action requirements are modified by a footnote that allows a Mode change when the Drywell Floor Drain Sump Monitoring System is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

With the drywell particulate radioactivity monitoring system inoperable, operation may continue as long as grab samples are taken every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to analyze the drywell atmosphere. The action requirements are modified by a footnote that allows a Mode change when the drywell particulate radioactivity monitoring system is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

With the required leakage detection instrumentation inoperable, no means of detecting leakage is available. This condition does not provide the required means of leakage detection. The required action is to restore one channel of the inoperable monitoring systems (Drywell Floor Drain Sump Monitoring System or drywell particulate radioactivity monitoring system) to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to regain the intended leakage detection capability. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time ensures that the plant will not be operated in a degraded configuration for a lengthy time period.

The sensitivity of the sump leakage detection systems for detection of leak rate changes is better than one gpm in a one hour period. Other leakage detection methods provide warning of abnormal leakage and are not directly calibrated to provide leak rate measurements.

E. Safety/Relief Valves The reactor coolant system safety/relief valves assure that the reactor coolant system pressure safety limit is never reached. In compliance with Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, the safety/relief valves must be set to open at a pressure no higher than 105 percent of design pressure, with at least one safety/relief valve set to open at a pressure no greater than design pressure, and they must limit the reactor pressure to no more than 110 percent of design pressure. The safety/relief valves are sized according to the Code for a condition of MSIV closure while operating at 1775 MWt, followed by no MSIV closure scram but scram from an indirect (high flux) means. With the safety/relief valves set as specified herein, the maximum vessel pressure remains below the 1375 psig ASME Code limit. Only five of the eight valves are assumed to be operable in this analysis and the valves are assumed to open at 3% above their setpoint of 1109 psig with a 0.4 second delay. The upper limit on safety/relief 3.6/4.6 BASES 152a 10/09/03 Amendment No. 137a l

Bases 3.6/4.6 (Continued):

valve setpoint is established by the operating limit of the HPCI and RCIC systems of 1120 psig. The design capability of the HPCI and RCIC systems has been conservatively demonstrated to be acceptable at pressures 3% greater than the safety/relief valve setpoint of 1109 psig. HPCI and RCIC pressures required for system operation are limited by the Low-Low Set SRV System to well below these values.

The safety/relief valves have two functions; 1) over-pressure relief (self-actuation by high pressure), and 2) Depressurization/

Pressure Control (using air actuators to open the valves via ADS, Low-Low Set system, or manual operation).

The safety function is performed by the same safety/relief valve with selfoactuated integral bellows and pilot valve causing main valve operation. Article 9, Section N-911 .4(a)(4) of the ASME Pressure Vessel Code Section III Nuclear Vessels (1965 and 1968 editions) requires that these bellows 'be monitored for failure since this would defeat the safety function of the safety/relief valve.

Low-Low Set Logic has been provided on three non-Automatic Pressure Relief System valves. This logic is discussed in detail in the Section 3.2 Bases. This logic, through pressure sensing instrumentation, reduces the opening setpoint and increases the blowdown range of the three selected valves following a scram to eliminate the discharge line water leg clearing loads resulting from multiple valve openings.

Testing of the safety/relief valves in accordance with ANSI/ASME OM-1 -1981 each refueling outage ensures that any valve deterioration is detected. An as-found tolerance value of 3% for safety/relief valve setpoints is specified in ANSI/ASME OM-1 -1981.

Analyses have been performed with the valves assumed to open at 3% above their setpoint of 1109 psig. The 1375 psig Code limit is not exceeded in any case. When the setpoint is being bench checked, it is prudent to disassemble one of the safety/relief valves to examine for crud buildup, bending of certain actuator members or other signs of possible deterioration.

Provision also has been made to detect failure of the bellows monitoring system. Testing of this system once per cycle provides assurance of bellows integrity.

F. Deleted 3.6/4.6 BASES 152b 10/09/03 Amendment No. 137a

With one or more penetration flow paths with one PCIV inoperable, the affected penetration must be returned to operable status or isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for MSIVs and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for Excess Flow Check Valves (EFCVs)). The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time for MSIVs allows a period of time to restore the MSIVs to operable status given the fact that MSIV closure will result in a potential for plant shutdown. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time for EFCVs is reasonable considering the instrument and the small diameter of the penetration piping combined with the ability of the penetration to act as an isolation boundary. With one or more penetrations with two PCIVs inoperable, either the inoperable PCIVs must be returned to operable status or the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Specification 3.7.D.3 requires the containment to be purged and vented through the standby gas treatment system except during inerting and deinerting operations. This provides for iodine and particulate removal from the containment atmosphere.

Use of the 2-inch flow path prevents damage to the standby gas treatment system in the event of a loss of coolant accident during purging or venting. Use of the reactor building plenum and vent flow path for inerting and deinerting operations permits the control room operators to monitor the activity level of the resulting effluent by use of the Reactor Building Vent Wide Range Gas Monitors.

E. Combustible Gas Control System The function of the Combustible Gas Control System (CGCS) is to maintain oxygen concentrations in the post-accident containment atmosphere below combustible concentrations. Oxygen may be generated in the hours following a loss of coolant accident from radiolysis of reactor coolant.

The Technical Specifications limit oxygen concentrations during operation to. less than four percent by volume during operation.

The maintenance of an inert atmosphere during operation precludes the build-up of a combustible mixture due to a fuel metal-water reaction. The other potential mechanism for generation of combustible mixtures is radiolysis of coolant which has been found to be small.

A special report is required to be submitted to the Commission to outline CGCS equipment failures and corrective actions to be taken if inoperability of one train exceeds thirty days. In addition, if both trains are inoperable for more than 30 days, the plant is required to shutdown until repairs can be made.

3.7 BASES 182a 09/25/03 Amendment No. 4130-, 136a

Bases 3.13:

The alternate shutdown system panel is provided to assure the capability of achieving cold shutdown, external to the control room, in the unlikely event the control room becomes uninhabitable or safe shutdown equipment in the control room or cable spreading room is damaged by fire. Control of those systems on the alternate shutdown system panel is taken when the locking master transfer switch is moved from the normal to the transfer position and each system's individual transfer switch is put in the transfer mode. When control is established at the alternate shutdown system panel no control of those systems is available from the control room and all automatic initiation signals have been disabled. The master transfer switch shall remain in the locked position at all times when not in use, being tested or being maintained. If the master transfer switch is moved to the transfer position there is an alarm in the control room.

"System controls on the ASDS panel" refers to portions of control circuits and instrumentation necessary solely to support ASDS functions, including equipment not located within the ASDS panel. The ASDS LCO is entered for any of the following conditions:

1. 12 RHRSW pump inoperable.

2. System controls on the ASDS panel inoperable.

3.13 BASES 225 10/14/03 Amendment No. 7, 33, 77, 85, 11 9, 137b

ATTACHMENT 4 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 I

J 10 pages follow

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

A 137b 36 128 71a 129b 122 135 B 137b 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 0 126 137 F 115 42 103 79 0 126a 137 G 115 45 0 80 29 127 137 H 119 46 70 81 3 128 42 1 130a 46a 37 82 123 129 122 J 137b 47 40 82a 63 130 82 i 128 48 89 83 24 131 122 ii 137 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 100a 134 133 vi 121 51a 117 87 100a 135 133 vii 122 52 128 88 100a 136 133 1 119 53 128 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 146 135 5 137 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 137a 8 128 59a 103 97 57 151 137a 9 128 60 128 98 56 152 137a 10 128 60a 31 99 104 152a 137a 11 128 60b 62 100 100a 152b 137a 12 128 60c 30 101 122 153 100a 25a 127 60d 128 102 122 154 129a 25b 127 60e 89 103 122 155 122 25c 127 61 104 104 122 156 93 25d 127 62 117 105 122 157 130 26 5 63 117 106 79 158 132 27 81 63a 117 107 97 159 132 27a 81 64 135a 108 128 160 132 28 128 65 117 109 100a 163 130 29 128 66 119a 110 100a 164 104 30 103 67 117 111 133a 165 130 31 104 68 129b 112 130a 166 130 32 103 69 129b 113 130a 167 112 33 103 69a 129b 114 133a 168 94 34 83 70 117 115 130a 169 94 35 100a 71 100a 121 0 170 130 A

Amendment No. 137b 10/14/03

MONTICELLO NUCLEAR GENERATING PLANT APPENDIX A TECHNICAL SPECIFICATIONS RECORD OF REVISIONS Amend Amend Pace No. Page No.

171 130 218 120 171a 130 223 119 172 71 224 119 175 107 225 137b I 175a 117 226 119 176 100a 229a 63 177 130 229b 104 178 100a 229c 104 179 123a 229d 63 180 130 229e 122 181 130 229u 104 182 130 229v 112 182a 136a 229vv 112 183 117 229w 112 184 100a 229ww 112 185 134 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 0 254 136 207 123 255 120 208 63 256 122 209 123 257 122 209a 100a 258 134 210 100a 258a 132 211 131 259 120 212 109 260 120 213 99 261 120 216 100a 262 120 217 128 B

Amendment No. 137b 10/14/03

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 Maior 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 l 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 Limits 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 2.4 Tech. Specs.

C 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 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 8x8R Authorization, 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 (REV' 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 Radiation 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 Line 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 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 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 90 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 NSP License Revision DPR-22 (REV) No. Amend No. & Date Major Subject 117 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 STA/LSO 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 l-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 08/28/98 Reactor Coolant Equivalent Radioiodine Concentration and Control Room Habitability 102 09/16/98 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 02/16/00 Safety Limit Minimum Critical Power Ratio for Cycle 20 110 08/07/00 Transfer of Operating Authority from NSP to NMC 111 08/18/00 Transfer of Operating License from NSP to a New Utility Operating Company 112 08/18/00 Emergency Filtration Train Testing Exceptions and Technical Specification Revisions 113 10/02/00 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 98Q1 45 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 11 8a 05/10/01 Bases Change - 50'F Loop Temperature, Bus Transfer &

Rerate Correction H

Amendment No. 119 04/05/01

i MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (REV} No. Amend No. & Date Maeor Subject 119 04/05101 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 subsystem 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

4 I MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (REV) No. Amend No. & Date Maior Subject 131 10/02/02 Update the Multiplier Values for Single Loop Operation Average Planar Linear Heat Generation Rate (APLHGR) 132 02/04/03 Conversion to Option B for Containment Leak Rate Testing 133 02/24/03 Revision to Pressure-Temperature Curves 133a 03/28/03 Bases Change - Adequate Reactor Steam Flow for HPCI/RCIC Testing 134 03/31/03 One-Time Extension of Containment Integrated Leak-Rate Test Interval 135 04/22/03 Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program 135a 04/24/03 Bases Change - Clarify description of head cooling Group 2 valves 136 06/17/03 Elimination of Requirements for Post Accident Sampling System 136a 09/25/03 Bases Change - Editorial change to define the abbreviation "EFCV."

137 08/21/03 Drywell Leakage and Sump Monitoring Detection System 137a 10/09/03 Bases Change - RCS Leakage Requirements for TS 3.6.4.D 137b 10/14/03 Bases Change - Clarification of system control boundary for ASDS

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 100a. For remaining Tech Spec pages, no other changes made and current Amendment Nos. used.

J Amendment No. 137b 10/14/03