ML022750435

Technical Specification Bases Pages

ML022750435
Person / Time
Site:	Monticello
Issue date:	09/19/2002
From:	Forbes J Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML022750435 (29)

Text

NMC Committed to NuclerExcellence Monticello Nuclear Generatinq Plant Operated by Nuclear Management Company, LLC September 19, 2002 Technical Specification 6.8.K US 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 suffixes "a" and "b." 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. Ttie current copy of our list of effective pages and record of revision is attached for your information, as Attachment D.

Please contact Paul Hartmann at 763-271-5172 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 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 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-129a Technical Specification Involved - 3.6.1H.1 Page affected - 154 Summary of Change: Modification of the description of when snubbers are required for operation. The Bases description was changed to be consistent with the associated Limiting Condition of Operation.

TSBC-129b Technical Specification Involved - 3.1 and 3.2 Pages affected - 39, 40, 68, 69, 69a, 71a Summary of Change: This TSBC removed language from the Technical Specification Bases that implies deviation values can be used to modify technical specification instrumentation trip settings and Safety Relief Value Setpoints (Low-Low Set Logic).

Other changes include relocation of Bases language between adjacent pages.

A-1

Attachment B 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:

Pagie 39 40 68 69 69a 71 a 154 B-1

Bases 3.1 (Continued):

10. Main Steamline Isolation Valve Closure The main steamline isolation valve closure scram is set to scram when the isolation valves are -<10% closed from full open.

This scram anticipates the pressure and flux transient, which would occur when the valves close. By scramming at this setting the resultant transient is insignificant.

11. Turbine control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass. This transient is less severe than the turbine stop valve closure with bypass failure and therefore adequate margin exists. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power. In order to ensure the availability of this scram above 45% rated thermal power, this scram is only bypassed below 30% thermal power as indicated by turbine first stage pressure. This takes into account the possibility of 14% power being passed directly to the condenser through the bypass valves.

12. Turbine Stop Valve Closure The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10% of valve closure from full open, the resultant increase in -7 surface heat flux is limited such that MCPR remains above the Safety Limit (T.S.2.1 .A) even during the worst case transient thaL assumes the turbine bypass is closed. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power. In order to ensure the availability of this scram above 45% rated thermal power, this scram is only bypassed below 30% thermal power as indicated by turbine first stage pressure. This takes into account the possibility of 14%, power being passed directly to the condenser through the bypass valves.

Although the operator will set the set points within the trip settings specified on Table 3.1.1, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. For power rerate, GE setpoint methodology provided in NEDC 31336, "General Electric Setpoint Methodology," is used in establishing setpoints. The deviations could be caused by inherent.

instrument error, operator setting error, drift of the set point, etc. Therefore, such deviations have been accounted for in the various transient analyses. and the actual trip s*ttings may va.y by the following amounts:

3.1 BASES 39 Amendment No. 50, 66, 76, 81, l 00a, 102,-12

Bases 3. 1 (Continued):

Trip Function Deviation Trip Function Deviation

3. High Flux IRM +2/125 of scale *7. Reactor Low Water Level -6 inches

5. High Reactor Pressure +10 psi 8. Scram Discharge Volume High Level +1 gallon

6. High Drywell Pressure +1 psi 9. Turbine Condenser Low Vacuum" -1/2 in. Hg This indication is reactor coolant temperature sensitive. The calibration is thus made for rated conditions. The level error at low pressures and temperatures is bounded by the safety analysis which reflects the weight-of-coolant above the lower tap, and not the indicated level.

Aviolation of this specification is assumed to occur only when a dVic, is knowingly set outside of the I,miting trip setting, O a sufficient numbehr f dcVices have been affe*ted by an" means such that the automatic function is incapable of operating within the allowablwe depviatioan hiein.Areac-tor modep- in.whlich the specifed-fu-nc~tion mrust be operable, or the actions specificd- in 33.1.B- are not initiated as specified.

If an unsafe failure is detected during surveillance testing, it is desirable to determine as soon as possible if other failures of a similar type have occurred and whether the particular function involved is still operable or capable of meeting the single failure criterion. To meet the requirements of Table 3.1.1, it is necessary that all instrument channels in one trip system be operable to permit testing in the other trip system. Thus, when failures are detected in the first trip system tested, they would have to be repaired before testing of the other system could begin. In the majority of cases, repairs or replacement can be accomplished quickly. If repair or replacement cannot be completed in a reasonable time, operation could continue with one tripped trip system until the surveillance testing deadline.

The ability to bypass one instrument channel when necessary to complete surveillance testing will preclude continued operation with scram functions which may be either unable to meet the single failure criterion or completely inoperable. It also eliminates the need for an unnecessary shutdown if the remaining channels are found to be operable. The conditions under which the bypass is permitted require an immediate determination that the particular function is operable. However, during the time a bypass is applied, the function will not meet the single failure criterion; therefore, it is prudent to limit the time the bypass is in effect by requiring that surveillance testing proceed on a continuous basis and that the bypass be removed as soon as testing is completed.

3.1 BASES NEXT PAGE IS 42 40 Amendment No. n 1n0an -12

Bases 3.2 (Continued):

The RBM bypass time delay is set low enough to assure minimum rod movement while upscale trips are bypassed.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the Safety Limit (T.S.2.1.A).

A downscale indication of an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus control rod motion is prevented. The downscale rod blocks assure that there will be proper overlap between the neutron monitoring systems and thus, that adequate coverage is provided for all ranges of reactor operation. The downscale trips are set at 3/125 of full scale.

For effective emergency core cooling for the small pipe break the HPCI or Automatic Pressure Relief system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The arrangement of the tripping contacts is such as to provide this functior when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria is met. Reference Section 6.2.4 and 6.2.6 FSAR. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., onlyione instrument channel out of service.

Four radiation monitors (two reactor building vent plenum and two refueling floor) are provided which initiate isolation of the reactor building and operation of the standby gas treatment system following a refueling accident. The monitors measure radioactivity in the reactor building ventilation exhaust and on the refueling floor. One upscale trip signal or two downscale/inoperable trip signals, from a pair of monitors performing the same function, will cause the desired action. Trip settings of 100 mR/hr for the reactor building vent plenum monitors and.the iefueling floor monitors are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

The recirculation pump trip is provided to minimize reactor pressure in the highly unlik6ly event of a plant transient coincident with the failure of all control rods to scram. The rapid flow reduction [ increases core voiding, a negative reactivity feedback. High pressure sensors initiate the pump trip in the event of an isolation tranrsient. Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety/relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative. ]

[moved first paragraph from page 69 to this page 1 3.2 BASES 68 Amendment No. 15, 29, 40, 50, 71, 100a, 10n2

Bases 3.2 (Continued):

The ATWS high reactor pressure and low-low water level logic also initiates the Alternate Rod Injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve to vent the header and causes rod insertion. This greatly reduces the long term consequences of an ATWS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an alternate source when a loss of voltage condition or a degraded voltage condition is sensed. On loss of voltage, the voltage sensing relays trip immediately and energize auxiliary relays that control the bus transfer sequence. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage '

for a steady state LOCA load'that maintains adequate voltage at the 480V essential MCCS. An allowance for relay tolerance is included.

Safety/relief valve low-low set logic is provided to prevent any safety/relief valve from opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when steam condenses in the line. If the valve reopens before the discharge line vacuum breakers act to return water level to normal, water clearing thrust loads on the discharge line may exceed their design limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves following a scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to [ open and instrumentation drift has caused the nominal 80-psi blowdown range to be reduced to 60 psi. Maximum water leg clearing time has been calculated to be less than 6 seconds for the Monticello design.

Inhibit timers are provided for each valve to prevent the valve from being manually opened less than 10 seconds following valve closure. Valve opening is sensed by pressure switches in the valve discharge line. Each-valve is provided with two trip, or actuation, systems. Each system is provided with two channels of instrumentation for each of the above described functions. A

- two-out-of-two-once logic scheme ensures that no single failure will defeat the low-low set function and no single failure will cause spurious operation of a safety/relief valve. Allowable deviations are provided for each specified instrument setpoint. Valve [

operationSetpei&s within the specified allowable deviations provide assurance that subsequent safety/relief valve actuations are sufficiently spaced to allow for discharge line water leg clearing. ]

[moved first paragraph from page 69a to this page ]

3.2 BASES 69 Amendment No. 30, 31, 4.5, 104, !! a

Bases 3.2 (Continued):

Control room habitability protection instrumentation assures that the control room operators will be adequately protected against the effects of accidental releases of radioactive leakage which may bypass secondary containment following a loss of coolant accident or radioactive releases from a steam line break accident, thus assuring that the Monticello Nuclear Generating Plant can be operated or shutdown safely.

Although the operator will set the setpoints within the trip settings specified in Tables 3.2.1 through 3.2.9, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. The deviations could be caused by inherent instrument error, operator setting error, drift of the set point, etc. Therefore, these deviations have been accounted for in the various transient analyses. and the actu-.al trip settings may v.ary by the following amounts:

3.2 BASES 69a Amendment No. 65, 70, 89, !00a

Bases 3.2 (Continued):

Trip Function Deviation Instrumentation for Safety/Relief Valve Reactor Coolant System +/- 20 psig Low Low Set Logic Pressure for Opening/Closing Opening- Closing Pressure >__

60 psi Discharge Pipe Pressure Inhibit +/- 10 psid Timer Inhibit -3 sec

+10 sec Other Instrumentation

• High Reactor Water Level +6 inches

• Low-Low Reactor Water Level -3 inches This indication is reactor coolant temperature sensitive. The calibration is thus made for rated conditions. The level error at low pressures and temperatures is bounded by the safety analysis which reflects the weight-of-coolant above the lower tap, and not the indicated level.

A vieltion of this specification is assucd,to *cGcu only when a devi-e Is knuoingly set outsidc of the limiting trip settings, or-, whca sufficientnumber of devi*eS have been affected bh an" means- suc-h that the alutomAt*i fi-uncti-o ISincapable of operating within the allowable devation while On a reactor monde in which thc specified function must be operable Aor;A whe*n actionr specified are noi initiated as speGifled.

3.2 BASES 71 a Amendment No. 30, 37, 66, 00a, 105

Bases 3.6/4.6 (Continued):

H. Snubbers All safety related snubbers are required to be operable whenever the supported system is required to be operable. The Limiting ConditonsfrOpeationisbaedonensuring

...... that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

Snubbers excluded from this inspection program are those installed on non-safety related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval is determined by the sample population size and the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from beihg counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design feature directly related to rejection of the snubber.

When a snubber is found inoperable, an engineering evaluation or inspection is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The evaluation or inspection will determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

3.6/4.6 BASES 154 Amendment No. 9, 42,2,)8 100a

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 Technical Specifications. The pages included are listed below:

Page 39 40 68 69 69a 71a 154 C-1

Bases 3.1 (Continued):

10. Main Steamline Isolation Valve Closure The main steamline isolation valve closure scram is set to scram when the isolation valves are _*10% closed from full open.

This scram anticipates the pressure and flux transient, which would occur when the valves close. By scramming at this setting the resultant transient is insignificant.

11. Turbine control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass. This transient is less severe than the turbine stop valve closure with bypass failure and therefore adequate margin exists. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power. In order to ensure the availability of this scram above 45% rated thermal power, this scram is only bypassed below 30% thermal power as indicated by turbine first stage pressure. This takes into account the possibility of 14% power being passed directly to the condenser through the bypass valves.

12. Turbine Stop Valve Closure The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10% of valve closure from full open, the resultant increase in -"

surface heat flux is limited such that MCPR remains above the Safety Limit (T.S.2.1 .A)even during the worst case transient that assumes the turbine bypass is closed. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power. In order to ensure the availability of this scram above 45% rated thermal power, this scram is only bypassed below 30% thermal power as indicated by turbine first stage pressure. This takes into account the possibility of 14% power being passed directly to the condenser through the bypass valves.

Although the operator will set the set points within the trip settings specified on Table 3.1.1, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. For power rerate, GE setpoint methodology provided in NEDC 31336, "General Electric Setpoint Methodology," is used in establishing setpoints. The deviations could be caused by inherent instrument error, operator setting error, drift of the set point, etc. Therefore, such deviations have been accounted for in the various transient analyses.

3.1 BASES 39 09/12/02 Amendment No. 50, 66, 76, 8, 4 Ob, 402, 12*, 129b

Bases 3. 1 (Continued):

Trip Function Deviation Trig Function Deviation

3. High Flux IRM +2/125 of scale *7. Reactor Low Water Level -6 inches

5. High Reactor Pressure +10 psi 8. Scram Discharge Volume High Level +1 gallon

6. High Drywell Pressure +1 psi 9. Turbine Condenser Low Vacuum -1/2 in. Hg

• This indication is reactor coolant temperature sensitive. The calibration is thus made for rated conditions. The level error at low pressures and temperatures is bounded by the safety analysis which reflects the weight-of-coolant above the lower tap, and not the indicated level.

I If an unsafe failure is detected during surveillance testing, it is desirable to determine as soon as possible if other failures of a similar type have occurred and whether the particular function involved is still operable or capable of meeting the 'single failure criterion. To meet the requirements of Table 3.1.1, it is necessary that all instrument channels in one trip system be operable to permit testing in the other trip system. Thus, when failures are detected in the first trip system tested, they would have to be repaired before testing of the other system could begin. In the majority of cases, repairs or replacement can be accomplished quickly. If repair or replacement cannot be completed in a reasonable time, operation could continue with one tripped trip system until the surveillance testing deadline.

The ability to bypass one instrument channel when necessary to complete surveillance testing will preclude continued operation with scram functions which may be either unable to meet the single failure criterion or completely inoperable. It also eliminates the need for an unnecessary shutdown if the remaining channels are found to be operable. The conditions under which the bypass is permitted require an immediate determination that the particular function is operable. However, during the time a bypass is applied, the function will not meet the single failure criterion; therefore, it is prudent to limit the time the bypass is in effect by requiring that surveillance testing proceed on a continuous basis and that the bypass be removed as soon as testing is completed.

3.1 BASES NEXT PAGE IS 42 40 09/12/02 Amendment No. 0,-100Oa,4 26r 129b

Bases 3.2 (Continued):

The RBM bypass time delay is set low enough to assure minimum rod movement while upscale trips are bypassed.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the Safety Limit (T.S.2.1 .A).

A downscale indication of an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus control rod motion is prevented. The downscale rod blocks assure that there will be proper overlap between the neutron monitoring systems and thus, that adequate coverage is provided for all ranges of reactor operation. The downscale trips are set at 3/125 of full scale.

For effective emergency core cooling for the small pipe break the HPCI or Automatic Pressure Relief system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria is met. Reference Section 6.2.4 and 6.2.6 FSAR. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

Four radiation monitors (two reactor building vent plenum and two refueling floor) are provided which initiate isolation of the reactor building and operation of the standby gas treatment system following a refueling accident. The monitors measure radioactivity in the reactor building ventilation exhaust and on the refueling floor. One upscale trip signal or two downscale/inoperable trip signals,"

from a pair of monitors performing the same function, will cause the desired action. Trip settings of 100 mR/hr for the reactor. 7 building vent plenum monitors and the refueling floor monitors are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

The recirculation pump trip is provided to minimize reactor pressure in the highly unlikely event of a plant transient coincident with the failure of all control rods to scram. The rapid flow reduction increases core voiding, a negative reactivity feedback. High pressure sensors initiate the pump trip in the event of an isolation transient. Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety/relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative.

3.2 BASES 68 09/12/02 Amendment No. 15, 29, 40, 50, 74, ln00, 102, 129b

Bases 3.2 (Continued):

The ATWS high reactor pressure and low-low water level logic also initiates the Alternate Rod Injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve to vent the header and causes rod insertion. This greatly reduces the long term consequences of an ATWS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an alternate source when a loss of voltage condition or a degraded voltage condition is sensed. On loss of voltage, the voltage sensing relays trip immediately and energize auxiliary relays that control the bus transfer sequence. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage-:

for a steady state LOCA load that maintains adequate voltage at the 480V essential MCCS. An allowance for relay tolerance is included.

Safety/relief valve low-low set logic is provided to prevent any safety/relief valve from opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when steam condenses in the line. If the valve reopens before the discharge line vacuum breakers act to return water level to normal, water clearing thrust loads on the discharge line may exceed their design limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves following a scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to open and instrumentation drift has caused the nominal 80-psi blowdown range to be reduced to 60 psi. Maximum water leg clearing time has been calculated to be less than 6 seconds for the Monticello design.

Inhibit timers are provided for each valve to prevent the valve from being manually opened less than 10 seconds following valve closure. Valve opening is sensed by pressure switches in the valve discharge line. Each valve is provided with two trip, or actuation, systems. Each system is provided with two channels of instrumentation for each of the above described functions. A two-out-of-two-once logic scheme ensures that no single failure will defeat the low-low set function and no single failure will cause--

spurious operation of a safety/relief valve. Allowable deviations are provided for each specified instrument setpoint. Valve operation within the specified allowable deviations provide assurance that subsequent safety/relief valve actuations are sufficiently spaced to allow for discharge line water leg clearing.

3.2 BASES 69 09/12/02 Amendment No. 30, 31, 45, 15,101, !!8a, 129b

Bases 3.2 (Continued):

Control room habitability protection instrumentation assures that the control room operators will be adequately protected against the effects of accidental releases of radioactive leakage which may bypass secondary containment following a loss of coolant accident or radioactive releases from a steam line break accident, thus assuring that the Monticello Nuclear Generating Plant can be operated or shutdown safely.

Although the operator will set the setpoints within the trip settings specified in Tables 3.2.1 through 3.2.9, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. The deviations could be caused by inherent instrument error, operator setting error, drift of the set point, etc. Therefore, these deviations have been accounted for in-.

the various transient analyses.

3.2 BASES 69a 09/12/02 Amendment No. 65, 70, 89, !00a, 129b

Bases 3.2 (Continued):

Trip Function DeviaLtion Instrumentation f6r Safety/Relief Valve Reactor Coolant System +/-20 psig Low Low Set Logic Pressure for Opening/Closing Opening- Closing Pressure _>60 psi Discharge Pipe Pressure Inhibit +/- 10 psid "Timer Inhibit -3 sec

+10 sec Other Instrumentation

• High Reactor Water Level +6 inches

• Low-Low Reactor Water Level -3 inches

• This indication is reactor coolant temperature sensitive. The calibration is thus made for rated conditions. The level error at low pressures and temperatures is bounded by the safety analysis which reflects the weight-of-coolant above the lower tap, and not the indicated level.

I 3.2 BASES 71 a 09/12/02 Amendment No. 30, 37, 66, ! Goa, !05, 129b

Bases 3.6/4.6 (Continued):

H. Snubbers All safety related snubbers are required to be operable whenever the supported system is required to be operable. The Limiting Conditions for Operation is based on ensuring that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection I

program are those installed on non-safety related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval is determined by the sample population size and the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the samedesign feature directly related to rejection of the snubber.

When a snubber is found inoperable, an engineering evaluation or inspection is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the ,"

inoperability of the snubber. The evaluation or inspection will determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

3.6/4.6 BASES 154 09/12/02 Amendment No. 9, 4283,,4490a, 129a

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

D-1

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

A B

C 129b 129 115 I 36 37 38 128 128 128 71 a 72 76 129b 104 0

123-124 125 117 121 104 D 115 39 129b 77 86 126 104 E 115 40 129b 78 0 .126a 87 F 115 42 103 79 0 127 128 G 115 45 0 80 29 128 42 H

I J

119 129b 129b I 46 46a 47 70 37 40 81 82 82a 3

123 63 129 130 131 122 82 122 i 128 48 89 83 24 132 39 ii 104 49 128 83a 24 132a 122 iii 120 50 128 84 100a 133 i06 iv 128 50a 117 85 100a 134 106 v 120 51 117 86 100a 135 106 vi 121 51a 117 87 100a 136 106 vii 122 52 128 88 100a 137 0 1 119 53 128 89 104 138 100a 2 70 54 128 90 100a 145 118a 3 21 55 103 91 123 146 106 4 102 56 102 92 100a 147 107 5 120 57 70 93 122 148 117 5a 120 58 84 94 106 149 100a 6 128 58a 29 95 77 150 128 7 128 59 128 96 77 151 128 8 128 59a 103 97 57 153 100a 9 128 60 128 98 56 154 129a I 10 128 60a 31 99 104 155 122 11 128 60b 62 100 100a 156 93 12 128 60c 30 101 122 157 117 25a 127 60d 128 102 122 158 107 25b 127 60e 89 103 122 159 95 25c 127 61 104 104 122 160 95 25d 127 62 117 105 122 163 0 26 5 63 117 106 79 164 104 27 81 63a 117 107 97 165 64 27a 81 64 128 108 128 166 94 28 128 65 117 109 100a 167 112 29 128 66 119a 110 100a 168 94 30 103 67 117 111 122 169 94 31 104 68 129b 112 124a 170 122 32 103 69 129b 113 122 171 96 33 103 69a 129b 114 122 172 71 34 83 70 117 121 0 175 107 35 100a 71 100a 122 106 175a 117 A

Amendment No. 129b 09/12/02

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

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

200 129 236 115 201 129 243 128 202 129 244 124 203 204 204a 205 41 129 129 129 I 248 249 250 251 59 120 128 124 206 0 252 120 207 123 253 120 208 63 254 120 209 123 255 120 209a 100a 256 122 210 100a 257 122 211 109 258 122 212 109 259 120 213 99 260 120 216 100a 261 120 217 128 262 120 218 120 223 119 224 119 225 119 B

Amendment No. 129 08/27/02

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 waiyer 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 8/24173 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 3/30174 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 8/20174 Note 1 Changed byproduct material allowance 11 Note 3 Note 3 10/24/74 Inverted Tube (CRD) Limits 12 5 15 1/15175 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 7/8175 Note 1 Increased allowed quantity of U-235 16 12 20 9/15/75 Snubber Requirements 17 11 19 9/17175 Removed byproduct material allowance 18 13 21 10/6175 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 4/13176 Adoptioln 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 Ucense (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 o09 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 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 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 115a 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

I MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision (REV) 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 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

Amendment No. 129b 09/12/02

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

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. 129b I

09/12/02