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.

A violation 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. A reac-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 1 n0an -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:

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

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 sufficient number of devi*eS have been affected bh an" means-suc-h that the alutomAt*i fi-uncti-o IS incapable 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 speGif led.

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

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

3.

High Flux IRM

5.

High Reactor Pressure

6.

High Drywell Pressure Deviation

+2/125 of scale

+10 psi

+1 psi Trig Function

• 7. Reactor Low Water Level

8. Scram Discharge Volume High Level

9. Turbine Condenser Low Vacuum

• 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.

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 Deviation

-6 inches

+1 gallon

-1/2 in. Hg I

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.

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

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 I 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 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 Page No.

A 129b B

129 C

115 D

115 E

115 F

115 G

115 H

119 I

129b J

129b i

128 ii 104 iii 120 iv 128 v

120 vi 121 vii 122 1

119 2

70 3

21 4

102 5

120 5a 120 6

128 7

128 8

128 9

128 10 128 11 128 12 128 25a 127 25b 127 25c 127 25d 127 26 5

27 81 27a 81 28 128 29 128 30 103 31 104 32 103 33 103 34 83 35 100a I

I Amend Page No.

36 128 37 128 38 128 39 129b 40 129b 42 103 45 0

46 70 46a 37 47 40 48 89 49 128 50 128 50a 117 51 117 51a 117 52 128 53 128 54 128 55 103 56 102 57 70 58 84 58a 29 59 128 59a 103 60 128 60a 31 60b 62 60c 30 60d 128 60e 89 61 104 62 117 63 117 63a 117 64 128 65 117 66 119a 67 117 68 129b 69 129b 69a 129b 70 117 71 100a Amend Page No.

71 a 129b 72 104 76 0

77 86 78 0

79 0

80 29 81 3

82 123 82a 63 83 24 83a 24 84 100a 85 100a 86 100a 87 100a 88 100a 89 104 90 100a 91 123 92 100a 93 122 94 106 95 77 96 77 97 57 98 56 99 104 100 100a 101 122 102 122 103 122 104 122 105 122 106 79 107 97 108 128 109 100a 110 100a 111 122 112 124a 113 122 114 122 121 0

122 106 Amend Page No.

123-117 124 121 125 104 126 104

.126a 87 127 128 128 42 129 122 130 82 131 122 132 39 132a 122 133 i06 134 106 135 106 136 106 137 0

138 100a 145 118a 146 106 147 107 148 117 149 100a 150 128 151 128 153 100a 154 129a 155 122 156 93 157 117 158 107 159 95 160 95 163 0

164 104 165 64 166 94 167 112 168 94 169 94 170 122 171 96 172 71 175 107 175a 117 A

Amendment No. 129b 09/12/02 I

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

176 100a 177 117 178 100a 179 123a 180 128a 181 100a 182 112 183 117 184 100a 185 100a 188 104 189 100a 190 104 191 0

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

207 123 208 63 209 123 209a 100a 210 100a 211 109 212 109 213 99 216 100a 217 128 218 120 223 119 224 119 225 119 I I Amend Page No.

226 119 229a 63 229b 104 229c 104 229d 63 229e 122 229u 104 229v 112 229v v 112 229w 112 229ww 112 229x 112 229y 115a 229z 112 230 54 231 34 232 119 233 124 234 119 235 115 236 115 243 128 244 124 248 59 249 120 250 128 251 124 252 120 253 120 254 120 255 120 256 122 257 122 258 122 259 120 260 120 261 120 262 120 B

Amendment No. 129 08/27/02

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

Original Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 3

4 6/17/74 6

8/20174 Note 3 5

7 8

9 10 7/8175 12 11 13 14 15 1/22/76 16 2/3/76 17 3/16/76 NOTE 3 18 4/14/76 21 5/20/76 License DPR-22 Amend No. & Date 1

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

2/20[73 Note 2 Note 2 NOTE 4 4/13176 AEC Tech Spec Change Issuance "No. and date Note 1 2

1/14[72 3

10/31/72 4

12/8[72 Note 1 5

3/2/73 1

4/28/71 &

6 4/3/73 7

5/4/73 8

7/2/73 9

8/24173 10 10/2/73 11 11/27/73&

12 11/15/73 13 3/30174 14 5/14/74 Note 1.

Note 1 Note 3 10/24/74 15 1/15175 16 2/3/75 17 2/26/75 18 4/10/75 Note 1 20 9/15/75 19 9/17175 21 10/6175 22 10/30/75 C

Amendment No. 115 12/21/00 Major Subject Appendix-A Technical Specifications incorporated in DPR-22 on 9/8/70 Removed 5 MWt restriction MOGS Technical Specification changes issued by AEC but never distributed or put into effect, superseded by TS Change 12 11/15/73 RHR service water pump capability change Temporary surveillance test waiyer Increase in U-235 allowed in fission chambers Miscellaneous Technical Specification changes, Respiratory Protection, & Administrative Control Changes Respiratory Protection Changes Relief Valve and CRD Scram Time Changes Fuel Densification Limits Safety Valve Setpoint Change Offgas Holdup System, RWM, and Miscellaneous Changes 8x8 Fuel Load Authorization 8x8 Full Power authorization Changed byproduct material allowance Changed byproduct material allowance Inverted Tube (CRD) Limits REMP Changes Reactor Vessel Surveillance Program Changes Vacuum Breaker Test Changes Corrects Errors & Provides Clarification Increased allowed quantity of U-235 Snubber Requirements Removed byproduct material allowance Suppression Pool Temperature Limits Appendix K and GETAB Limits Reporting Requirements CRD Collet Failure Surveillance NSP Organization Changes Adoptioln of GETAB Containment Isolation Valve Testing Interim Appendix B, Section 2.4 Tech. Specs.

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 I

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

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

D Amendment No. 115 12/21/00 Major Subject Low Steamline Pressure Setpoint and MCPR Changes APLHGR, LHGR, MCPR Limits Correction of Errors and Environmental Reporting Standby Gas Treatment System Surveillance CRD Test Frequency Snubber Testing Changes APRS Test Method MAPLHGR Clamp at Reduced Flow Radiation Protection Supervisor Qualification REMP Changes More Restrictive MCPR Inservice Inspection Changes Reporting Requirements Fire Protection Requirements Increase in spent fuel storage capacity RPT Requirements Suppression Pool Surveillance 8x8R Authorization, MCPR Limits & SRV Setpoints Corrected Downcomer Submergence Incorporation of Physical Security Plan into License Revised LPCI Flow Capability Respiratory Protection Program Changes Fire Protection Safety Evaluation Report MAPLHGR vs. Exposure Table MCPR & MAPLHGR Changes for Cycle 8 and Extended Core Burnup° ILRT Requirements Order for Modification of License-Environmental Qualification Revised Order for Modification of License-Environmental Qualification Order for Modification of License-Environmental Qualification Records Issuance of Facility Operating Ucense (FTOL)

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

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

2/12/81 2

3/2/81 3

4 3/27/81 3/27/81 5

5/4/81 6

6/3/81 7

6/30/81 8

11/5/81 9

12/28/81 1/19/82 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 5/20/82 10/8/82 11/30/82 12/6/82 12/10/82 12/17/82 4/18/83 4/17/83 11/28/83 12/30/83 1/16/84 1/23/84 2/2/84 4/3/84 5/1/84 8/15/84 9/24/84 10/31/84 11/2/84 11/16/84 11/16/84 11/27/84 NSP Revision (REV) No.

NOTE 1 I

E Amendment No. 115 12/21/00 53 54 NOTE 1 Major Subject Order for Modification Mark I Containment Revision of License Conditions Relating to Fire Protection Modifications TMI Lessons Learned & Safety Related Hydraulic Snubber Additions Low voltage protection, organization and miscellaneous Incorporation of Safeguards Contingency Plan and Security Force Qualification and Training Plan into License Cycle 9 - ODYN Changes, New MAPLHGR's, RPS Response time change Inservice Inspection Program Fire Protection Technical Specification Changes Mark I Containment Modifications Inservice Surveillance Requirements for Snubbers Revised Order for Modification Mark I Containment Scram Discharge Volume New Scram Discharge Volumes RPS Power Monitor Cycle 10 Recirc Piping and Coolant Leak Detection Appendix I Technical Specifications (removed App. B)

Organizational Changes Miscellaneous Changes Steam Line Temperature Switch Setpoint Radiation Protection Program SRM Count Rate Definition of Operability Miscellaneous Technical Specification Changes RPS Electrical Protection Assembly Time Delay Scram Discharge Volume Vent and Drain Valves Miscellaneous Technical Specification Changes Cycle 11 RHR Intertie Line Addition Hybrid I Control Rod Assembly ARTS Low Low Set Logic Degraded Voltage Protection Logic 55 56 57 58 59 NOTE 1 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81

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

82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 o09 110 111 112 113 114 115 116 License DPR-22 Amend No. & Date 32 5/28/85 33 10/7/85 34 10/8/85 35 12/3/85 36 12/23/85 37 1/22/86 38 2/12/86 39 3/13/86 40 3/18/86 41 3/24/86 42 3/27/86 43 4/8/86 44 5/27/86 45 7/1/86 46 7/1/86 47 10/22/86 48 12/1/86 49 8/26/87 50 8/26/87 51 10/16/87 52 11/18/87 53 11/19/87 54 11/25/87 55 11/25/87 56 12/11/87 57 9/23/88 58 12/13/88 59 2/16/89 60 2/28/89 61 3/29/89 62 3/31/89 63 4/18/89 64 5/10/89 65 5/30/89 66 5/30/89 F

Amendment No. 115 12/21/00 Major Subject Surveillance Requirements Screen Wash/Fire Pump (Partial)

Fuel Enrichment Limits Combustible Gas Control System Vacuum Breaker Cycling NUREG-0737 Technical Specifications Environmental Technical Specifications Administrative Changes Clarification of Radiation Monitor Requirements 250 Volt Battery Jet Pump Surveillance Simmer Margin Improvement Cycle 12 Operation Miscellaneous Changes LER Reporting and Miscellaneous Changes Single Loop Operation Offgas System Trip Rod Block Monitor APRM and IRM Scram Requirements 2R Transformer Surveillance Intervals - ILRT Schedule Extension of Operating License Cycle 13 and Misc Changes Appendix J Testing ATWS - Enriched Boron Increased Boron Enrichment Physical Security Plan Miscellaneous Administrative Changes Miscellaneous Administrative Changes Fire Protection and Detection System ADS Logic and S/RV Discharge Pipe Pressure Miscellaneous Technical Specification Improvements Containment Vent and Purge Valves NUREG-0737 - Generic Letter 83-36 Reactor Vessel Level Instrumentation I

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

NSP License Revision DPR-22 (REV) No.

Amend No. & Date 117 67 6/19/89 118 68 7/14/89 119 69 9/12/89' 120 70 9/28/89 121 71 10/19/89 122 72 11/2/89 123 73 5/1/90 124 74 6/5/90 125 75 10/12/90 126 76 12/20/90 127 77 2/15/91 128 78 3/28/91 125 79 4/9/91 130 80 8/12/91 131 81 4/16/92 132 82 7/15/92 133 83 8/18/92 134 84 1/27/93 135 85 6/29/93 136 86 7/12/93 137 87 4/15/94 138 88 6/30/94 139 140 141 142 143 144 145 146 147 89 8/25/94 90 91 92 93 94 95 96 97 9/7/94 9/9/94 9/15/94 7/12/95 10/2/95 4/3/96 4/9/96 9/17/96 Major Subject Extension of MAPLHGR. Exposure for One Fuel Type SRO Requirements & Organization Chart Removal Operations Committee Quorum Requirements Relocation of Cycle-Specific Thermal-Hydraulic Limits Deletion of Primary Containment Isolation Valve Table RG 1.99, Rev 2, ISI & ILRT Combined STA/LSO Position Removal of WRGM Automatic ESF Actuation Diesel Fuel Oil Storage Miscellaneous Administrative Changes Redundant and IST Testing Alarming Dosimetry SAFER/GESTR Torus Vacuum Breaker Test Switch/EDG Fuel Oil Tank Level Surveillance Test Interval Extension - Part I Alternate Snubber Visual Inspection Intervals Revisions to Reactor Protection System Tech Specs MELLIA and Increase Core Flow Revision to Diesel Fire Pump Fuel Oil Sampling Requirements Revisions to Control Rod Drive Testing Requirements Revised Coolant Leakage Monitoring Frequency Average Planar Linear Heat Generation Rate (APLHGR)

Specification & Minimum Critical Power Ratio Bases Revisions Removal of Chlorine Detection Requirements and Changes to Control Room Ventilation System Requirements Revisions to Radiological Effluent Specifications Secondary Containment System and Standby Gas Treatment System Water Level Setpoint Change Change in Safety Relief Valves Testing Requirements Revised Core Spray Pump Flow Standby Gas Treatment and Secondary Containment Systems MSIV Combined Leakrate, and Appendix J, Option B Purge and Vent Valve Seal Replacement Interval Implementation of BRWOG Option I-D core Stability Solution and re-issue of pages 11, 12, 82 and 231 to reflect pages issued by NRC amendments.

I G

Amendment No. 115 12/21/00

NSP Revision (REV) No.

148 149 NOTE 5 150 NOTE 6 MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date Major Subject 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.

99 10/29/97 SLMCPR for Cycle 18 and reissue pages vi, 155, 202, 207, 219, 229u 11/25/97 Reissue pages a, b, g, iii, vi, 14, 25a, 155, 198y, 198z, 202, 207, 209, 219, 229k, 229p, 229r, 229u, 230, 245 100 4/20/98 SLMCPR for Cycle 19 1 00a 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

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

I Amendment No. 129b 09/12/02 I

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

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.

I J

Amendment No. 129b 09/12/02