Text

MNGP Technical Specification Bases Updating Instructions
	ML13261A146
Person / Time
Site:	Monticello
Issue date:	09/12/2013
From:	; Nuclear Management Co
To:	; Office of Nuclear Reactor Regulation
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TECHNICAL SPECIFICATION BASES UPDATING INSTRUCTIONS MNGP TECHNICAL SPECIFICATION BASES REMOVE Pane(s) Revision Document Type Table 1 25 List of Effective Sections I Specifications Table 2 25 Record of Revisions B 3.1.6-1 through 4 Bases B 3.1.6-4 Spec. 3.1.6 B 3.3.2.2-1 through 12 Bases B 3.3.2.2-13 Spec. 3.3.2.1 I I Destroy removed pages.

Revision 26

TABLE 1 (Page 1 of 1)

MONTICELLO NUCLEAR GENERATING PLANT BASES LIST OF EFFECTIVE SECTIONS/SPECIFICATIONS Section/Specification Revision No. Section/Specification Revision No.

B 2.1.1 4 B 3.6.1.4 0 B 2.1.2 6 B 3.6.1.5 25 B 3.0 14 B 3.6.1.6 0 B 3.1.1 0 B 3.6.1.7 0 B3.1.2 0 B 3.6.1.8 0 B 3.1.3 11 B 3.6.2.1 0 B 3.1.4 19 B 3.6.2.2 0 B 3.1.5 0 B 3.6.2.3 0 B 3.1.6 26 B 3.6.3.1 0 B 3.1.7 4 B 3.6.4.1 0 B 3.1.8 4 B 3.6.4.2 0 B 3.2.1 0 B 3.6.4.3 0 B 3.2.2 0 B 3.7.1 0 B 3.2.3 0 B 3.7.2 0 B 3.3.1.1 24 B 3.7.3 17 B 3.3.1.2 0 B 3.7.4 15 B 3.3.2.1 26 B 3.7.5 8 B 3.3.2.2 0 B 3.7.6 4 B 3.3.3.1 3 B 3.7.7 16 B 3.3.3.2 3 B 3.7.8 0 B 3.3.4.1 0 B 3.8.1 19 B 3.3.5.1 23 B 3.8.2 19 B 3.3.5.2 0 B 3.8.3 1 B 3.3.6.1 4 B 3.8.4 7 B 3.3.6.2 0 B 3.8.5 4 B 3.3.6.3 3 B 3.8.6 0 B 3.3.7.1 4 B 3.8.7 0 B 3.3.7.2 4 B 3.8.8 4 B 3.3.8.1 22 B 3.3.8.2 0 B 3.9.1 0 B 3.4.1 12 B 3.9.2 0 B 3.4.2 0 B 3.9.3 0 B 3.4.3 25 B 3.9.4 0 B 3.4.4 0 B 3.9.5 0 B 3.4.5 0 B 3.9.6 0 B 3.4.6 4 B 3.9.7 21 B 3.4.7 0 B 3.9.8 0 B 3.4.8 0 B 3.10.1 0 B 3.4.9 25 B 3.10.2 0 B 3.4.10 0 B 3.10.3 3 B 3.5.1 25 B 3.10.4 0 B 3.5.2 0 B 3.10.5 0 B 3.5.3 0 B 3.10.6 0 B 3.6.1.1 0 B 3.10.7 0 B 3.6.1.2 0 B 3.10.8 0 B 3.6.1.3 18 Rev. 26

TABLE 2 (Page 1 of 4)

TECHNICAL SPECIFICATION BASES RECORD OF REVISIONS Revision Affected Description of Revision Number Section/

Specification 0 All Amendment 146 - Original ITS Revision 1 B 3.8.3 SR 3.8.3.3, Diesel Fuel Oil Testing Description 2 B 3.5.1 LCO 3.5.1, ACTION D, changed description of LPCI injection pathway.

3 B 3.3.3.1, Miscellaneous ITS Bases B 3.3.3.2, 'Clarifications/Corrections B 3.3.6.3, B 3.10.3 4 B 2.1.1, Amendment 148 - Bases Changes B 2.1.2, implementing Full Scope AST.

B 3.1.6, B 3.1.7, B 3.1.8, B 3.3.6.1, B 3.3.7.1, B 3.3.7.2, B 3.4.6, B 3.6.1.3, B 3.7.4, B 3.7.5, B 3.7.6, B 3.8.2, B 3.8.5, B 3.8.8 5 B 3.3.5.1 Amendment 151 - Extend Surveillance Interval B 3.8.1(1) and AV for the LPCI Loop Select TD Relays.

6 B 2.1.2 Clarify RCS Safety Limit values.

B 3.3.2.1 Correct that initial MCPR values are specified in the COLR.

7 B.3.8.1, Clarify that the 2R and 1AR transformers are B 3.8.2 considered as a single off-site source when 1AR is supplied from 345 kV Bus 1. 1

1. Replaces page B 3.8.1-25 in Sharepoint version of the TS. Page inadvertently deleted during implementation of Amendment 148 (CAP 01095053).

Rev. 26

TABLE 2 (Page 2 of 4)

TECHNICAL SPECIFICATION BASES RECORD OF REVISIONS Revision Affected Description of Revision Number Section/

Specification 7 (con't) B.3.8.4 Correct the float voltage for the 125 VDC batteries in SR 3.8.4.1.

B.3.8.4 Amendment 153 - Specify in SR 3.8.4.2 that the Division 2 battery charger supplies

_-110 amps.

8 B.3.5.1 Amendment 155 - Revise SR 3.5.1.3 to correct Alternate Nitrogen System supply pressure to ADS and clarify OPERABILITY during bottle changeout.

B.3.7.4, Amendment 154 - Revise Bases for B.3.7.5 Specification 3.7.5 to reflect adoption of TSTF-477, which allows both CRV subsystems to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Clarify the OPERABILITY requirements of certain CRV fans currently required to support CREF subsystem operation.

9 B 3.5.1 Clarify RHR intertie discussion.

B 3.5.1 Clarify Action M to indicate that the plant may not be in a condition outside the accident analysis but is in a condition not specifically justified for continued operation.

B 3.6.1.3 Add Action E to describe actions for when the MSIVs are not within leakage limits and re-label subsequent actions.

10 B 3.5.1 Add HPCI "Keep-fill" discussion.

B 3.9.7 Clarify Action A.1 for what is meant by inoperable.

11 B 3.1.3, Amendment 158- Change control rod notch B 3.1.4 testing frequency from every 7 days to only once per 31 days in accordance with TSTF-475, Revision 1.

Rev. 26

TABLE 2 (Page 3 of 4)

TECHNICAL SPECIFICATION BASES RECORD OF REVISIONS Revision Affected Description of Revision Number Section/

Specification 12 B 3.3.1.1, Amendment 159- PRNMS.

B 3.3.2.1, B 3.4.1 B 3.3.5.1 Amendment 161 - LPCI Recirculation Riser Differential Pressure - High (Break Size) allowable value and channel calibration interval change.

13 B 3.0 Amendment 157 - Add Bases for new LCO 3.0.9 for the unavailability of barriers, reflecting adoption of TSTF-427.

B 3.4.9 Clarify that the shift in Figure 3.4.9-1 includes both delta RTNDT and margin.

B 3.5.1 Amendment 162 -Add new Conditions to Specification 3.5.1 for restoration of various low-pressure ECCS subsystem out-of-service combinations.

14 B 3.0 Section B 3.0 reissued in entirety. Page numbers at end of LCO Applicability over-lapped SR Applicability page numbers (CAP 01192534).

15 B 3.7.4 Amendment 160 - Revise Bases for the specification reflecting adoption of a Control Room Envelope Habitability program in accordance with TSTF-448.

16 B 3.3.1.1 Replace IRM - Neutron Flux - High High (1.a) for calibrating IRMs by a heat balance by referring to IRM/APRM overlap and APRM Setdown Scram meeting reactivity requirements.

B 3.7.7 Correct Turbine Bypass Valve capacity.

17 B 3.7.3 Correct EDG-ESW Background description.

Rev. 26

TABLE 2 (Page 4 of 4)

TECHNICAL SPECIFICATION BASES RECORD OF REVISIONS Revision Affected Description of Revision Number Section/

Specification 18 B 3.6.1.3 Clarify PCIV definition.

19 B 3.1.4 Clarify spacing requirements of adjacent slow control rods.

B 3.8.1, Revise Bases to reflect separation of MARS B 3.8.2 Transformer and Bus 1.

20 B 3.5.1 Reissue section, missing text on HPCI keep-fill discussion. (CAP 01328422) 21 B 3.9.7 Correct prior clarification to Action A.1 for what is meant by inoperable. (CAP 01257096) 22 B 3.3.8.1 Amendment 169 - Revise licensing basis to reflect removal of the capability to automatically transfer to the IAR Transformer as a source of power to the essential buses on degraded voltage and instead directly transfer to the EDGs.

23 B 3.3.5.1 Amendment 170- Revised to reflect ancillary change related to ADS 20-minute Bypass Timer.

24 B 3.3.1.1 Amendment 171 - Revised to provide restoration period before declaring the APRMs inoperable when SR 3.3.1.1.2 is not met.

25 B 3.4.3, Amendment 168 - Revised surveillance B 3.5.1, requirements within these specifications to B 3.6.1.5 allow crediting overlapping testing rather than requiring a lift-test during plant startup.

B 3.4.9 Amendment 172- Revised specification to adopt PTLR.

26 B 3.1.6 Amendment 173 - Revised to reflect B 3.3.2.1 incorporation of TSTF-376 for improved BPWS.

Rev. 26

Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP. The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1, 2, and 3.

APPLICABLE The analytical methods and assumptions used in evaluating the CRDA SAFETY are summarized in References 1, 2, and 3. CRDA analyses assume that ANALYSES the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis.

The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity.

Since the failure consequences for U0 2 have been shown to be insignificant below fuel energy depositions of 300 cal/gm (Ref. 4), the fuel design limit of 280 cal/gm provides a margin of safety from significant core damage which would result in release of radioactivity (Ref. 5).

Generic evaluations (Refs. 6 and 7) of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal/gm) have shown that if the peak fuel enthalpy remains below 280 cal/gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref..8) and the calculated offsite doses will be well within the required limits (Ref. 9).

Control rod patterns analyzed in Reference I follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation.

Monticello B 3.1.6-1 Revision No. 26

Rod Pattern Control B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued)

Generic analysis of the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel design limit will not be violated during a CRDA while following the BPWS mode of operation. The generic BPWS analysis (Ref. 10) also evaluates the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e.,

eight) and distribution of fully inserted, inoperable control rods.

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 11) may be used. Before reducing power to the low power setpoint (LPSP), control rod coupling integrity shall be confirmed for all rods that are fully withdrawn. Control rods that have not been confirmed coupled and which are in intermediate positions must be fully inserted prior to power reduction to the LPSP. No action is required for fully-inserted control rods. If a shutdown is required and all rods which are not confirmed coupled cannot be fully inserted prior to the power dropping below the LPSP, then the original BPWS must be adhered to.

The rods may be fully inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled (Ref. 11). It is recommended that control rods be inserted in the same order as specified for the original BPWS as much as reasonably possible. When in the process of shutting down following optional BPWS with the power below the LPSP, no control rod shall be withdrawn unless the control rod pattern is in compliance with original BPWS requirements.

When using the Reference 11 control rod sequence for shutdown, the rod worth minimizer may be bypassed in accordance with the allowance provided in the Applicability Note for the Rod Worth Minimizer in Table 3.3.2.1-1.

In order to use the Reference 11 BPWS shutdown process, an extra check is required in order to consider a control rod to be "confirmed" to be coupled. This extra check ensures that no Single Operator Error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Details on this coupling confirmation requirement are provided in Reference 11. If the requirements for use of the BPWS control rod insertion process contained in Reference 11 are followed, the plant is considered to be in compliance with BPWS requirements, as required by LCO 3.1.6.

Rod pattern control satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS analysis. Compliance with the optional BPWS control rod insertion process prevents a CRDA from occurring.

This LCO only applies to OPERABLE control rods. For inoperable control Monticello B 3.1.6-2 Revision No. 26

Rod Pattern Control B 3.1.6 BASES LCO (continued) rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

APPLICABILITY In MODES 1 and 2, when THERMAL POWER is < 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel design limit during a CRDA (Ref. 2). In MODES 3 and 4, the reactor is shut down and the control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied, therefore a CRDA is not postulated to occur. In MODE 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to < 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence.

Required Action A.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator (Operator or Senior Operator) or by a qualified member of the technical staff (e.g., engineer). This helps to ensure that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.

Monticello B 3.1.6-3 Revision No. 26

Rod Pattern Control B 3.1.6 BASES ACTIONS (continued)

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator (Operator or Senior Operator) or by a qualified member of the technical staff (e.g.,

engineer).

When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at

< 10% RTP.

Monticello B 3.1.6-4 Revision No. 26

Rod Pattern Control B 3.1.6 BASES REFERENCES 1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel" (revision specified in Specification 5.6.3).

2. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),

"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," BWROG-8644, August 15, 1986.

3. USAR, Section 14.7.1.

4. NUREG-0979, Section 4.2.1.3.2, April 1983.

5. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

7. NEDO-10527, "Rod Drop Accident Analysis for Large BWRs,"

(including Supplements 1 and 2), March 1972.

8. ASME, Boiler and Pressure Vessel Code.

9. 10 CFR 50.67.

10. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

11. NEDO-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

Monticello B 3.1.6 Last Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes.

Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions; control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.

The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one of the four redundant average power range monitor (APRM) channels supplies a reference signal for one of the RBM channels and a signal from another of the APRM channel supplies the reference signal to the second RBM channel. This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 1). Furthermore, the Bypass Time Delay, which bypasses the RBM upscale trips for a short period of time, is not utilized (it is permanently disabled). Thus, if it is not disabled, the associated RBM channel is inoperable. In addition, to preclude rod movement with an inoperable RBM, an inoperable trip is provided. A RBM channel is considered inoperable if less than half the total number of inputs are available.

Revision No. 26 B 3.3.2.1-1 Monticello Monticello B 3.3.2. 1-1 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND (continued)

The purpose of the RWM is to control rod patterns during startup, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits.

With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.

APPLICABLE Allowable Values are specified for each applicable Rod Block Function SAFETY listed in Table 3.3.2.1-1. The nominal trip setpoints (NTSPs) (actual trip ANALYSES, LCO, setpoints) are selected to ensure that the setpoints are conservative with and APPLICABILITY respect to the Allowable Value. A channel is inoperable if its actual trip setpoint selected is non-conservative with respect to its required Allowable Value.

NTSPs are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,

trip unit) changes state. The Analytical Limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the Analytical Limits, corrected for calibration, process, and some of the instrument errors. The NTSPs are then determined, accounting for the remaining channel uncertainties. The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, and drift are accounted for. The Limiting Trip Setpoint is the value of the setpoint within its specified as-found tolerance which most closely approaches the Allowed Value. For the Rod Block Monitor, which is a digital system with a zero as-found tolerance, the Limiting Trip Setpoint is the NTSP.

Monticello B 3.3.2.1-2 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Rod Block Monitor Low, Intermediate and High Power Range -

Upscale functions (Functions Ia, lb and lc, respectively) are Limiting Safety System Setting (LSSS), SL-related, as determined in the NRC Safety Evaluation for Amendment 159 (Ref. 12).

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

1. Rod Block Monitor The RBM is designed to prevent violation of the MCPR SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The NTSP and Allowable Values are chosen as a function of power level. NTSP operating limits are established based on the specified Allowable Values.

The RBM Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block from this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.

NTSPs are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the NTSP, but within its Allowable Value, is acceptable. NTSPs are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter, the calculated RBM flux (RBM channel signal).

When the normalized RBM flux value exceeds the applicable trip setpoint, the RBM provides a trip output. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values and NTSPs are derived, using the General Electric setpoint methodology guidance, as specified in the Monticello setpoint methodology. The Allowable Values are derived from the analytic limits. The difference between the analytic limit and the Allowable Value allows for channel instrument accuracy, calibration accuracy, process measurement accuracy, and primary element Monticello B 3.3.2.1-3 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) accuracy. Use of this method and verification provides the assurance that ifthe setpoint is found conservative to the Allowable Value during surveillance testing, the instrumentation would have provided the required trip function by the time the process reached the analytic limit for the applicable events, thereby protecting the SL.

For the digital RBM, there is a normalization process initiated upon rod selection, so that only RBM input signal drift over the interval from the rod selection to rod movement needs to be considered in determining the nominal trip setpoints. The RBM has no drift characteristic with no as-left or as-found tolerances since it only performs digital calculations on the digitized input signals provided by the APRMs.

The NTSP (or Limiting Trip Setpoint) is the LSSS since the RBM has no drift characteristic. The RBM Allowable Value demonstrates that the analytic limit would not be exceeded, thereby protecting the safety limit.

The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and environment errors are accounted for and appropriately applied for the RBM. There are no margins applied to the RBM nominal trip setpoint calculations which could mask RBM degradation.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 30% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 3). When operating < 90% RTP, analyses have shown that with an initial MCPR > the cycle and power specific limit specified in the current COLR, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at > 90% RTP with MCPR Ž the cycle and power specific limit specified in the current COLR, no RWE event will result in exceeding the MCPR SL. Therefore, under these conditions, the RBM is also not required to be OPERABLE.

2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated.

The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, 7 and 14. The standard BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattern Control."

Monticello B 3.3.2.1-4 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 14) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions. When using the Reference 14 control rod insertion sequence for shutdown, the rod worth minimizer may be bypassed if it is not programmed to reflect the optional BPWS shutdown sequence, as permitted by the Applicability Note for the RWM in Table 3.3.2.1-1.

The RWM Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Since the RWM is a system designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY,"

and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is < 10% RTP.

When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 and 7). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Monticello B 3.3.2.1-5 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.

During shutdown conditions (MODES 3 and 4, and MODE 5 when the reactor mode switch is in the shutdown position), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position; the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2, "Refuel Position One-Rod-Out Interlock")

provides the required control rod withdrawal blocks.

ACTIONS A.1 With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.

B._1 If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.

Revision No. 26 B 3.3.2.1-6 Monticello Monticello B 3.3.2.1-6 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS (continued)

C.1, C.2.1.1, C.2.1.2, and C.2.2 With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram. Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM was not performed in the last 12 months. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications.

Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Operator or Senior Operator) or other qualified member of the technical staff (engineer).

The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.

D. 1 With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Operator or Senior Operator) or other qualified member of the technical staff (engineer). The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.

E.1 and E.2 With one Reactor Mode Switch - Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function.

However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch - Shutdown Position Monticello B 3.3.2.1-7 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS (continued)

Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.

In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1, "SHUTDOWN MARGIN (SDM)." Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Control Rod Block REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.2.1-1.

The Surveillances are modified by a second Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.

SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control System input. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.

This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Revision No. 26 B 3.3.2.1-8 Monticello B 3.3.2.1-8 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 184 days is based on reliability analyses (Ref. 10).

SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.

This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the RWM is performed by: a) attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs; b) verifying proper annunciation of the selection error of at least one out-of-sequence control rod in each fully inserted group; and c) performing a RWM computer on-line diagnostic test. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at < 10% RTP in MODE 2, and SR 3.3.2.1.3 is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is < 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Frequencies are based on reliability analysis (Ref. 8).

SR 3.3.2.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.6.

Monticello B 3.3.2.1-9 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)

The Frequency is based upon the assumption of a 24 month calibration interval (Refs. 10 and 11).

SR 3.3.2.1.4 for the following RBM functions is modified by two Notes as identified in Table 3.3.2.1-1. These functions, in accordance with the guidance of Regulatory Issue Summary 2006-17 (Ref. 13) and as determined in the NRC Safety Evaluation for Amendment 159 (Ref. 12),

are LSSS SL-related.

Function No. RBM Function 1.a Low Power Range - Upscale 1 .b Intermediate Power Range - Upscale 1 .c High Power Range - Upscale Note (h) requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. For digital channel components, no as-found tolerance or as-left tolerance can be specified.

Evaluation of instrument performance will verify that the instrument will continue to behave in accordance with design basis assumptions. The purpose of the assessment is to ensure confidence in the instrument performance prior to returning the instrument to service. This nonconformance will be entered into the Corrective Action Program.

Entry into the Corrective Action Program will ensure required review and documentation of the condition for continued OPERABILITY.

Note (i) requires that the as-left setting for the instrument be returned to the NTSP. If the as-left instrument setting cannot be returned to the NTSP, then the instrument channel shall be declared inoperable. The NTSPs and Allowable Values for Rod Block Monitor Functions 1 .a, 1 .b and 1.c are specified in the COLR. The methodology used to determine the NTSPs are specified in Appendix C to the Technical Requirements Manual, a document controlled under 10 CFR 50.59.

SR 3.3.2.1.5 The RBM setpoints are automatically varied as a function of power. The three control rod block Allowable Values required in Table 3.3.2.1-1, each within a specific power range, are specified in the COLR. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These control rod block bypass setpoints must be verified periodically to be less than or equal to the specified values. If any power range setpoint is Monticello B 3.3.2. 1-10 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e., enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.6. The 24 month Frequency is based on the actual trip setpoint methodology utilized for these channels.

SR 3.3.2.1.6 The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be verified periodically to be > 10% RTP.

If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The 24 month Frequency is based on engineering judgment considering the reliability of the components, and that indication of whether or not the RWM is bypassed is provided in the control room.

SR 3.3.2.1.7 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch - Shutdown Position Function to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch - Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.

As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links.

Monticello B 3.3.2. 1-11 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This allows entry into MODES 3 and 4 if the 24 month Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.

REFERENCES 1. USAR, Section 7.3.5.3.

2. USAR, Section 7.8.2.

3. NEDC-30492-P, "Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvements (ARTS) Program for Monticello Nuclear Generating Plant," April 1984.

4. NEDE-2401 1-P-A, "General Electrical Standard Application for Reload Fuel" (revision specified in Specification 5.6.3).

5. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),

"Amendment 17 to General Electric Licensing Topical Report NEDE-2401 1-P-A," BWROG-8644, August 15, 1986.

6. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

7. NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.

Monticello B 3.3.2.1-12 Revision No. 26

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES (continued)

8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.

9. GENE-770-06-1-A, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," December 1992.

10. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," October 1995.

11. NEDC-3241OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," November 1987.

12. Amendment No. 159, "Issuance of Amendment Re: Request to Install Power Range Neutron Monitoring System, dated February 3, 2009. (ADAMS Accession No. ML083440681)

13. U.S. NRC Regulatory Issue Summary 2006 17, "NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications,"

Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006.

14. NEDC-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

Revision No. 26 Monticello Monticello B 3.3.2.1-13 Last B 3.3.2.1 Last Revision No. 26

OPERATING LICENSE AND TECHNICAL SPECIFICATIONS UPDATING INSTRUCTIONS TECHNICAL SPECIFICATIONS REMOVE INSERT'...

Paqe Amendment Document Type Page 'Amendment 3 172 Operating 3 17.3.

License Table 1 172 Operating Table 1.: 173 (3 pages) License TS LEFPand (3pages)

Table 2 172 Record of TS Table 2 :173 (13 pages) Changes and (13 pages)

License Amendments 3.3.2.1-5 159 Specification 3.3.2.1-5 173 .

3.3.2.1 (Do not insert in TS binder, only an updating aid.)

Monticello Am 173

2. Pursuant to the Act and 10 CFR Part 70,. NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and -amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated. August 16, 1974 (those portions dealing with handling of reactor fuel) and August 17, 1977 (those portions dealing with fuel assembly storage capacity);

3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

5. Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C. This renewed operating license shall be.deemed to contain and is subject to the conditions specified in the Commission's regulations i.n.10 CFR Chapter I and is subject to all applicable provisions of the Act and to the'rules, regulations, and.

orders of the Commission, now OF hereafter in effect; and is subject to the additional conditions specified or incorporated bel.ow:

1. Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1775 megawatts (thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 173, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

3. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No, 1-thru 173

TABLE 1 (Page 1 of 3)

. MONTICELLO NUCLEAR GENERATING PLANT OPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES ace Amend. Page Amend. Page Amend. Page Amend.

No. No. No. No.

1 _ (2) 3.1.4-2 146 3.3.5.1-1 146 3.4.5-2 146 i _ (2) 3.1.4-3 158 3.3.5.1-2 146 3.4.6-1 148 ii __ (2) 3.1.5-1 146 3.3.5.1-3 146 3.4.6-2 148 1.1-1 146 3.1.5-2 146 3.3.5.1-4 146 3.4.7-1 146 1.1-2 148 3.1.5-3 146 3.3.5.1-5 151 3.4.7-2 146 1.1-3 146 3.1.6-1 146 3.3.5.1-6 170 3.4.8-1 146 1.1-4 172 3.1.6-2 146 3.3.5.1-7 146 3.4.8-2 146

_ (4) 1.1-4a 1.1.5 146 3.1.7-1 148 3.3.5.1-8 170(1) 3.4.9-1 172 1.1-6 146 3.1.7-2 146 3.3.5.1-9 161 3.4.9-2 172 1.2-1 146 3.1.7-3 146 3.3.5.1-10 146 3.4.9-3 172 1.2-2 146 3.1.7-4 146 3.3.5.1-11 146 1.3-1 146 3.1.7-5 146 3.3.5.2-1 146 1.3-2 146 3.1.7-6 146 3.3.5.2-2 146 1.3-3 146 3.1.8-1 146 3.3.5.2-3 146 1.3-4 146 3.1.8-2 146 3.3.5.2-4 146 3.4.10-1 146 1.3-5 146 3.2.1-1 146 3.3.6.1-1 146 3.5.1-1 146 1.3-6 146 3.2.2-1 146 3.3.6.1-2 146 3.5.1-2 162"5) 1.3-7 146 3.2.2-2 146 3.3.6.1-3 146 3.5.1-3 162 1.3-8 146 3.2.3-1 146 3.3.6.1-4 146 3.5.1-4 162 1.3-9 146 3.3.1.1-1 171 3.3.6.1-5 146 3.5.1-5 162 1.3-10 146 3.3.1.1-2 171 3.3.6.1-6 146 3.5.1-6 167 3.5.1-7 168 1.3-11 146 3.3.1.1-3 159 3.3.6.1-7 164 3.5.2-1 146 1.4-1 146 3.3.1.1-4 159 3.3.6.2-1 146 3.5.2-2 146 1.4-2 146 3.3.1.1-5 159 3.3.6.2-2 146 3.5.2-3 167 1.4-3 146 3.3.1.1-6 159 3.3.6.2-3 146 3.5.3-1 146 1.4-4 158 3.3.1.1-7 159 3.3.6.3-1 146 3.5.3-2 146 1.4-5 146 3.3.1.1-8 159 3.3.6.3-2 146 3.6.1.1-1 146 1.4-6 146 3.3.1.1-9 159 3.3.6.3-3 146 3.6.1.1-2 146 1.4-7 146 3.3.1.2-1 146 3.3.7.1-1 148 3.6.1.2-1 146 2.0-1 165 3.3.1.2-2 146 3.3.7.1-2 148 3.6.1.2-2 146 3.3.1.2-3 146 3.3.7.1-3 148 3.3.1.2-4 146 3.3.7.2-1 148(3) 3.3.7.2-2 148(3) 3.0-1 157 3.3.1.2-5 146 3.3.8.1-1 146 3.6.1.2-3 146 3.0-2 146 3.3.2.1-1 146 3.3.8.1-2 146 3.6.1.2-4 146 3.0-3 157 3.3.2.1-2 146 3.3.8.1-3 147 3.6.1.3-1 148 3.0-4 146 3.3.2.1-3 159 3.3.8.2-1 146 3.6.1.3-2 146 3.0-5 146 3.3.2.1-4 159 3.3.8.2-2 146 3.6.1.3-3 148 3.1.1-1 146 3.3.2.1-5 173 3.3.8.2-3 146 3.6.1.3-4 146 3.1.1-2 146 3.3.2.2-1 146 3.4.1-1 159 3.6.1.3-5 148 3.1.1-3 146 3.3.2.2-2 146 3.4.1-2 159 3.6.1.3-6 148 3.1.2-1 146 3.3.3.1-1 146 3.6.1.3-7 146 3.1.2-2 146 3.3.3.1-2 146 3.4.2-1 146 3.6.1.3-8 148

.3.1.3-1 146 3.3.3.1-3 146 3.4.3-1 146 3.6.1.4-1 146 3.1.3-2 158 3.3.3.2-1 146 3.4.3-2 168 3.6.1.5-1 146 3.1.3-3 158 3.3.4.1-1 146 3.4.4-1 146 3.6.1.5-2 168 3.1.3-4 158 3.3.4.1-2 146 3.4.4-2 146 3.6.1.6-1 146 3.1.4-1 146 3.3.4.1-3 146 3.4.5-1 146 3.6.1.6-2 146 Am. 173

TABLE I (Page 2 of 3)

MONTICELLO NUCLEAR GENERATING PLANT OPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES Page Amend. Page Amend. Page Amend.

No. No. No.

3.6.1.7-1 146 3.8.2-2 148 3.10.6-1 146 3.6.1.7-2 146 3.8.2-3 146 3.10.6-2 146 3.6.1.8-1 146 3.8.3-1 146 3.10.7-1 146 3.6.1.8-2 146 3.8.3-2 146 3.10.7-2 146 3.6.2.1-1 146 3.8.3-3 146 3.10.8-1 146 3.6.2.1-2 146 3.8.4-1 146 3.10.8-2 146 3.6.2.1-3 146 3.8.4-2 153 3.10.8-3 146 3.6.2.2-1 146 3.8.4-3 146 4.0-1 146 3.6.2.3-1 146 3.8.5-1 148 4.0-2 150 3.6.2.3-2 146 3.8.5-2 146 5.1-1 146 3.6.3.1-1 146 3.8.6-1 146 5.2-1 146 3.6.4.1-1 146 3.8.6-2 146 5.2-2 146 3.6.4.1-2 146 3.8.6-3 146 5.3-1 146 3.6.4.2-1 146 3.8.6-4 146 5.4-1 146 3.6.4.2-2 146 3.8.7-1 146 5.5-1 146 3.6.4.2-3 146 3.8.7-2 146 5.5-2 163 3.6.4.3-1 146 3.8.8-1 148 5.5-3 146 3.6.4.3-2 146 3.8.8-2 146 5.5-4 146 3.7.1-1 146 3.9.1-1 146 5.5-5 146 3.7.1-2 146 3.9.1-2 146 5.5-6 146 3.7.2-1 146 3.9.2-1 146 5.5-7 146 3.7.2-2 146 3.9.3-1 146 5.5-8 146 3.7.3-1 146 3.9.4-1 146 5.5-9 146 3.7.4-1 146 3.9.4-2 146 5.5-10 148 3.7.4-2 146 3.9.5-1 146 5.5-11 148 3.7.4-3 146 3.9.6-1 146 5.6-1 146 3.7.5-1 154 3.9.7-1 146 5.6-2 159 5.6-3 172 3.7.5-2 154 3.9.7-2 146 5.7-1 146 3.7.6-1 146 3.9.8-1 146 5.7-2 146 3.7.6-2 146 3.9.8-2 146 5.7-3 146 3.7.7-1 146 3.10.1-1 146 5.7-4 146 3.7.8-1 146 3.10.1-2 146 3.8.1-1 146 3.10.1-3 146 Operating License 3.8.1-2 146 3.10.2-1 146 1 156 3.8.1-3 146 3.10.2-2 146 2 156 3.8.1-4 146 3.10.3-1 146 3 173 3.8.1-5 146 3.10.3-2 146 4 166 3.8.1-6 146 3.10.3-3 146 5 156 3.8.1-7 146 3.10.4-1 146 6 169 3.8.1-8 146 3.10.4-2 146 6A 169 3.8.1-9 146 3.10.4-3 146 7 160 3.8.1-10 146 3.10.5-1 146 3.8.2-1 148 3.10.5-2 146 Am. 173

TABLE 1 (Page 3 of 3)

MONTICELLO NUCLEAR GENERATING PLANT OPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES NOTES NOTE 1: Am. 146 corrected by letter dated 10/12/06.

NOTE 2: Am. 152 removed Table of Contents (TOC) from NRC issued Tech. Specs. TOC at Revision 0.

NOTE 3: Am. 148 corrected by letter dated 3/9/2009.

NOTE 4: Am. 172 inadvertently removed two definitions. Page 1.1-4a (not part of TS) inserted to provide definitions until an amendment can be processed and received.

NOTE 5: Condition F in Specification 3.5.1 (Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with both Core Spray subsystems inoperable) is annotated and is being treated as a non-conservative TS (see AR 01372600).

Am. 173

TABLE 2 (Page 1 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License AEC Tech Spec Maior Subeect Revision DPR-22 Change Issuance (REV) No. Amend No. & Date No. & Date Original Appendix A Technical Specifications incorporated in DPR-22 on 9/8/70 1 1/19/71 Note 1 Removed 5 MWt restriction Note 2 2 1/14/72 MOGS Technical Specification changes issued by AEC but never distributed or put into effect, superseded by TS Change 12 11/15/73 1 Note 2 3 10/31/72 RHR service water pump capability change Note 2 4 12/8/72 Temporary surveillance test waiver 2 2/20/73 Note 1 Increase in U-235 allowed in fission chambers 2 Note 2 5 3/2/73 Miscellaneous Technical Specification changes, 3 Note 2 1 4/28/71 & Respiratory Protection, & Administrative 6 4/3/73 Control Changes 4 Note 2 7 5/4/73 Respiratory Protection Changes 5 Note 2 8 7/2/73 Relief Valve and CRD Scram Time Changes 6 Note 2 9 8/24/73 Fuel Densification Limits 7 Note 2 10 10/2/73 Safety Valve Setpoint Change 8 Note 2 11 11/27/73& Offgas Holdup System, RWM, and 12 11/15/73 Miscellaneous Changes 9 Note 2 13 3/30/74 8x8 Fuel Load Authorization 10 3 14 5/14/74 8x8 Full Power authorization 4 6/17/74 Note 1 Changed byproduct material allowance 6 8/20/74 Note 1 Changed byproduct material allowance 11 Note 3 Note 3 10/24/74 Inverted Tube (CRD) Limits 12 5 15 1/15/75 REMP Changes 13 7 16 2/3/75 Reactor Vessel Surveillance Program Changes 14 8 17 2/26/75 Vacuum Breaker Test Changes 15 9 18 4/10/75 Corrects Errors & Provides Clarification 10 7/8/75 Note 1 Increased allowed quantity of U-235 16 12 20 9/15/75 Snubber Requirements 17 11 19 9/17/75 Removed byproduct material allowance 18 13 21 10/6/75 Suppression Pool Temperature Limits 19 14 22 10/30/75 Appendix K and GETAB Limits Am. 173

TABLE 2 (Page 2 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subject Revision DPR-22 (REV) No. Amend No. & Date 20 15 1/22/76 NOTE 4 Reporting Requirements 21 16 2/3/76 CRD Collet Failure Surveillance 22 17 3/16/76 NSP Organization Changes 23 NOTE 3 4/13/76 Adoption of GETAB 24 18 4/14/76 Containment Isolation Valve Testing 25 21 5/20/76 Interim Appendix B, Section 2.4 Tech. Specs.

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 30 31 23 24 25 9/27/76 10/15/76 10/27/76 Standby Gas Treatment System Surveillance CRD Test Frequency 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 Am. 173

TABLE 2 (Page 3 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subject Revision DPR-22 (REV) No. Amend No. & Date 48 42 12/28/79 MAPLHGR vs. Exposure Table 49 43 2/12/80 MCPR & MAPLHGR Changes for Cycle 8 and Extended Core Burnup 50 44 2/29/80 ILRT Requirements NOTE 1 - 8/29/80 Order for Modification of License-Environmental Qualification NOTE 1 - 9/19/80 Revised Order for Modification of License-Environmental Qualification 51 - 10/24/80 Order for Modification of License-Environmental Qualification Records 52 - 1/9/81 Issuance of Facility Operating License (FTOL)

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

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 Am. 173

TABLE 2 (Page 4 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Maior Subiect Revision DPR-22 (REV) No. Amend No. & Date 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 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 Am. 173

TABLE 2 (Page 5 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Malor Subject Revision DPR-22 (REV) No. Amend No. & Date 91 41 3/24/86 250 Volt Battery 92 42 3/27/86 Jet Pump Surveillance 93 43 4/8/86 Simmer Margin Improvement 94 44 5/27/86 Cycle 12 Operation 95 45 7/1/86 Miscellaneous Changes 96 46 7/1/86 LER Reporting and Miscellaneous Changes 97 47 10/22/86 Single Loop Operation 98 48 12/1/86 Offgas System Trip 99 49 8/26/87 Rod Block Monitor 100 50 8/26/87 APRM and IRM Scram Requirements 101 51 10/16/87 2R Transformer 102 52 11/18/87 Surveillance Intervals - ILRT Schedule 103 53 11/19/87 Extension of Operating License 104 54 11/25/87 Cycle 13 and Misc Changes 105 55 11/25/87 Appendix J Testing 106 56 12/11/87 ATWS - Enriched Boron 107 57 9/23/88 Increased Boron Enrichment 108 58 12/13/88 Physical Security Plan 109 59 2/16/89 Miscellaneous Administrative Changes 110 60 2/28/89 Miscellaneous Administrative Changes 111 61 3/29/89 Fire Protection and Detection System 112 62 3/31/89 ADS Logic and S/RV Discharge Pipe Pressure 113 63 4/18/89 Miscellaneous Technical Specification Improvements 114 64 5/10/89 Containment Vent and Purge Valves 115 65 5/30/89 NUREG-0737 - Generic Letter 83-36 116 66 5/30/89 Reactor Vessel Level Instrumentation 117 67 6/19/89 Extension of MAPLHGR. Exposure for One Fuel Type 118 68 7/14/89 SRO Requirements & Organization Chart Removal Am. 173

TABLE 2 (Page 6 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Maior Subiect Revision DPR-22 (REV) No. Amend No. & Date 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 Removal of Chlorine Detection Requirements and 139 89 8/25/94 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 Am. 173

TABLE 2 (Page 7 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subiect Revision DPR-22 (REV) No. Amend No. & Date 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.

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 1.1/24/99 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 Am. 173

TABLE 2 (Page 8 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Maior Subiect Revision DPR-22 (REV) No. Amend No. & Date 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 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 Am. 173

TABLE 2 (Page 9 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subject Revision DPR-22 (REV) No. Amend No. & Date 127 05/31/02 Missed Surveillance Requirement Technical Specification Changes (TSTF-358) 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 130 09/23/02 Containment Systems Technical Specification Changes 130a 09/26/02 Bases Change - HPCI - Change Wording / HPCI & RCIC -

Enhance with Wording Consistent with NUREG-1433-Rev 1 Update the Multiplier Values for Single Loop Operation 131 10/02/02 Average Planar Linear Heat Generation Rate (APLHGR) 132 02/04/03 Conversion to Option B for Containment Leak Rate Testing 133 02/24/03 Revision to Pressure-Temperature Curves 133a 03/28/03 Bases Change - Adequate Reactor Steam Flow for HPCI/RCIC Testing 134 03/31/03 One-Time Extension of Containment Integrated Leak-Rate Test Interval 135 04/22/03 Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program 135a 04/24/03 Bases Change - Clarify description of head cooling Group 2 valves 136 06/17/03 Elimination of Requirements for Post Accident Sampling System (TSTF-413) 136a 09/25/03 Bases Change - Editorial change to define the abbreviation "EFCV."

137 08/21/03 Drywell Leakage and Sump Monitoring Detection System Am. 173

TABLE 2 (Page 10 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subiect Revision DPR-22 (REV) No. Amend No. & Date 137a 10/09/03 Bases Change - RCS Leakage Requirements for TS 3.6.4.D 137b 10/14/03 Bases Change - Clarification of system control boundary for ASDS 138 05/21/04 Elimination of Requirements for Hydrogen Recombiners and Hydrogen and Oxygen Monitors (TSTF-447) 138a 06/10/04 Bases Change - Clarification of Tech Spec Table 4.1.1 Manual Scram 139 06/02/04 Revised Analysis of Long-Term Containment Response and Net Positive Suction Head (Design Bases and USAR change) 140 11/02/04 Revision to Technical Specification Tables 3.2.1 and 3.2.4 140a 01/13/05 Bases Change - Removal of Drywell Vent Coolers from 3.6/4.6 Bases 141 01/28/05 Revision to Technical Specifications Table 3.2.3 and Section 3.7/4.7 141a 02/24/05 Bases Change - Implement Improved BPWS as Described in NEDO-33091-A 141b 03/10/05 Bases Change - Bases Changes for License Amendments 138 and 140 141c 03/10/05 Bases Change - Removal of 3% Delta-K from Standby Liquid Control Bases 3.4.A/4.4.A 142 02/01/05 Deletion of Requirements for Submittal of Occupational Radiation Reports, Monthly Operating Reports, and Report of Safety/Relief Valve Failures and Challenges (TSTF-369) 143 09/30/05 Implementation of 24-Month Fuel Cycles NOTE 1: 10/20/05 Change to Facility Operating License Pursuant to Commission Order EA-03-086 Regarding Revised Design Basis Treat (DBT); and Revisions to Physical Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan 144 01/12/06 Surveillance Test Intervals for various instruments (Second part of 24-month Fuel Cycle amendment.)

144a 04/05/06 TS Bases changes to conform with implementation of License Amendments 143 and 144 (24-Month Fuel Cycles).

145 04/24/06 Alternate Source Term - Fuel Handling Accident (TSTF-51)

Am. 173

TABLE 2 (Page 11 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subiect Revision DPR-22 (REV) No. Amend No. & Date 146 06/05/06 Improved Technical Specifications (TSTF-359, 372, 439, 455, 458, 460, 464, 479, 480, 485)

Correction Letter 10/12/06 Correction of typo in Amendment 146 (page 3.3.5.1-8) 147 07/05/06 Degraded Voltage Allowable Value Change (Second part of ITS - Follow-on ITS Amendment)

NOTE 1: Renewed OL 11/08/06 Renewed Facility Operating License 148 12/07/06 Alternate Source Term - Full Scope Appendix J Exemptions In conjunction with issuance of Amendment 148, exemptions to 10 CFR 50.54(o) and to 10 CFR 50, Appendix J, Option B, Sections III.A and III.B were issued.

149 01/18/07 One-Time Extension of LPCI Loop Select Logic Time Delay Relay Surveillance Interval Correction Letter 02/23/07 Correction: Remove Am 148 from first page of OL and added Renewed License No. DPR-22 to footer on page 2.

150 03/09/07 Increase SFP allowed Heat Load and allow installation of 64 cell PaR Fuel Storage Rack (if required) to maintain Full Core Offload capability during ISFSI construction.

151 07/20/07 Extend Surveillance Interval from 92-days to 24-months and increase Allowable Values for LPCI Loop Select Logic Time Delay Relays. (Also, correct typo on page 3.3.5.1-6.)

NOTE 1: 08/23/07 Conforming License Amendment to incorporate the Mitigation Strategies Required by Section B.5.b of Commission Order EA-02-026 and the Radiological Protection Mitigation Strategies Required by Commission Order EA-06-137 152 11/08/07 Remove the Table of Contents (TOC) out of the Technical Specifications and place under licensee control. TS TOC initial revision is Revision 0.

153 01/30/08 Revise Surveillance Requirement 3.8.4.2 to specify that the Division 2 battery chargers are verified to supply greater than or equal to 110 amps.

154 01/23/08 Add Action Statement for two inoperable Control Room Ventilation subsystems to Specification 3.7.5 (TSTF-477).

155 02/21/08 Revise Surveillance Requirement 3.5.1.3.b to specify that the Alternate Nitrogen System supply pressure to the ADS valves is verified to be greater than or equal to 410 psig.

NOTE 1: 156 09/22/08 Transfer of Operating License from NMC to NSP - Minnesota 157 10/22/08 Add LCO 3.0.9 on unavailable Barriers (TSTF-427).

158 11/19/08 Revise Control Rod notch testing frequency from once per 7 days to every 31 days (TSTF-475).

159 1/30/09 Power Range Neutron Monitoring System Am. 173

TABLE 2 (Page 12 of 13)

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP Page License Major Subject Revision DPR-22 (REV) No. Amend No. & Date Correction Letter 3/9/09 Correct footer of Specification 3.3.7.2 - AST Amendment 148.

160 3/17/09 Control Room Envelope Habitability (TSTF-448).

161 4/7/09 Revise the Allowable Value and channel calibration frequency for Table 3.3.5.1-1, Function 2.j, Recirculation Riser Differential Pressure - High (Break Detection).

162 7/10/09 Add new Conditions to Specification 3.5.1 for restoration of various low-pressure ECCS subsystems out-of-service (OOS) combinations (e.g., one low-pressure ECCS division OOS).

163 8/19/09 Deleted paragraph d concerning work hour limitations under Specification 5.2.2 to align with rule changes under 10 CFR 26, Subpart I (TSTF-511).

164 9/28/09 Corrected Modes in Table 3.3.6.1-1, Function 5.d, RWCU System isolation on a SLC System initiation, to match SLC System modes (Specification 3.1.7) after adoption of full-scope AST, i.e.,

added Mode 3 to Function 5.d 165 5/4/11 Revise MCPR Safety Limit to 1.15 to reflect reload analyses (which include EPU and MELLLA+ considerations).

166 7/29/11 Add Cyber Security Plan license condition under OL Section 3, Physical Protection.

167 1/11/12 Revise Core Spray flowrate from 2800 gpm to 2835 gpm.

168 7/27/12 Revise surveillance requirements in Specifications 3.4.3, 3.5.1 and 3.6.1.5 to remove requirements to lift-test SRVs during plant startup.

169 8/27/12 Revise licensing basis to reflect removal of the capability to automatically transfer to the 1AR Transformer as a source of power to the essential buses on degraded voltage and instead directly transfer to the EDGs. (Operating License change only.)

170 9/7/12 Revise Table 3.3.5.1-1, Functions 1.e and 2.e, "Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive)",

(i.e., 20 minute ADS bypass timer), to remove the lower limit of the allowable value.

171 1/25/13 Revise Required Actions table for LCO 3.3.1.1 to provide restoration period before declaring the APRMs inoperable when SR 3.3.1.1.2 is not met.

172 1/20/13 Revise Specification 3.4.9, add new Specification 5.6.5, and add a new definition to specify the adoption of a PTLR.

173 7/15/13 Add footnote reflecting that RWM can be bypassed when an improved BPWS is used for reactor shutdown (TSTF-476).

Am. 173

TABLE 2 (Page 13 of 13)

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

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 affected pages not changed.

6. All pages reissued using INTERLEAF in different font. Using NRC Amendment Nos. and issue date. For Bases and Table of Contents, spelling errors corrected and editorial corrections made and all Amendment Nos. changed to 100a. For remaining Tech Spec pages, no other changes made and current Amendment Nos. used.

Am. 173

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 As specified in SR 3 .3.2.1. 4 (h)(i) COLR SR 3.3.2.1.5

b. Intermediate Power Range - (b) 2 SR 3.3.2.1.1 As specified in Upscale SR 3 .3.2.1.41h)li COLR SR 3.3.2.1.5

c. High Power Range - Upscale (c), (d) 2 SR 3.3.2.1.1 As specified in SR 3 .3.2.1.4(h)() COLR SR 3.3.2.1.5

d. Inop (d), (e) 2 SR 3.3.2.1.1 NA

2. Rod Worth Minimizer 1 (, 2(l 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown (g) 2 SR 3.3.2.1.7 NA Position (a) THERMAL POWER > 30% and < 65% RTP and MCPR is below the limit specified in COLR.

(b) THERMAL POWER > 65% and < 85% RTP and MCPR is below the limit specified in COLR.

(c) THERMAL POWER > 85% and < 90% RTP and MCPR is below the limit specified in COLR.

(d) THERMAL POWER > 90% RTP and MCPR is below the limit specified in COLR.

(e) THERMAL POWER > 30% and < 90% RTP and MCPR is below the limit specified in COLR.

(f) With THERMAL POWER _ 10% RTP, except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.

(g) Reactor mode switch in the shutdown position.

(h) If the as-found channel setpoint is not the Nominal Trip Setpoint (NTSP) but is conservative with respect to the Allowable Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(i) The instrument channel setpoint shall be reset to the Nominal Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The NTSP shall be specified in the COLR.

The methodology used to determine the NTSP is specified in the Technical Requirements Manual.

Monticello 3.3.2.1-5 Amendment No. 44.6, ! 59, 173

ML13261A146

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