ML15275A040

Revision 37 to Technical Specification Bases

ML15275A040
Person / Time
Site:	Monticello
Issue date:	10/02/2015
From:	Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML15275A040 (47)

Text

TECHNICAL SPECIFICATION BASES UPDATING INSTRUCTIONS MNGP TECHNICAL SPECIFICATION BASES REMOVE Pagce(s) Revision Document Type Table 1 36 List of Effective Sections I

______________ ________ Specifications Table 2 36 Record of Revisions B 3.3.1.1-1 through 32 Bases B 3.3.1.1-39 _____ Spec. 3.3.1.1 Destroy removed pages.

1061 Revision 37

TABLE 1 (Page 1 ofl1)

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

B 2.1.1 34 B 3.6.1.4 0 B 2.1.2 6 B 3.6.1.5 25 B 3.0 27 B 3.6.1.6 0 B 3.1.1 0 B 3.6.1.7 0 B 3.1.2 0 B 3.6.1.8 29 B 3.1.3 11 B 3.6.2.1 30 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 29 B 3.6.3.1 0 B-3.117 -4 B 316.4:1- 0 B 3.1.8 4 B 3.6.4.2 0 B 3.2.1 29 B 3.6.4.3 31 B 3.2.2 0 B 3.7.1 29 B 3.2.30 B 3.7.2 0 B 3.3.1.1 37 I B 3.7.3 17 B 3.3.1.2 0 I B 3.7.4 31 B 3.3.2.1 26 B 3.7.5 8 B 3.3.2.2 29 B 3.7.6 4 B 3.3.3.1 3 B 3.7.7 29 B 3.3.3.2 3 B 3.7.8 0 B 3.3.4.1 0 B 3.8.1 36 B 3.3.5.1 29 B 3.8.2 27 B 3.3.5.2 0 B 3.8.3 36 B 3.3.6.1 34 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 32 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 28 B 3.4.9 25 B 3.10.2 0 B 3.4.10 0 B 3.10.3 3 B 3.5.1 34 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 29 B 3.10.7 0 B 3.6.1.2 29 B 3.10.8 0 B 3.6.1.3 29 Rev. 37

TABLE 2 (Page I of 6)

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/Cor'rections -

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

Rev. 37

TABLE 2 (Page 2 of 6)

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.

8B.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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Clarify the OPERABILITY requirements of certain CRV fans currently required to support CREF subsystem operation.

9B 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. 37

TABLE 2 (Page 3 of 6)

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

... ..... Differenti~alPressure - 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. 37

TABLE 2 (Page 4 of 6)

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 1ARS 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 1AR 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-476 for improved BPWS.

Rev. 37

TABLE 2 (Page 5 of 6)

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

Specification B 3.0 Clarify application of SR 3.0.2 and SR 3.0.3 to 27 IST tests to clarify compliance to Enforcement Guidance Memorandum (EGM) 12-001 (CAP 01389604).

Correct referenced SR number from SR 3.8.1.8 B 3.8.2 to SR 3.8.1.6.

28 B 3.10.1 Amendment 174 - Revised specification to incorporate TSTF-484, to allow scram time testing in conjunction with hydrostatic testing.

Amendment 177- Implement Extended Power 29 B 3.1.6 B 3.2.1 Uprate.

B 3.3.1.1 B 3.3.2.2 B 3.3.5.1 B 3.4.1 B 3.5.1 B 3.6.1.1 B 3.6.1.2 B 3.6.1.3 B 3.6.1.8 B 3.7.1 B 3.7.7 30 B 3.6.2.1 Revised TS Bases to indicate local Suppression Pool temperature limits were eliminated with Amendment 126.

B 3.8.3 Amendment 178 - Revised specification to relocate stored fuel and lube oil volumes to TS bases and replace with duration requirements in accordance with TSTF-501.

B 3.6.4.3 Amendment 181 - Revised specifications to 31 B 3.7.4 reflect reduced runtime for SBGT and CREF from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to 15 minutes in accordance with TSTF-522.

Rev. 37

TABLE 2 (Page 6 of 6)

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

Specification 32 B 3.3.1.1, Amendment 180- Implement Maximum B 3.4.1 Extended Load Line Limit Analysis, Plus (MELLLA+)

33 B 3.8.1, Add EDG Fuel Oil Transfer System train_

... . B 3.8.3 ... .des-crip-tior for each-EDG.

34 B 2.1.1, Amendment 185 - Reduce the Reactor Steam B 3.3.6.1 Dome Pressure specified in the Reactor Core Safety Limits, resolved GE Part 21 condition for a potential to violate the safety limit during a Pressure Regulator Failure Downscale event.

B 3.5.1 Amendment 184 - Removes former Condition F which allowed both Core Spray subsystems to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

35 B 3.8.1, Revised to reflect EDG Fuel Oil Transfer B 3.8.3 System separation modifications.

36 B 3.8.1, Revised to reflect completion of EDG Fuel Oil B 3.8.3 Transfer System modifications and differentiate between 6-day and 7-day EDG fuel oil storage tank level requirements for each division.

37 B 3.3.1.1 Revised to clarify Automated BSP Scram Region implementation to protect MCPR safety limit during thermal-hydraulic power oscillations when the OPRM Upscale function is inoperable.

Rev. 37

RPS Instrumentation B 3.3.1.1 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) and minimize the

-- - energy that must be absorbed following~a loss of coolant accident--

(LOCA). This can be accomplished either automatically or manually.

The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. Technical Specifications are required by 10 CFR 50.36 to contain LSSS defined by the regulation as

".settings for automatic protective devices.. .so chosen that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded." The Analytic Limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytic Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the Analytic Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The trip setpoint is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process variable reaching the Analytic Limit and thus ensuring that the SL would not be exceeded. As such, the trip setpoint accounts for uncertainties in setting the device (e.g., calibration), uncertainties in how the device might actually perform (e.g., repeatability), changes in the point of action of the device over time (e.g., drift during surveillance intervals), and any other factors which may influence its actual performance (e.g., harsh accident environments). In this manner, the trip setpoint plays an important role in ensuring that SLs are not exceeded. As such, the trip setpoint meets the definition of an LSSS (Ref. 1) and could be used to meet the requirement that they be contained in the Technical Specifications.

Technical Specifications contain values related to the OPERABILITY of equipment required for safe operation of the facility. OPERABLE is defined in Technical Specifications as ". ..being capable of performing its safety function(s)." For automatic protective devices, the required safetY function is to ensure that a SL is not exceeded and therefore the LSSS as Monticello B 3.3.1.1-1 B.3.11-1Revision Montcell No. 37

RPS Instrumentation B 3.3.1.1 BASES BACKGROUND (continued) defined by 10 CFR 50.36 is the same as the OPERABILITY limit for these devices. However, use of the trip setpoint to define OPERABILITY in Technical Specifications and its corresponding designation as the LSSS required by 10 CFR 50.36 would be an overly restrictive requirement if it were applied as an OPERABILITY limit for the "as found" value of a protective device setting during a Surveillance. This would result in Tec~hn~ical Sp~ecificati~on _comp~lian~ce problems, as well as reports and - --

corrective actions required by the rule which are not necessary to ensure safety. For example, an automatic protective device with a selling that has been found to be different from the trip setpoint due to some drift of the setting may still be OPERABLE since drift is to be expected. This expected drift would have been specifically accounted for in the setpoint methodology for calculating the trip setpoint and thus the automatic protective action would still have ensured that the SL would not be exceeded with the "as found" setting of the protective device. Therefore, the device would still be OPERABLE since it would have performed its safety function and the only corrective action required would be to reset the device to the trip setpoint to account for further drift during the next surveillance interval.

Use of the trip setpoint to define "as found" OPERABILITY and its designation as the LSSS under the expected circumstances described above would result in actions required by both the rule and Technical Specifications that are clearly not warranted. However, there is also some point beyond which the device would have not been able to perform its function due, for example, to greater than expected drift. This value needs to be specified in the Technical Specifications in order to define OPERABILITY of the devices and is designated as the Allowable Value which, as stated above, is the same as the LSSS.

The Allowable Valuable specified in Table 3.3.1-1 serves as the LSSS such that a channel is OPERABLE ifthe trip setpoint is found not to exceed the Allowable Value. As such, the Allowable Value differs from the trip setpoint by an amount primarily equal to the expected instrument loop uncertainties, such as drift, during the surveillance interval. In this manner, the actual setting of the device will still meet the LSSS definition and ensure that a SL is not exceeded at any given point of time as long as the device has not drifted beyond that expected during the surveillance interval. If the actual setting of the device is found to have exceeded the Allowable Value the device would be considered inoperable from a Technical Specification perspective. This requires corrective action including those actions required by 10 CFR 50.36 when automatic protective devices do not function as required. Note that, although the Monticello B 3.3.1.1-2 Montcell No. 37 B.3.11-2Revision

RPS Instrumentation B 3.3.1.1 BASES BACKGROUND (continued) channel is "OPERABLE" under these circumstances, the trip setpoint should be left adjusted to a value within the established trip setpoint calibration tolerance band, in accordance with uncertainty assumptions stated in the referenced setpoint methodology (as-left criteria), and confirmed to be operating within the statistical allowances of the uncertainty terms assigned.

The RPS, as described in USAR, Section 7.6.1.2.1 (Ref. 2), includes sensors, relays, bypass circuits, and switches that are necessary to cause initiation of a reactor scram. Functional diversity is provided by monitoring a wide range of dependent and independent parameters. The input parameters to the scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam line isolation valve position, turbine control valve (TCV) acceleration relay oil pressure, turbine stop valve (TSV) position, drywell pressure, and scram discharge volume (SDV) water level, as well as reactor mode switch in shutdown position and manual scram signals.

There are at least four redundant sensor input signals from each of these parameters (with the exception of the reactor mode switch in shutdown and manual scram signals). Some channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an RPS trip signal to the trip logic.

The RPS is comprised of two independent trip systems (A and B) with three logic channels in each trip system (logic channels Al, A2, and A3, B1, B2, and B3) as described in Reference 3. The automatic trip logics of trip system A are logic channels Al and A2; the manual trip logic of trip system A is logic channel A3. Similarly, the trip logics for trip system B are logic channels B1, B2, and B3. The outputs of the automatic logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram. This logic arrangement is referred to as a one-out-of-two taken twice logic. The outputs of the manual logic channels in a trip system are combined in a one-out-of-one logic. The tripping of both manual logic channels will produce a scram.

Each trip system can be reset by use of a reset switch. If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for a short time delay after the full scram signal is received. The short time delay on reset ensures that the scram function will be completed.

Monticello B 3.3.1.1-3 Montcell B.3.11-3Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES BACKGROUND (continued)

Two scram pilot valves are located in the hydraulic control unit for each control rod drive (CRD). Each scram pilot valve is solenoid operated, with the solenoids normally energized. The scram pilot valves control the air supply to the scram inlet and outlet valves for the associated CRD. When either scram pilot valve solenoid is energized, air pressure holds the scram valves closed and, therefore, both scram pilot valve solenoids must be de-energized to cause a control rod to scram. The scram valves control the supply and discharge paths for the CRD water during a scram.

One of the scram pilot valve solenoids for each CR0 is controlled by trip system A, and the other solenoid is controlled by trip system B. Any trip of trip system A in conjunction with any trip in trip system B results in de-energizing both solenoids, air bleeding off, scram valves opening, and control rod scram.

The backup scram valves, which energize on a scram signal to depressurize the scram air header, are also controlled by the RPS.

Additionally, the RPS System controls the SDV vent and drain valves such that when both trip systems trip, the SOV vent and drain valves close to isolate the SDV.

APPLICABLE The actions of the. RPS are assumed in the safety analyses of SAFETY References 4 and 5. The RPS initiates a reactor scram when ANALYSES, LCO, monitored parameter values exceed the Allowable Values, specified by and APPLICABILITY the setpoint methodology and listed in Table 3.3.1.1-1Ito preserve the integrity of the fuel cladding, the reactor coolant pressure boundary (RCPB), and the containment by minimizing the energy that must be absorbed following a LOCA.

RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Functions not specifically credited in the accident analysis are retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The OPERABILITY of the RPS is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1.

Each Function must have a required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Each channel must also respond within its assumed response time, where applicable.

Revision No. 37 B 3.3.1.1-4 Monticello Monticello B 3.3.1.1-4 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCD, and APPLICABILITY (continued)

Allowable Values are specified for most of the RPS Functions specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.

Trip setpoints 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 vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values and nominal trip setpoints (NTSP) 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 accuracy. The margin between the Allowable Value and the NTSP allows for instrument drift that might occur during the established surveillance period. Two separate verifications are performed for the calculated NTSP. The first, a Spurious Trip Avoidance Test, evaluates the impact of the NTSP on plant availability. The second verification, an LER Avoidance Test, calculates the probability of avoiding a Licensee Event Report (or exceeding the Allowable Value) due to instrument drift. These two verifications are statistical evaluations to provide additional assurance of the acceptability of the NTSP and may require changes to the NTSP. Use of these methods and verifications 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.

The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCD.

The individual Functions are required to be OPERABLE in the MODES or other specified conditions indicated in Table 3.3.1.1-1, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals.

Monticello B 3.3.1.1-5 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The RPS is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. During normal operation in MODES 3 and 4, all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1, "Control Rod Block Instrumentation") does not allow any control rod to be withdrawn. In MODE 5, control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram. Provided all other control rods remain inserted, the RPS function is not required. In this condition, the required SDM (LCO 3.1.1, "SHUTDOWN MARGIN") and refuel position one-rod-out interlock (LCO 3.9.2, "Refueling Position One-Rod-Out Interlock") ensure that no event requiring RPS will occur. Under these conditions, the RPS function is not required to be OPERABLE.

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

Intermediate Rangqe Monitor (IRM) l.a. Intermediate Rangqe Monitor Neutron Flux - Higqh Higqh The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the average power range monitors (APRMs). The IRMs are capable of generating trip signals that can be used to prevent fuel damage resulting from abnormal operating transients in the intermediate power range. In this power range, the most significant source of reactivity change is due to control rod withdrawal.

The IRM provides diverse protection for the rod worth minimizer (RWM),

which monitors and controls the movement of control rods at low power.

The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an unacceptable neutron flux excursion (Ref. 6). The IRM provides mitigation of the neutron flux excursion. To demonstrate the capability of the IRM System to mitigate control rod withdrawal events, generic analyses have been performed (Ref. 7) to evaluate the consequences of control rod withdrawal events during startup that are mitigated only by the IRM. This analysis, which assumes that one IRM channel in each trip system is bypassed, demonstrates that the IRMs provide protection against local control rod withdrawal errors and results in peak fuel energy depositions below the 170 cal/gm fuel failure threshold criterion. The IRMs are capable of limiting other reactivity excursions during startup, such as cold water injection events, although no credit is specifically assumed. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

Monticello B 3.3.1.1-6 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The IRM System is divided into two groups of IRM channels, with four IRM channels inputting to each trip system. The analysis of Reference 7 assumes that one channel in each trip system is bypassed. Therefore, six channels with three channels in each trip system are required for IRM OPERABILITY to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. For an IRM to be considered OPERABLE it must be fully inserted. This trip is active in each of the 9 ranges of the IRM (center IRMs 13, 16 and 17 highest range is Range 10), which must be selected by the operator to maintain the neutron flux within the monitored level of an IRM range.

The calculation in Reference 26 has adequate conservatism to permit an IRM Allowable Value of 121.5 divisions of a 125 division scale. The Intermediate Range Monitor Neutron Flux - High High SCRAM at the top of scale (125 divisions) provides effective protection against reactivity insertion transients in the startup mode (MODE 2). During startup, the operator changes IRM ranges to keep the IRM onscale and, by procedure, assures the IRM is > 10 and -*75divisions of scale on any range for reliable monitoring. Thus, any local power transient occurring during startup (on any range but the highest) cannot cause power to increase beyond a factor of 10 before reaching the IRM Neutron Flux -

High High SCRAM setpoint. In the startup range, this provides effective safety protection. There is enough overlap with the APRM on the highest IRM range so the Average Power Range Monitor Neutron Flux - High (Setdown) SCRAM, which is OPERABLE in MODE 2, will provide protection if required. The APRM Neutron Flux - High (Setdown) setpoint assures the SCRAM will occur before power exceeds 25% RTP in MODE 2 and, per procedure, that the plant changes from MODE 2 to RUN (MODE 1) before power reaches 25% RTP. This also assures that MODE 1 thermal limits monitoring is performed as required for power

> 25% RTP.

The Intermediate Range Monitor Neutron Flux - High High Function must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists. In MODE 5, when a cell with fuel has its control rod withdrawn, the IRMs provide monitoring for and protection against unexpected reactivity excursions. In MODE 1, the APRM System and the RWM provide protection against control rod withdrawal error events and the IRMs are not required.

1.b. intermediate Range Monitor- Inop This trip signal provides assurance that a minimum number of IRMs are OPERABLE. Anytime an IRM mode switch is moved to any position other than "Operate," the detector voltage drops below a preset level, or when a module is not plugged in, an inoperative trip signal will be Monticello B 3.3.1.1-7 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) received by the RPS unless the IRM is bypassed. Since only one IRM in each trip system may be bypassed, only one IRM in each RPS trip system may be inoperable without resulting in an RPS trip signal.

This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

Six channels of Intermediate Range Monitor - Inop with three channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.

This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux - High High Function is required.

Average Power Rangqe Monitor The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.

The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. Each APRM channel also includes an Oscillation Power Range Monitor (OPRM) Upscale Function which monitors small groups of LPRM signals to detect thermal-hydraulic instabilities.

The APRM System is divided into four APRM channels and four 2-out-of-4 voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each; with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one un-bypassed APRM will result in a "half-trip" in all four of the voter channels, but no trip inputs to either RPS trip system. Because APRM trip Functions 2.a, 2.b, 2.c and 2.f are implemented in the same hardware, these trip Functions are combined with APRM Inop trip Function 2.d. Any Function 2.a, 2.b, 2.c or 2.d trip from any two un-bypassed APRM channels will result in a full trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip system logic channel (A, A2, BI and B2). Similarly, any Function 2.d or 2.f trip from any two un-bypassed APRM channels Monticello B 3.3.1.1-8 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) will result in a full trip from each of the four voter channels.

Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b and 2.c, at least 14 LPRM inputs, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel.

For the OPRM Upscale Function (Function 2.f), LPRMs are assigned to "cells". A minimum of 8 responsive cells, each with a minimum of 2 LPRMs, must be OPERABLE for the OPRM Upscale Function to be OPERABLE (Ref. 25).

2.a. Average Power Ranqe Monitor Neutron Flux - High (Setdown')

For operation at low power (i.e., Mode 2), the Average Power Range Monitor Neutron Flux - High (Setdown) Function is capable of generating a trip signal to prevent fuel damage resulting from abnormal operating transients in this power range. During most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Setdown)

Function will provide a secondary scram to the Intermediate Range Monitor (IRM) Neutron Flux - High Function because of the relative setpoints. When the IRMs are on Range 9 or 10, it is possible that the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide the primary trip signal for a core wide increase in power.

No specific safety analyses take credit for the Average Power Range Monitor Neutron Flux - High (Setdown) Function. However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% (SL 2.1.1.1) when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER *<25% RTP.

The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP.

The Average Power Range Monitor Neutron Flux - High (Setdown)

Function must be OPERABLE during MODE 2 when control rods may be withdrawn because the potential for criticality exists.

In MODE 1, the Average Power Range Monitor Neutron Flux - High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events.

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In accordance with the NRC Safety Evaluation for Amendment 159 (Ref. 24), the Average Power Range Monitor Neutron Flux - High (Setdown) Function is not LSSS SL-reiated.

2.b. Averagqe Power Rangqe Monitor Simulated Thermal Power - Hiqh The Average Power Range Monitor Simulated Thermal Power - High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux is electronically filtered with a time constant, representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod pattern) but is clamped at an upper limit that is always lower than the Average Power Range Monitor Neutron Flux - High Function Allowable Value.

A note is included, applicable when the plant is in single recirculation loop operation per LCO 3.4.1, which requires reducing by Delta W the flow value used in the Allowable Value equation. The value of Delta W is defined in the COLR. The value of Delta W is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. This adjusted Allowable Value thus maintains thermal margins essentially unchanged from those for two-loop operation.

No specific safety analyses take credit for the Average Power Range Monitor Simulated Thermal Power - High Function, however it provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and therefore mitigates over-power and delta CPR for such events. During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram. For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Neutron Flux - High Function will provide a scram signal before the Average Power Range Monitor Simulated Thermal Power - High Function setpoint is exceeded.

Each APRM channel uses one total drive flow signal representative of total core flow. The total drive flow signal is generated by the flow processing logic, part of the APRM channel, by summing up the flow calculated from two flow transmitter signal inputs, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this Function.

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Although no specific safety analyses take credit for the clamped Allowable Value at Monticello, it is based on analyses that do take credit for the Average Power Range Monitor Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.

The Average Power Range Monitor Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER.

In accordance with the NRC Safety Evaluation for Amendment 159 (Ref. 24), the Average Power Range Monitor Simulated Thermal Power -

High Function is not LSSS SL-related.

Automated Backup Stability Protection (BSP)

When Oscillation Power Range Monitor (OPRM) Upscale (Function 2.f) trip capability is not maintained, alternate backup stability protection is required in accordance with Action 1.2. An Automated BSP Scram Region (see Section 7.4 of Reference 18) is implemented to prevent entry into the region of the power and flow-operating map susceptible to reactor instability. Modified Average Power Range Monitor Simulated Thermal Power - High scram setpoints initiate a reactor trip for flow reduction events that would terminate in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram to provide protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.

2.c. Averagqe Power Rangqe Monitor Neutron Flux - Higqh The Average Power Range Monitor Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 9, high neutron flux is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety/relief valves (SRVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 10) takes credit for high neutron flux to terminate the CRDA.

The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

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The Average Power Range Monitor Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux - High Function is functional in MODE 2, the Average Power Range Monitor Neutron Flux - High (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux - High Function is not required in MODE 2.

In accordance with the guidance of Regulatory Issue Summary 2006-17 (Ref. 23) and the NRC Safety Evaluation for Amendment 159 (Ref. 24),

the Average Power Range Monitor Neutron Flux - High Function is LSSS SL-related.

2.d. Averaqe Power Rangqe Monitor - Inop, Three of the four APRM channels are required to be OPERABLE for each of the APRM Functions. This Function (Inop) provides assurance that the minimum numbers of APRM channels are OPERABLE.

For any APRM channel, any time its mode switch is in any position other than "Operate," an APRM module is unplugged, or the automatic self-test system detects a critical fault with the APRM channel, an Inop trip is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from all four voter channels to their associated trip system.

This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

There is no Allowable Value for this Function.

This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

2.e. 2-Out-Of-4 Voter The 2-Out-Of-4 Voter Function provides the interface between the APRM Functions, including the OPRM Upscale Function, and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the 2-Out-Of-4 Voter Function needs to be OPERABLE in MODES 1 and 2.

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All four voter channels are required to be OPERABLE. Each voter channel includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, a trip is issued from that voter channel to the associated trip system.

The 2-Out-Of-4 Voter Function votes APRM Functions 2.a, 2.b and 2.c independently of Function 2.f. This voting is accomplished by the 2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module. The voter also includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs).

The voter Function 2.e must be declared inoperable if any of its functionality is inoperable. However, due to the independent voting of APRM trips, and the redundancy of outputs, there may be conditions where the voter Function 2.e is inoperable, but trip capability for one or more of the other APRM Functions through that voter is still maintained.

This may be considered when determining the condition of other APRM Functions resulting from partial inoperability of the 2-Out-Of-4 Voter Function 2.e.

There is no Allowable Value for this Function.

2.f. Oscillation Power Rangqe Monitor (OPRM Upscale)

The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations; Reference 18 describes the Detect and Suppress - Confirmation Density (DSS-CO) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). Use of DSS-CD was approved by Amendment 180 (Ref. 31). There are three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA),

and the growth rate algorithm (GRA) are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA.

The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the CDA.

The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.

DSS-CD operability requires at least 8 responsive OPRM cells per channel.

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RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The OPRM Upscale Function is required to be OPERABLE when the plant is Ž 20% RTP, which is established as a power level that is greater than or equal to 5% below the lower boundary of the Armed Region. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is > 25% RTP corresponding to the plant specific MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note e allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing though the DSS-CD Armed Region during both a single startup and a single shutdown following DSS-CD implementation. Note e reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs.

However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary.

An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by period confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed its trip condition for one or more cells in that channel.

Three of the four channels are required to be operable. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded. There is no allowable value for this function.

The OPRM Upscale settings are not traditional instrumentation setpoints determined under an instrument setpoint methodology. In accordance with the NRC Safety Evaluation for Amendment 159 (Ref. 24), the OPRM Upscale Function is not LSSS SL-related and Reference 26 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument's setpoint methodology.

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3. Reactor Vessel Steam Dome Pressure - Hi~qh An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure -

High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressbrization protection analysis of Reference 9, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure switches that sense reactor pressure. The Reactor Vessel Steam Dome Pressure

- High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level - Low Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at this level to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low Function is not assumed in the analysis of the recirculation line break (Ref. 11) since the scram occurs in the beginning of the event due to the loss of offsite power. However, analyses have been performed that indicate that the difference between a scram initiated at the beginning of the event and a scram initiated by Reactor Vessel Water Level - Low is negligible. The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

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Reactor Vessel Water Level - Low signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Water Level - Low Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

The Reactor Vessel Water Level - Low Allowable Value is selected to ensure that for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water Level - Low Low will not be required.

The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. EGOS initiations at Reactor Vessel Water Level - Low Low provide sufficient protection for level transients in all other MODES.

5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.

However, for the overpressurization protection analysis of Reference 9, the Average Power Range Monitor Neutron Flux - High Function, along with the S/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.

Additionally, MSIV closure is assumed in the main steam line break accident analyzed in Reference 12. The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged Monticello B 3.3.1.1-16 Montcell B33.1.-16Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.

The Main Steam Isolation Valve - Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.

Sixteen channels of the Main Steam Isolation Valve - Closure Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 and MODE 2 with reactor pressure Ž_600 psig since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2 with reactor pressure < 600 psig, the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection. This Function is automatically bypassed when the reactor mode switch is in a position other than run and the reactor pressure is < 600 psig.

6. Drvwell Pressure - Hicqh High pressure in the drywell could indicate a break in the RCPB. A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell.

The Drywell Pressure - High Function is a secondary scram signal to Reactor Vessel Water Level - Low for LOCA events inside the drywell.

However, no credit is taken for a scram initiated from this Function for any of the DBAs analyzed in the USAR. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

High drywell pressure signals are initiated from four pressure switches that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.

Four channels of Drywell Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS, resulting in the limiting transients and accidents.

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RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 7.a, 7.b. Scram Dischar~qe Volume Water Level - Hi~qh The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered. Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level

- High Functions are an input to the RPS logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the USAR. However, they are retained to ensure the RPS remains OPERABLE.

SDy water level is measured by two diverse methods. The level in each of the two SDVs is measured by two float type level switches and two thermal probes for a total of eight level signals. The outputs of these devices are arranged so that there is a signal from a level switch and a thermal probe to each RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 13.

The Allowable Value is chosen low enough to ensure that there is sufficient volume in the SDV to accommodate the water from a full scram.

The Allowable Value refers to the volume of water in the discharge volume receiver tank and does not include the volume in the lines to the level switches.

Four channels of each type of Scram Discharge Volume Water Level -

High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES I and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.

8. Turbine Stop Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of the transients that would result from the closure of these valves. The Turbine Stop Valve - Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 14. For this event, Monticello B 3.3.1.1-18 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LOCO, and APPLICABILITY (continued) the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.

Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs. One position switch and two independent contacts are associated with each stop valve. One of the two contacts provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be enabled at THERMAL POWER > 40% RTP (Refs. 29 and 30). This is normally accomplished automatically by pressure switches sensing turbine first stage pressure. The pressure switches are normally adjusted lower (26.6% RTP - Refs. 29 and 30) to account for the turbine bypass valves being opened, such that approximately 11.5%

of rated steam flow (Refs. 27 and 30) is being passed directly to the condenser.

The Turbine Stop Valve - Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.

Eight channels of Turbine Stop Valve - Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function even if one TSV should fail to close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is > 40% RTP.

This Function is not required when THERMAL POWER is _< 40% RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

9. Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure - Low Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure - Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 15. For this event, the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.

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Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure -

Low signals are initiated by loss of oil pressure at the acceleration relay.

Two pressure switches are mounted on one pressure tap while two other pressure switches are mounted at a distance on another pressure tap.

The pressure switches associated with one pressure tap are assigned to different RPS trip systems. This Function must be enabled at THERMAL POWER > 40% RTP (Refs. 29 and 30). This is normally accomplished automatically by pressure switches sensing turbine first stage pressure.

The pressure switches are normally adjusted lower (26.6% RTP -

Refs. 29 and 30) to account for the turbine bypass valves being opened, such that approximately 11.5% of rated steam flow (Refs. 27 and 30) is being passed directly to the condenser.

The Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure

- Low Allowable Value is selected high enough to detect imminent TCV fast closure.

Four channels of Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is > 40% RTP. This Function is not required when THERMAL POWER is _< 40% RTP, since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the two manual scram logic channels (A3 and B3), which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The reactor mode switch is a single switch with two channels, each of which provides input into one of the two manual scram logic channels.

There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position.

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Two channels of Reactor Mode Switch - Shutdown Position Function, with one channel in each trip system, are available and required to be OPERABLE. The Reactor Mode Switch - Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

11. Manual Scram The Manual Scram push button channels provide signals, via the two manual scram logic channels (A3 and B3), which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

There is one Manual Scram push button channel for each of the two manual scram logic channels. In order to cause a scram it is necessary that both channels be actuated.

There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.

Two channels of Manual Scram with one channel in each trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

ACTIONS Note 1 has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, Note 1 has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.

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RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

Note 2 has been provided to modify the ACTIONS for the RPS instrumentation functions of APRM Flow Biased Simulated Thermal Power - Upscale (Function 2.b) and APRM Fixed Neutron Flux - High (Function 2.c) such that an APRM that is not within the limit of SR 3.3.1.1.2 has a specified restoration period before declaring the associated APRM inoperable (Ref. 28). Therefore, Note 2 allows delaying the entry into associated Conditions and Required Actions to be delayed up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ifthe APRM is indicating a lower power value than the calculated power (non-conservative), and for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the APRM is indicating a higher power value than the calculated power (conservative).

Prior to expiration of the time allotted by the note, the absolute difference between the channel and calculated power is required to be restored to within the limit of SR 3.3.1.1.2 (< 2% RTP) or the applicable Condition entered and Required Actions taken. This note is based on the time required to perform APRM adjustments on multiple channels and the impact on safety; additional time is allowed when the APRM is indicating a higher power value than the calculated power, i.e., out of limits but conservative.

A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (Ref. 16) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2 and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition 0 must be entered and its Required Action taken. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance is not allowed for Reactor Mode Switch - Shutdown Position Function and Manual Scram Function channels since with one channel inoperable RPS trip capability is not maintained. In this case, Condition C must be entered and its Required Actions taken.

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RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

As noted, Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d or 2.f. Inoperability of one required APRM channel affects both trip systems. For that condition, Required Action A.1 must be satisfied, and is the only action (other than restoring operability) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel.

B.1 and B.2 Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system.

Required Actions B.1i and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in either trip system (e.g.,

one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in Reference 16 for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time. Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system.

Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in Reference 16, which justified a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowable out of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). If this action would result in a scram, it is permissible to place the other trip system or its inoperable channels in trip.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.

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RPS Instrumentation B 3.3.1:1 BASES ACTIONS (continued)

Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing tihe inoperable channel or associated trip system in trip would result in a scram, Condition D must be entered and its Required Action taken. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance is not allowed for Reactor Mode Switch - Shutdown Position Function and Manual Scram Function channels since with two channels inoperable RPS trip capability is not maintained. In this case, Condition C must be entered and its Required Action taken.

As noted, Condition B is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d or 2.f. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-out-of-4 voter and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of a Function in more than one required APRM channel results in loss of trip capability for that Function and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a, 2.b, 2.c, 2.d or 2.f, and these functions are not associated with specific trip systems as are the APRM 2-out-of-4 voter and other non-APRM channels, Condition B does not apply.

C._.1 Required Action C. 1 is intended to ensure that appropriate actions are taken if one or more inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of-two taken twice logic and the IRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip). For Function 5 (Main Steam Isolation Valve - Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip). For Function 8 (Turbine Stop Valve - Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip). For Function 10 (Reactor Mode Switch -Shutdown Position) and Function 11 (Manual Scram), since each trip system only has one Monticello B 3.3.1.1-24 Montcell B33.1.-24Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued) channel for each Function, with a channel inoperable, RPS trip capability is not maintained.

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

D...1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed.

Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.

E,1, F.1, G.1 and J.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Actions E.1 and J.1 are consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."

H._1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. 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.

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RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued) 1.1_

If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 18. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 18 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place. Use of BSP was approved by Amendment 180 (Ref. 31).

1.2 and 1.3 Actions 1.2 and 1.3 are both required to be taken in conjunction with Action 1.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 18, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow-operating map that is susceptible to reactor instability. The reactor trip would be initiated by the modified APRM STP scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection.

The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events.

Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The reporting requirements of Specification 5.6.6 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of immediately directs that action be taken in accordance with Specification 5.6.6 to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status.

Revision No. 37 B 3.3.1.1-26 Monticello B3.3.1.1-26 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

J.1 If the Required Actions I are not completed within the associated Completion Times, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time are addressed in the Bases for 1.1. The Manual BSP Regions are required in conjunction with the BSP Boundary.

J._22 The BSP Boundary, as described in Section 7.3 of Reference 18, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that the immediate final statepoint for a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 18).

The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant system.

J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified.

Consequently, the required channels are required to be restored to OPERABLE status within 120 days.

Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (see Action J.1) and the BSP Boundary (see J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B.

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RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

A Note is provided to indicate that LCO 3.0.4 is not applicable. The intent of the note is to allow plant startup while operating within the 120-day Completion Time for Required Action J.3. The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation.

K.._.I If the required channels are not restored to OPERABLE status and the Required Actions of J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the THERMAL POWER must be reduced to less than 20% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.1.1-1.

The Surveillances are modified by a Note to indicate that when a 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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. 16) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does not significantly reduce the probability that the RPS will trip when necessary.

SR 3.3.1.1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument Monticello B 3.3.1.1-28 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated .from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.6.

A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at > 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At > 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

SR 3.3.1.1.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the 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 Monticello B 3.3.1.1-29 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

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 Specification tests at least once per refueling interval with applicable extensions.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 ifthe 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 16).

SR 3.3.1.1.4 A functional test of each automatic scram contactor is performed to ensure that each automatic RPS logic channel will perform the intended function. There are four RPS channel test switches, one associated with each of the four automatic trip channels (Al, A2, B1 and B2). These test switches allow the operator to test the OPERABILITY of the individual trip logic channel automatic scram contactors as an alternative to using an automatic scram function trip. This is accomplished by placing the RPS channel test switch in the test position, which will input a trip signal into the associated RPS logic channel. The RPS channel test switches are not credited in the accident analysis, they just provide a method to test the automatic scram contactors. The Manual Scram Functions are not configured the same as the generic model used in Reference 16.

However, Reference 16 concluded that the Surveillance Frequency extensions for RPS Functions were not affected by the difference in configuration since each automatic RPS logic channel has a test switch that is functionally the same as the manual scram switches in the generic model. As such, a functional test of each RPS automatic scram contactor using either its associated test switch or by test of any of the associated automatic RPS Functions is required to be performed once every 7 days.

The Frequency of 7 days is based on the reliability analysis of Reference 16.

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RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the 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 Specification and non-Technical Specification tests at least once per refueling interval with applicable extensions. The 31 day Frequency is based on engineering judgment, operating experience, and reliability of this instrumentation.

SR 3.3.1.1.6 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD/T" Frequency (Ref. 30) is based on operating experience with LPRM sensitivity changes.

SR 3.3.1.1.7 and SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the 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 Specification and non-Technical Specification tests at least once per refueling interval with applicable extensions. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The CHANNEL FUNCTIONAL TEST (SR 3.3.1.1.10) for the Reactor Mode Switch - Shutdown Position channels will be performed by placing the reactor mode switch in the shutdown position.

The 92 day Frequency of SR 3.3.1.1.7 is based on the reliability analysis of Reference 16. The 24 month Frequency of SR 3.3.1.1.10 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 that these components usually pass the Surveillance when performed at the 24 month Frequency.

Monticello B3.3.1.1-31 B33.1.-31Revision Montcell No. 37

IRPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.1.8 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.

The Frequency of 92 days is based on the reliability analysis of Reference 16.

SR 3.3.1.1.9 and SR 3.3.1.1.11 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that 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. For the APRM Simulated Thermal Power -

High Function, this SR also includes calibrating the associated recirculation loop flow channel.

Note 1 to SIR 3.3.1.1.11 state that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in APIRM neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 MWDIT LPRM calibration (Ref. 30) against the TIPs (SR 3.3.1.1.6).

Note 2 to SIR 3.3.1.1.11 requires the IIRM SRs to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from MODE 1. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SIR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

Note 3 is added to SIR 3.3.1.1.11 to ciarify that the recirculation flow transmitters that feed the APIRMs are included in the CHANNEL CALIBRATION.

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RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

The Frequency of SR 3.3.1.1.9 is based upon the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.1.1.11 is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

The Frequency of SR 3.3.1.1.11 for the APRM / OPRM functions is based upon a 24 month calibration interval (Refs. 17 and 21).

SR 3.3.1.1.11 for Function 2.c, APRM Neutron Flux- High, is modified by two Notes. This function was determined by the NRC Safety Evaluation for Amendment 159 (Ref. 24) to be a LSSS for the protection of the reactor core SLs.

Note (t) requires evaluation of channel performance for the condition where the as-found selling for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. Evaluation of instrument performance will verify that the instrument will continue to perform 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 (g) 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 NTSP and the methodology used to determine the NTSP for the APRM Neutron Flux- High Function, (Function 2.c) in Table 3.3.1.1-1 are specified in Appendix C to the Technical Requirements Manual, a document controlled under 10 CFR 50.59.

SR 3.3.1.1.12 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LCO 3.1.3, "Control Rod OPERABILITY"), and SDV vent and drain valves (LCO 3.1.8, "Scram Discharge Volume Vent and Drain Valves"), overlaps this Surveillance to provide complete testing of the assumed safety function.

The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-out-of-4 voter channel inputs to check all combinations of two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM related redundant RPS relays.

Monticello B3.3.1.1-33 Montcell B33.1.-33Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

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 that these components usually pass the Surveillance when performed at the 24 month Frequency. The Frequency of SR 3.3.1.1.12 for the APRM 2-Out-Of-4 Voter Function is based upon a 24 month calibration interval (Refs. 17 and 21).

SR 3.3.1.1.13 This SR ensures that scrams initiated from the Turbine Stop Valve -

Closure and Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is > 40% RTP (Refs. 29 and 30). This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.

Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during in-service calibration at THERMAL POWER > 40% RTP (Refs. 29 and 30), if performing the calibration using actual turbine first stage pressure, to ensure that the calibration is valid. The pressure switches are normally adjusted lower (26.6% RTP - Refs. 29 and 30) to account for the turbine.

bypass valves being opened, such that approximately 11 .5% of rated steam flow (Refs. 27 and 30) is being passed directly to the condenser.

If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at > 40% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure

- Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

Monticello B3.3.1.1-34 Montcell B33.1.-34Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.1.14 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis.

RPS RESPONSE TIME may be verified by actual response time measurements in any series of sequential, overlapping, or total channel measurements.

The RPS RESPONSE TIME acceptance criterion is 50 milliseconds.

RPS RESPONSE TIME for the APRM 2-Out-Of-4 Voter Function (Function 2.e), includes the output relays of the voter and the associated RPS relays and contactors. (The digital portion of the APRM and 2-out-of-4 voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration checks the time base of the digital electronics. Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time.)

RPS RESPONSE TIME tests are conducted on a 24 month STAGGERED TEST BASIS. Note 2 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels per trip system, in lieu of the 8 channels specified in Table 3.3.1.1-1 for the MSIV- Closure Function.

APRM and OPRM RESPONSE TIME tests are conducted on a 24 month STAGGERED TEST BASIS. Note 1 requires the STAGGERED TEST BASIS to be determined based on 4 channels of APRM outputs and 4 channels of OPRM outputs, (total n = 8) being tested on an alternating basis.

This allows the STAGGERED TEST BASIS Frequency for Function 2.e to be determined based on 8 channels rather than the 4 actual 2-Out-Of-4 Voter channels. The redundant outputs from the 2-Out-Of-4 Voter channel (2 for APRM trips and 2 for OPRM trips) are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels for application of SR 3.3.1.1.14, so n = 8. The note further requires that testing of OPRM and APRM outputs from a 2-Out-Of-4 Voter be alternated. In addition to these commitments, References 17 and 21 require that the testing of inputs to each RPS Trip System alternate.

Monticello B 3.3.1.1-35 Montcell B33.1.-35Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Combining these frequency requirements, an acceptable test sequence is one that:

a) Tests each RPS Trip System interface every other cycle, b) Alternates the testing of APRM and OPRM outputs from any specific 2-Out-Of-4 Voter channel, and c) Alternates between divisions at least every other test cycle.

After 8 cycles, the sequence repeats.

Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-Out-Of-4 Voter channel for that Function and each of the corresponding relays in the RPS. Consequently, each of the RPS relays is tested every fourth cycle. This RPS relay testing frequency is twice the frequency justified by References 17 and 21.

This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. The 24 month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

SR 3.3.1.1.15 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing -- applicable to Function 2.b and 2.f only), the 2-out-of-4 voter channels,, and the interface connections into the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 184 day Frequency of SR 3.3.1.1.15 is based on the reliability analysis of References 17 and 21. (NOTE: The actual voting logic of the 2-Out-Of-4 Voter Function is tested as part of SR 3.3.1.1.12.)

Note 1 is provided for Function 2.a to clarify that this SR is required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without Monticello B3.3.1.1-36 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) utilizing jumpers or lifted leads. Note 1 allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.

Note 2 is added to clarify that the CHANNEL FUNCTIONAL TEST is limited to the recirculation flow input processing and does not include the flow transmitters.

REFERENCES Instrumentation."

2. USAR, Section 7.6.1.2.1.

3. USAR, Section 7.6.1.2.5.

4. USAR, Chapter 14.

5. USAR, Chapter 14A.

6. USAR, Section 7.8.2.1.

7. USAR, Section 7.3.4.3.

8. Not Used.

9. USAR, Section 14.5.1.

10. USAR, Section 14.7.1.

11. USAR, Section 14.7.2.

12. USAR, Section 14.7.3.

13. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.

14. USAR, Section 14.4.5.

15. USAR, Section 14.4.1.

16. NEDC-30851 -P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.

Monticello B 3.3.1.1-37 Montcell B33.1.-37Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES REFERENCES (continued)

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

• 18. NEDC-33075P-A, Revision 6, "General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density,"

January 2008

19. Not Used.

20. Not Used.

21. NEDC-32410OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function", November 1997.

22. Letter from GEH to NRC, NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density (DSS-CD) Analytical Limit (TAC No.

MD0277)," dated October 29, 2008.

23. 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.

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

25. GHNE-O000-0073-4167-R2, "Reactor Long-Term Stability Solution Option Ill: Licensing Basis Hot Channel Oscillation Magnitude for Monticello Nuclear Generating Plant," December 2007.

26. CA-I10-135, "Instrument Setpoint Calculation - Intermediate Range Monitor (IRM) High Flux SCRAM and CR Block" (including Attachment 5, GEH document 0000-0121-5727, IRM Calibration Design Bases).

27. Calculation 09-239, "Turbine Bypass Valve Capacity for EPU"

28. Amendment No. 171, "Issuance of Amendment Regarding the Restoration Period Before Declaring Average Power Range Monitors Inoperable," dated January 25, 2013. (ADAMS Accession No. ML12339A035)

Revision No. 37 B 3.3.1.1-38 Monticello B 3.3.1.1-38 Revision No. 37

RPS Instrumentation B 3.3.1.1 BASES REFERENCES (continued)

29. MNGP EPU Task Report T0506, Revision 4, "Technical Specification Setpoints"

30. Amendment No. 176, "Monticello Nuclear Generating Plant -

Issuance of Amendment 176 to Renewed Facility Operating License Regarding Extended Power Uprate," December 9, 2013. (ADAMS Accession No. ML12339A035)

31. Amendment No. 180, "Monticello Nuclear Generating Plant -

Issuance of Amendment 180 to Renewed Facility Operating License Regarding Maximum Extended Operating Domain (MELLLA+),"

March 28, 2014. (ADAMS Accession No. ML14035A248)

Monticello Monticello 3.3.1.1-39

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