Text

Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control
	ML17293A280
Person / Time
Site:	Monticello
Issue date:	10/20/2017
From:	Church C; Northern States Power Company, Minnesota, Xcel Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-MT-17-058
	Download: ML17293A280 (184)
	v • d • e

2807 West County Road 75 Monticello, MN 55362 800.895.4999 xcelenergy.com October 20, 2017 L-MT-17-058 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), is submitting a request for an amendment to the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP).

The proposed change replaces existing Technical Specifications (TS) requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.4. Safety Limit 2.1.1.4 requires the reactor vessel water level to be greater than the top of active irradiated fuel.

provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.

Approval of the proposed amendment is requested by December 20, 2018. Once approved, the amendment shall be implemented prior to the next refueling outage. MNGP requests the option to utilize Enforcement Guidance Memorandum (EGM 11-003), Revision 3, Dispositioning Boiler Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel, until implementation of the amendment.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Minnesota Official.

If there are any questions or if additional information is needed, please contact Mr. Leonard Sueper at (612) 330-6917.

L-MT-17-058 Document Control Desk Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury, that the foregoing is true and correct.

Executed on October 20, 2017.

~~~

Christopher R. Church Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Attachments (4) cc:

Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC State of Minnesota

Page 1 of 9 ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT Evaluation of Proposed Change Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control"

1.

DESCRIPTION

2.

ASSESSMENT 2.1 Applicability of Published Safety Evaluation 2.2 Variations

3.

REGULATORY ANALYSIS 3.1 No Significant Hazards Consideration Analysis

4.

ENVIRONMENTAL EVALUATION

5.

REFERENCE

L-MT-17-058 Page 2 of 9 DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

The proposed change replaces the existing Technical Specifications (TS) requirements related to "operations which have the potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.4. Safety Limit 2.1.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), has reviewed the safety evaluation provided to the Technical Specifications Task Force (TSTF) on December 20, 2016, as well as the information provided in TSTF-542. NSPM has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the NRC staff are applicable to Monticello Nuclear Generating Plant (MNGP) and justify this amendment for the incorporation of the changes to the MNGP TS.

The following MNGP TS reference, or are related to, OPDRVs and are affected by the proposed change:

3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation 3.3.6.1, Primary Containment Isolation Instrumentation 3.3.6.2, Secondary Containment Isolation Instrumentation 3.3.7.1, Control Room Emergency Filtration (CREF) System Instrumentation 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring 3.5.2, ECCS - Shutdown 3.6.1.3, Primary Containment Isolation Valves (PCIVs) 3.6.4.1, Secondary Containment 3.6.4.2, Secondary Containment Isolation Valves (SCIVs) 3.6.4.3, Standby Gas Treatment (SGT) System 3.7.4, Control Room Emergency Filtration (CREF) System 3.7.5, Control Room Ventilation System 3.8.2, AC Sources - Shutdown 3.8.5, DC Sources - Shutdown 3.8.8, Distribution Systems - Shutdown 2.2 Variations NSPM is proposing the following variations from the TS changes described in the TSTF-542 or the applicable parts of the NRC staffs safety evaluation. These variations

L-MT-17-058 Page 3 of 9 do not affect the applicability of TSTF-542 or the NRC staff's safety evaluation to the proposed license amendment.

a. The MNGP TS, in some cases, utilizes different numbering and titles than the NUREG-1433 Standard TS on which TSTF-542 was based. The table below shows the differences between the plant-specific TS numbering and titles and the TSTF-542 numbering and titles. These differences are editorial and do not affect the applicability of TSTF-542 to the MNGP TS.

NUREG-1433 Numbering and Titles MNGP TS Numbering and Titles TS 2.1.1.3 Safety Limit TS 2.1.1.4 Safety Limit TS 3.3.7.1 Main Control Room Environmental Control (MCREC) System Instrumentation TS 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrumentation TS 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control.

TS 3.5.2 RPV Water Inventory Control TS 3.7.4 Main Control Room Environmental Control (MCREC) System TS 3.7.4 Control Room Emergency Filtration (CREF) System TS 3.7.5 Control Room Air Conditioning (AC)

System TS 3.7.5 Control Room Ventilation System TS 3.8.8 Inverters - Shutdown None None TS 3.8.10 Distribution Systems -

Shutdown TS 3.8.8 Distribution Systems -

Shutdown

b. TSTF-542 and the associated safety evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). MNGP was not licensed to the 10 CFR 50, Appendix A, GDC. The MNGP Updated Final Safety Analysis Report (UFSAR), Appendix E contains a comparative evaluation of the design basis of MNGP with respect to the 70 General Design Criteria for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission for public comment in July, 1967.

This difference does not alter the conclusion that the proposed change is applicable to MNGP.

c. The MNGP TS include Amendment No. 189 (Reference 1) for TSTF-523, "Generic Letter 2008-01, Managing Gas Accumulation." The following changes have no effect on the adoption of TSTF-542 and are an acceptable variation:

SR 3.5.2.2 has been modified from; Verify, for the required ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve," to; "Verify, for the required ECCS injection/spray

L-MT-17-058 Page 4 of 9 subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

SR 3.5.2.3 has been modified to retain the NOTE, "Not required to be met for system vent flow paths opened under administrative control."

d. NSPM has chosen to implement the Reactor Pressure Vessel Water Inventory Control (WIC) Instrumentation specification as TS 3.3.5.3 and to not renumber the existing TS 3.3.5.2.

e. The MNGP design does not include a Reactor Vessel Water Level - Low Low Low function. The corresponding MNGP TS Table 3.3.5.1-1 Functions 1.a and 2.a and TS Table 3.3.7.1-1 Function 1 occur on a Reactor Vessel Water Level -

Low Low signal. Therefore, it is appropriate to revise MNGP TS Tables 3.3.5.3.-1 and 3.3.7.1-1 to reflect this design difference.

f. NUREG-1433 Table 3.3.5.1-1, Function 1.d, Core Spray Pump Discharge Flow -

Low (Bypass), is not included in the MNGP TSs. As a result, this function is not being included in TS 3.3.5.3, Reactor Pressure Vessel (RPV) Water Inventory Control, Table 3.3.5.3-1.

g. MNGP TS Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation, does not include functions for manual initiation of Core Spray (CS) and Low Pressure Coolant Injection (LPCI). Since the TS does not include this feature, proposed Table 3.3.5.3-1 does not include manual initiation functions for CS and LPCI. In addition, TS Table 3.3.5.3-1 does not include a surveillance requirement (SR) for a logic system functional test since the SR applies only to the manual initiation function.

Therefore, as an alternative to NUREG-1433 SR 3.5.2.8, which demonstrates ECCS injection/spray actuation on a manual initiation signal, NSPM proposes that TS 3.5.2, RPV Water Inventory Control, include a new SR 3.5.2.5 to verify the required CS or LPCI subsystem can be manually operated through the manipulation of the subsystem components from the Control Room. The manual operation of the LPCI and Core Spray subsystems for the control of reactor cavity or RPV inventory are relatively simple evolutions and involve the manipulation of a small number of components. These subsystem alignments can be performed by licensed operators from the Control Room. This alternative is justified by the fact that a draining event is a slow evolution when compared to a design basis loss of coolant accident (LOCA), which is assumed to occur at full power, and thus there is adequate time to take manual actions (i.e., hours versus minutes). Adequate time to take action is assured since the proposed TS 3.5.2, Condition E, prohibits plant conditions that result in Drain Times that are less than one hour. Therefore, there is sufficient time for the licensed operators to take manual action to stop an unanticipated draining event, and to manually start an ECCS injection/spray subsystem or the additional method of water injection.

L-MT-17-058 Page 5 of 9 Consequently, there is no need for manual initiation logic to actuate the required subsystem components. Manual operation of the required subsystem would be an equivalent alternative to system initiation via manual initiation logic.

h. The MNGP design does not provide the capability to perform channel checks for the following NUREG-1433 functions in proposed Table 3.3.5.3-1, RPV Water Inventory Control Instrumentation: Function 1.c, Reactor Steam Dome Pressure

- Low (Injection Permissive), Function 2.c, Reactor Steam Dome Pressure - Low (Injection Permissive), and Function 2.d, Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) and are not in the current MNGP Table 3.3.5.1-1.

Therefore, channel checks are not included in Table 3.3.5.3-1 for these functions.

i. SR 3.5.2.4 has been modified to remove the phrase through the recirculation line to avoid potential confusion with the reactor recirculation lines or the Core Spray and RHR pump minimum flow lines, which are sometimes referred to as recirculation lines in plant procedures.

j. Rather than renumber the existing TS 3.5.2 surveillance requirements new NUREG-1433 SRs 3.5.2.1 and 3.5.2.7 have been numbered SR 3.5.2.7 and SR 3.5.2.6 respectively.

k. MNGP TS 3.8.5, DC Sources - Shutdown, does not include NUREG-1433 Condition A, its associated Required Actions or COMPLETION TIMES. Nor does MNGP TS 3.8.5 contain NUREG-1433 Required Action B.1. Therefore, MNGP TS 3.8.5 Required Action A.1 corresponds with NUREG-1433 Required Action B.2.

The MNGP TS contain the following requirements that differ from NUREG-1433 on which TSTF-542 was based, but are encompassed in the TSTF-542 justification:

l. MNGP TS Table 3.3.5.1-1, Functions 1.d and 2.d describe the Reactor Steam Dome Pressure Permissive-Low (Pump Permissive) functions for Core Spray and LPCI. These channels delay Core Spray and LPCI pump starts on a Reactor Vessel Water Level - Low Low until reactor steam dome pressure is below the setpoint. This ensures that, prior to starting low pressure ECCS subsystem pumps, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. This function is bypassed during manual operation; therefore, these Mode 4 and 5 functions and their corresponding surveillance requirements can be removed from the TS because the required ECCS subsystem is proposed to be started by manual operation.

m. MNGP TS Table 3.3.5.1-1, Functions 1.e and 2.e describe Reactor Steam Dome Pressure Permissive bypass time delays for the Core Spray and LPCI pumps.

The Bypass Timer channels allow the Core Spray and LPCI pumps to start on a Reactor Vessel Water Level - Low Low signal after the bypass timer times out, even if the reactor steam dome pressure is above its permissive setpoint. This

L-MT-17-058 Page 6 of 9 ensures that starting the pumps of the low pressure ECCS subsystems will occur on a Reactor Vessel Water Level - Low Low signal after an 18 minute time delay.

This time delay is unnecessary for manual operation; therefore, these Mode 4 and 5 functions and their corresponding surveillance requirements can be removed from the TS because the required ECCS subsystem is proposed to be started by manual operation.

n. MNGP TS Table 3.3.5.1-1, Function 1.f, describes the Core Spray Pump Start -

Time Delay Relays for the Core Spray pumps. This function is not included in NUREG-1433; however, it is similar to the LPCI Pump Start - Time Delay Relays (Function 2.f). The purpose of the time delay relays is to stagger the start of the CS and LPCI pumps that are in Divisions 1 and 2, thus limiting the starting transients on the 4.16 kV essential buses. This time delay is unnecessary for manual operation; therefore, this function can be removed from the TS.

o. The MNGP TS includes Table 3.3.7.1-1 Function 3, Reactor Building Ventilation Exhaust Radiation - High in lieu of a Control Room Air Inlet Radiation - High function (NUREG 1433 Table 3.3.7.1-1 Function 5). Like the Control Room Air Inlet Radiation - High function, this function initiates the Control Room Emergency Filtration (CREF) System to isolate the control room envelope from untreated outside air. Therefore, the Reactor Building Ventilation Exhaust Radiation - High function is being revised accordingly.

p. MNGP TS Section 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring, is being revised to remove references to OPDRVs from the Applicability of the LCO and from Condition F.

q. In the MNGP TS, the corresponding NUREG -1433 SR 3.5.2.2 and SR 3.5.2.3 are combined into a single SR 3.5.2.1 because the requirements for suppression pool water level and Condensate Storage Tank (CST) levels are applicable to both the CS and LPCI Systems, i.e., the CS and LPCI Systems both have the capability to draw a suction flow path from the CSTs or the Suppression Pool.

r. TSTF-542 inadvertently omitted the corresponding TS Bases markup for the deletion of TS 3.3.6.1 Required Action I.2 regarding actions to isolate RHR shutdown cooling. The MNGP TS Bases include changes consistent with this TS change.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests adoption of TSTF-542, "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications (STS), into the Monticello Nuclear Generating Plant (MNGP) Technical

L-MT-17-058 Page 7 of 9 Specifications (TS). The proposed amendment replaces the existing requirements in the TS related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.4. Safety Limit 2.1.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.

NSPM has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.4. Draining of Reactor Pressure Vessel (RPV) water inventory in Mode 4 (cold shutdown) and Mode 5 (refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed change reduces the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed change reduces the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as

L-MT-17-058 Page 8 of 9 verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.4. The proposed change will not alter the design function of the equipment involved.

Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The

L-MT-17-058 Page 9 of 9 safety basis for the new requirements is to protect Safety Limit 2.1.1.4. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the TAF in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NSPM concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

5.0 REFERENCE

1.

Monticello Nuclear Generating Plant - Issuance of Amendment Re: Technical Specifications Task Force Standard Technical Specifications Change Traveler TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation, ADAMS Accession No. ML16125A165

2.

Final Safety Evaluation of Technical Specifications Task Force Traveler TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control" (TAC No. MF3487) ADAMS Accession No. ML16343B065

ATTACHMENT 2 MONTICELLO NUCLEAR GENERATING PLANT Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" TECHNICAL SPECIFICATION PAGES (Markups) 47 pages follow

Monticello i

Revision 01 TABLE OF CONTENTS Page Number 1.0 USE AND APPLICATION 1.1 Definitions............................................................................................................ 1.1-1 1.2 Logical Connectors.............................................................................................. 1.2-1 1.3 Completion Times................................................................................................ 1.3-1 1.4 Frequency........................................................................................................... 1.4-1 2.0 SAFETY LIMITS (SLs).............................................................................................. 2.0-1 2.1 Safety Limits 2.2 SL Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.......................... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY......................................... 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM).................................................................... 3.1.1-1 3.1.2 Reactivity Anomalies................................................................................... 3.1.2-1 3.1.3 Control Rod OPERABILITY........................................................................ 3.1.3-1 3.1.4 Control Rod Scram Times........................................................................... 3.1.4-1 3.1.5 Control Rod Scram Accumulators............................................................... 3.1.5-1 3.1.6 Rod Pattern Control.................................................................................... 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System......................................................... 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves............................. 3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)............................................................................................. 3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)........................................... 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optional)............................ 3.2.3-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation................................... 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation......................................... 3.3.1.2-1 3.3.2.1 Control Rod Block Instrumentation........................................................... 3.3.2.1-1 3.3.2.2 Feedwater Pump and Main Turbine High Water Level Trip Instrumentation.................................................................................. 3.3.2.2-1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation..................................... 3.3.3.1-1 3.3.3.2 Alternate Shutdown System..................................................................... 3.3.3.2-1 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation............................................................. 3.3.4.1-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation...................... 3.3.5.1-1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation.......... 3.3.5.2-1 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation3.3.5.3-1 3.3.6.1 Primary Containment Isolation Instrumentation........................................ 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation................................... 3.3.6.2-1 3.3.6.3 Low-Low Set (LLS) Instrumentation......................................................... 3.3.6.3-1 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrumentation.................................................................................. 3.3.7.1-1 3.3.7.2 Mechanical Vacuum Pump Isolation Instrumentation............................... 3.3.7.2-1 3.3.8.1 Loss of Power (LOP) Instrumentation...................................................... 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring..3.3.8.2-1

Monticello ii Revision 01 TABLE OF CONTENTS Page Number 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating.................................................................... 3.4.1-1 3.4.2 Jet Pumps................................................................................................... 3.4.2-1 3.4.3 Safety/Relief Valves (S/RVs)...................................................................... 3.4.3-1 3.4.4 RCS Operational LEAKAGE....................................................................... 3.4.4-1 3.4.5 RCS Leakage Detection Instrumentation.................................................... 3.4.5-1 3.4.6 RCS Specific Activity.................................................................................. 3.4.6-1 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown.............................................................................................. 3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown.............................................................................................. 3.4.8-1 3.4.9 RCS Pressure and Temperature (P/T) Limits.............................................. 3.4.9-1 3.4.10 Reactor Steam Dome Pressure................................................................ 3.4.10-1 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating....................................................................................... 3.5.1-1 3.5.2 RPV Water Inventory ControlECCS - Shutdown......................................... 3.5.2-1 3.5.3 RCIC System.............................................................................................. 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment............................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Lock.................................................................. 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs)........................................ 3.6.1.3-1 3.6.1.4 Drywell Air Temperature.......................................................................... 3.6.1.4-1 3.6.1.5 Low-Low Set (LLS) Valves....................................................................... 3.6.1.5-1 3.6.1.6 Reactor Building-to-Suppression Chamber Vacuum Breakers................. 3.6.1.6-1 3.6.1.7 Suppression Chamber-to-Drywell Vacuum Breakers............................... 3.6.1.7-1 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray.......................................... 3.6.1.8-1 3.6.2.1 Suppression Pool Average Temperature................................................. 3.6.2.1-1 3.6.2.2 Suppression Pool Water Level................................................................. 3.6.2.2-1 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling....................... 3.6.2.3-1 3.6.3.1 Primary Containment Oxygen Concentration........................................... 3.6.3.1-1 3.6.4.1 Secondary Containment........................................................................... 3.6.4.1-1 3.6.4.2 Secondary Containment Isolation Valves (SCIVs).................................... 3.6.4.2-1 3.6.4.3 Standby Gas Treatment (SGT) System.................................................... 3.6.4.3-1 3.7 PLANT SYSTEMS 3.7.1 Residual Heat Removal Service Water (RHRSW) System.......................... 3.7.1-1 3.7.2 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS)............................................................................................ 3.7.2-1 3.7.3 Emergency Diesel Generator Emergency Service Water (EDG-ESW) System.............................................................................. 3.7.3-1 3.7.4 Control Room Emergency Filtration (CREF) System................................... 3.7.4-1 3.7.5 Control Room Ventilation System............................................................... 3.7.5-1 3.7.6 Main Condenser Offgas.............................................................................. 3.7.6-1 3.7.7 Main Turbine Bypass System..................................................................... 3.7.7-1 3.7.8 Spent Fuel Storage Pool Water Level......................................................... 3.7.8-1

Definitions 1.1 Monticello 1.1-2 Amendment No. 146, 148, XXX 1.1 Definitions CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

The following exceptions are not considered to be CORE ALTERATIONS:

a.

Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and

b.

Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.3. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR)-11, "Limiting Values of Radionuclide Intake and Air Concentration Factors for Inhalation, Submersion and Ingestion," September 1988, and FGR-12, "External Exposure to Radionuclides in Air, Water and Soil,"

September 1993.

DRAIN TIME The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming:

a)

The water inventory above the TAF is divided by the limiting drain rate; b)

The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single

Definitions 1.1 Monticello 1.1-3 Amendment No. 194, XXX 1.1 Definitions DRAIN TIME (continued) human error), for all penetration flow paths below the TAF except:

1.

Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;

2.

Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation; or

3.

Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.

c)

The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory; d)

No additional draining events occur; and e)

Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-1 Amendment No. 146, XXX 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.5.1-1.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels inoperable.

A.1 Enter the Condition referenced in Table 3.3.5.1-1 for the channel.

Immediately B. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

B.1

NOTES-------------

1. Only applicable in MODES 1, 2, and 3.

2. Only applicable for Functions 1.a, 1.b, 2.a, 2.b, 2.f, 2.h, and 2.k.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

AND 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divisions

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.2

NOTE--------------

Only applicable for Functions 3.a and 3.b.

Declare High Pressure Coolant Injection (HPCI)

System inoperable.

AND B.3 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

C.1

NOTES-------------

1. Only applicable in MODES 1, 2, and 3.

2. Only applicable for Functions 1.c, 1.d, 1.e, 1.f, 2.c, 2.d, 2.e, 2.i, 2.j, 2.l, and 2.m.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

AND C.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divisions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-3 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

D.1

NOTE--------------

Only applicable if HPCI pump suction is not aligned to the suppression pool.

Declare HPCI System inoperable.

AND D.2.1 Place channel in trip.

OR D.2.2 Align the HPCI pump suction to the suppression pool.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours E. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

E.1

NOTES-------------

1. Only applicable in MODES 1, 2, and 3.

2. Only applicable for Function 2.g.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

AND E.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for subsystems in both divisions 7 days

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-6 Amendment No. 146, 151, 170, 176, XXX Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

a.

Reactor Vessel Water Level - Low Low 1, 2, 3, 4(a), 5(a) 4(ab)

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Drywell Pressure -

High 1, 2, 3 4(ab)

B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.8 2 psig

c.

Reactor Steam Dome Pressure -

Low (Injection Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 397 psig and 440 psig 4(a), 5(a) 2 B

SR 3.3.5.1.2 SR 3.3.5.1.4(c)(d)

SR 3.3.5.1.8 397 psig and 440 psig

d.

Reactor Steam Dome Pressure Permissive - Low (Pump Permissive) 1, 2, 3 4(a), 5(a) 2 2

C B

SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 SR 3.3.5.1.2 SR 3.3.5.1.4(c)(d)

SR 3.3.5.1.8 397 psig 397 psig

e.

Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive) 1, 2, 3 4(a), 5(a) 2 2

C B

SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.7 SR 3.3.5.1.8 18 minutes 18 minutes (a)

When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2, "ECCS - Shutdown."

(ab) Also required to initiate the associated emergency diesel generator (EDG).

(bc) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(cd) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the Technical Requirements Manual (TRM).

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-7 Amendment No. 146, XXX Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

f.

Core Spray Pump Start - Time Delay Relay 1, 2, 3 4(a), 5(a) 1 per pump C

SR 3.3.5.1.7 SR 3.3.5.1.8 15.86 seconds

2. Low Pressure Coolant Injection (LPCI) System

a.

Reactor Vessel Water Level - Low Low 1, 2, 3, 4(a), 5(a) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Drywell Pressure -

High 1, 2, 3 4

B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.8 2 psig

c.

Reactor Steam Dome Pressure -

Low (Injection Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 397 psig and 440 psig 4(a), 5(a) 2 B

SR 3.3.5.1.2 SR 3.3.5.1.4(c)(d)

SR 3.3.5.1.8 397 psig and 440 psig

d.

Reactor Steam Dome Pressure Permissive - Low (Pump Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 397 psig 4(a), 5(a) 2 B

SR 3.3.5.1.2 SR 3.3.5.1.4(c)(d)

SR 3.3.5.1.8 397 psig (a)

When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2.

(bc) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(cd) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-8 Correction letter of 10/12/06 Amendment No. 146, 149, 151, 161, 170 176, XXX Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

2. LPCI System

e.

Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive) 1, 2, 3 4(a), 5(a) 2 2

C B

SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.7 SR 3.3.5.1.8 18 minutes 18 minutes

f.

Low Pressure Coolant Injection Pump Start - Time Delay Relay 1, 2, 3, 4(a), 5(a) 4 per pump B

SR 3.3.5.1.7 SR 3.3.5.1.8 Pumps A, B 5.33 seconds Pumps C, D 10.59 seconds

g.

Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 1, 2, 3, 4(a), 5(a) 1 per pump E

SR 3.3.5.1.2 SR 3.3.5.1.7 SR 3.3.5.1.8 360 gpm and 745 gpm

h.

Reactor Steam Dome Pressure -

Low (Break Detection) 1, 2, 3 4

B SR 3.3.5.1.2 SR 3.3.5.1.7 SR 3.3.5.1.8 873.6 psig and 923.4 psig

i.

Recirculation Pump Differential Pressure

- High (Break Detection) 1, 2, 3 4 per pump C

SR 3.3.5.1.2 SR 3.3.5.1.7 SR 3.3.5.1.8 63.5 inches wc

j.

Recirculation Riser Differential Pressure

- High (Break Detection) 1, 2, 3 4

C SR 3.3.5.1.2 SR 3.3.5.1.7(bc)(cd)

SR 3.3.5.1.8 100.0 inches wc (a)

When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2.

(bc) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(cd) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-9 Amendment No. 146, 149, 151, 161, XXX Table 3.3.5.1-1 (page 4 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

2. LPCI System

k.

Recirculation Steam Dome Pressure -

Time Delay Relay (Break Detection) 1, 2, 3 2

B SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.9 2.97 seconds

l.

Recirculation Pump Differential Pressure

- Time Delay Relay (Break Detection) 1, 2, 3 2

C SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.9 0.75 seconds

m. Recirculation Riser Differential Pressure

- Time Delay Relay (Break Detection) 1, 2, 3 2

C SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.9 0.75 seconds

3. High Pressure Coolant Injection (HPCI) System

a.

Reactor Vessel Water Level - Low Low 1, 2(de), 3(de) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Drywell Pressure -

High 1, 2(de), 3(de) 4 B

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.8 2 psig

c.

Reactor Vessel Water Level - High 1, 2(de), 3(de) 2 C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8 48 inches

d.

Condensate Storage Tank Level - Low 1, 2(de), 3(de) 2 D

SR 3.3.5.1.7 SR 3.3.5.1.8 29.3 inches

e.

Suppression Pool Water Level - High 1, 2(de), 3(de) 2 D

SR 3.3.5.1.5 SR 3.3.5.1.6 SR 3.3.5.1.8 3.0 inches

f.

High Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 1, 2(de), 3(de) 1 E

SR 3.3.5.1.5 SR 3.3.5.1.6 SR 3.3.5.1.8 362 gpm and 849 gpm (de) With reactor steam dome pressure > 150 psig.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-10 Amendment No. 146, XXX Table 3.3.5.1-1 (page 5 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

4. Automatic Depressurization System (ADS) Trip System A

a.

Reactor Vessel Water Level - Low Low 1, 2(de), 3(de) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Automatic Depressurization System Initiation Timer 1, 2(de), 3(de) 1 G

SR 3.3.5.1.7 SR 3.3.5.1.8 120 seconds

c.

Core Spray Pump Discharge Pressure

- High 1, 2(de), 3(de) 2 G

SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 75 psig and 125 psig

d.

Low Pressure Coolant Injection Pump Discharge Pressure - High 1, 2(de), 3(de) 4 G

SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 75 psig and 125 psig

5. ADS Trip System B

a.

Reactor Vessel Water Level - Low Low 1, 2(de), 3(de) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Automatic Depressurization System Initiation Timer 1, 2(de), 3(de) 1 G

SR 3.3.5.1.7 SR 3.3.5.1.8 120 seconds (bc) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(cd) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

(de) With reactor steam dome pressure > 150 psig.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-11 Amendment No. 146, XXX Table 3.3.5.1-1 (page 6 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5. ADS Trip System B

c.

Core Spray Pump Discharge Pressure

- High 1, 2(de), 3(de) 2 G

SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 75 psig and 125 psig

d.

Low Pressure Coolant Injection Pump Discharge Pressure - High 1, 2(de), 3(de) 4 G

SR 3.3.5.1.2 SR 3.3.5.1.4(bc)(c d)

SR 3.3.5.1.8 75 psig and 125 psig (bc) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(cd) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

(de) With reactor steam dome pressure > 150 psig.

RPV Water Inventory Control Instrumentation 3.3.5.3 Monticello 3.3.5.3-1 Amendment No. XXX 3.3 INSTRUMENTATION 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.3 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.5.3-1.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels inoperable.

A.1 Enter the Condition referenced in Table 3.3.5.3-1 for the channel.

Immediately B. As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

B.1 Declare associated penetration flow path(s) incapable of automatic isolation.

AND B.2 Calculate DRAIN TIME.

Immediately Immediately C. As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

C.1 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

D.1 Restore channel to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

RPV Water Inventory Control Instrumentation 3.3.5.3 Monticello 3.3.5.3-2 Amendment No. XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and associated Completion Time of Condition C or D not met.

E.1 Declare associated low pressure ECCS injection/spray subsystem inoperable.

Immediately SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

Refer to Table 3.3.5.3-1 to determine which SRs apply for each ECCS Function.

SURVEILLANCE FREQUENCY SR 3.3.5.3.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.5.3.2 Perform CHANNEL FUNCTIONAL TEST.

92 days

RPV Water Inventory Control Instrumentation 3.3.5.3 Monticello 3.3.5.3-3 Amendment No. XXX Table 3.3.5.3-1 (page 1 of 1)

RPV Water Inventory Control Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

a.

Reactor Steam Dome Pressure - Low (Injection Permissive) 4, 5 2

C SR 3.3.5.3.2 397 psig and 440psig

2. Low Pressure Coolant Injection (LPCI)

System

a.

Reactor Steam Dome Pressure -

Low (Injection Permissive) 4, 5 2

C SR 3.3.5.3.2 397 psig and 440 psig

b.

Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 4, 5 1 per pump(a)

D SR 3.3.5.3.2 360 gpm and 745 gpm

3. RHR System Isolation

a.

Reactor Vessel Water Level - Low (b) 2 in one trip system B

SR 3.3.5.3.1 SR 3.3.5.3.2 7 inches

4. Reactor Water Cleanup (RWCU) System Isolation

a.

Reactor Vessel Water Level - Low Low (b) 2 in one trip system B

SR 3.3.5.3.1 SR 3.3.5.3.2

-48 inches (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "RPV Water Inventory Control."

(b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.

Primary Containment Isolation Instrumentation 3.3.6.1 Monticello 3.3.6.1-3 Amendment No. 146 XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I.

As required by Required Action C.1 and referenced in Table 3.3.6.1-1.

I.1 Initiate action to restore channel to OPERABLE status.

OR I.2 Initiate action to isolate the Residual Heat Removal (RHR) Shutdown Cooling System.

Immediately Immediately SURVEILLANCE REQUIREMENTS

NOTES----------------------------------------------------------

1.

Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2.

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains primary containment isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.6.1.3 Calibrate the trip unit.

92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION.

92 days

Primary Containment Isolation Instrumentation 3.3.6.1 Monticello 3.3.6.1-7 Amendment No. 146, 164, XXX Table 3.3.6.1-1 (page 3 of 3)

Primary Containment Isolation Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION C.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5. RWCU System Isolation

d.

SLC System Initiation 1, 2, 3 1

H SR 3.3.6.1.6 NA

e.

Reactor Vessel Water Level - Low Low 1, 2, 3 2

F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6

-48 inches

6. Shutdown Cooling System Isolation

a.

Reactor Steam Dome Pressure -

High 1, 2, 3 2

F SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 81.8 psig

b.

Reactor Vessel Water Level - Low 3, 4, 5 2(a)

I SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 7 inches

7. Traversing Incore Probe System Isolation

a.

Reactor Vessel Water Level - Low 1, 2, 3 2

G SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 7 inches

b.

Drywell Pressure -

High 1, 2, 3 2

G SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 2.0 psig (a)

Only one channel per trip system, with an isolation signal available to one shutdown cooling supply isolation valve, is required in MODES 4 and 5, provided RHR Shutdown Cooling System integrity is maintained.

Secondary Containment Isolation Instrumentation 3.3.6.2 Monticello 3.3.6.2-3 Amendment No. 146, XXX Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Reactor Vessel Water Level - Low Low 1, 2, 3, (a) 2 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

-48 inches

2. Drywell Pressure - High 1, 2, 3 2

SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 2 psig

3. Reactor Building Ventilation Exhaust Radiation - High 1, 2, 3, (a), (b) 2 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 100 mR/hr

4. Refueling Floor Radiation - High 1, 2, 3, (a), (b) 2 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 100 mR/hr (a)

During operations with a potential for draining the reactor vessel.

(ab) During movement of recently irradiated fuel assemblies in secondary containment.

CREF System Instrumentation 3.3.7.1 Monticello 3.3.7.1-3 Amendment No. 146, 148, XXX Table 3.3.7.1-1 (Page 1 of 1)

Control Room Emergency Filtration System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1, 2, 3, (a) 2 SR 3.3.7.1.1

- 48 inches Level - Low Low SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6

2. Drywell Pressure -

1, 2, 3 2

SR 3.3.7.1.2 2 psig High SR 3.3.7.1.4 SR 3.3.7.1.6

3. Reactor Building 1, 2, 3, (a), (b) 2 SR 3.3.7.1.1 100 mR/hr Ventilation Exhaust SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.4 SR 3.3.7.1.6

4. Refueling Floor 1, 2, 3, (a), (b) 2 SR 3.3.7.1.1 100 mR/hr Radiation - High SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.6 (a)

During operations with a potential for draining the reactor vessel.

(b)(a)

During movement of recently irradiated fuel assemblies in the secondary containment.

RPS Electric Power Monitoring 3.3.8.2 Monticello 3.3.8.2-1 Amendment No. 146, XXX 3.3 INSTRUMENTATION 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring LCO 3.3.8.2 Two RPS electric power monitoring assemblies shall be OPERABLE for each inservice RPS motor generator set or alternate power supply.

APPLICABILITY:

MODES 1, 2, and 3, MODES 4 and 5 with residual heat removal (RHR) shutdown cooling supply isolation valves open, MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both inservice power supplies with one electric power monitoring assembly inoperable.

A.1 Remove associated inservice power supply(s) from service.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. One or both inservice power supplies with both electric power monitoring assemblies inoperable.

B.1 Remove associated inservice power supply(s) from service.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

RPS Electric Power Monitoring 3.3.8.2 Monticello 3.3.8.2-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition A or B not met in MODE 4 or 5 with RHR shutdown cooling supply isolation valves open.

D.1 Initiate action to restore one electric power monitoring assembly to OPERABLE status for inservice power supply(s) supplying required instrumentation.

OR D.2 Initiate action to isolate the RHR Shutdown Cooling System.

Immediately Immediately E. Required Action and associated Completion Time of Condition A or B not met in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies.

E.1 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Immediately F. Required Action and associated Completion Time of Condition A or B not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

F.1.1 Isolate the associated secondary containment penetration flow path(s).

OR F.1.2 Declare the associated secondary containment isolation valve(s) inoperable.

AND F.2.1 Place the associated standby gas treatment (SGT) subsystem(s) in operation.

OR Immediately Immediately Immediately

ECCS - Operating 3.5.1 Monticello 3.5.1-1 Amendment No. 146, XXX 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM (RCIC) 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of three safety/relief valves shall be OPERABLE.

NOTE--------------------------------------------

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling supply isolation interlock in MODE 3, if capable of being manually realigned and not otherwise inoperable.

APPLICABILITY:

MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One LPCI pump inoperable.

A.1 Restore LPCI pump to OPERABLE status.

30 days B. One LPCI subsystem inoperable for reasons other than Condition A.

OR One Core Spray subsystem inoperable.

B.1 Restore low pressure ECCS injection/spray subsystem to OPERABLE status.

7 days

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-1 Amendment No. 146, XXX 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) SYSTEM 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control ECCS - Shutdown LCO 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AND OneTwo low pressure ECCS injection/spray subsystems shall be OPERABLE.

NOTE--------------------------------------------

A One Low Pressure Coolant Injection (LPCI) subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

APPLICABILITY:

MODES 4,

MODE and 5, except with the spent fuel storage pool gates removed and water level 21 ft 11 inches over the top of the reactor pressure vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Rrequired ECCS injection/spray subsystem inoperable.

A.1 Restore required ECCS injection/spray subsystem to OPERABLE status.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion Time of Condition A not met.

B.1 Initiate action to establish a method of water injection capable of operating without offsite electrical power.Initiate action to suspend operations with a potential for draining the reactor vessel (OPDRVs).

Immediately

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-2 Amendment No. 146, XXX C. Two required ECCS injection/spray subsystems inoperable.

C.1 Initiate action to suspend OPDRVs.

AND C.2 Restore one required ECCS injection/spray subsystem to OPERABLE status.

Immediately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.1 Verify secondary containment boundary is capable of being established in less than the DRAIN TIME.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-3 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.2 Verify each secondary containment penetration flow path is capable of being isolated in less than the DRAIN TIME.

AND C.3 Verify one standby gas treatment subsystem is capable of being placed in operation in less than the DRAIN TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours D. DRAIN TIME < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Required Action C.2 and associated Completion Time not met.

D.1

NOTE ---------------

Required ECCS injection/spray subsystem or additional method of water injection shall be capable of operating without offsite electrical power.

Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AND D.21 Initiate action to restore secondary containment boundaryto OPERABLE status.

AND Immediately Immediately

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-4 Amendment No. 146, XXX D.2 Initiate action to restore one standby gas treatment subsystem to OPERABLE status.

AND D.3 Initiate action to restore isolation capability in each required secondary containment penetration flow path not isolated.

Immediately ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.3 Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room.

AND D.4 Initiate action to verify one standby gas treatment subsystem is capable of being placed in operation.

Immediately Immediately E. Required Action and associated Completion Time of Condition C or D not met.

OR DRAIN TIME < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

E.1 Initiate action to restore DRAIN TIME to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Immediately SURVEILLANCE REQUIREMENTS

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-5 Amendment No. 146, XXX SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for eachthe required ECCS injection/spray subsystem, the:

a. Suppression pool water level is -3 ft; or

b. ---------------------------NOTE----------------------------

Only one required ECCS injection/spray subsystem may take credit for this option during OPDRVs.

Condensate storage tank(s) water level is 7 ft for one tank operation and 4 ft for two tank operation.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-6 Amendment No. 146, 167, 189, 194, XXX SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for eachthe required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

31 days SR 3.5.2.3

NOTE------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify, for eachthe required ECCS injection/spray subsystem, each manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

31 days SR 3.5.2.4 Operate the required ECCS injection/spray subsystem for 10 minutes. Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor to containment pressure.

System Head Corresponding to a Reactor to No. of Containment System Flow Rate Pumps Pressure of Core Spray 2835 gpm 1 130 psi LPCI 3870 gpm 1 20 psi In accordance with the INSERVICE TESTING PROGRAM92 days SR 3.5.2.5

NOTE------------------------------

Vessel injection/spray may be excluded.

Verify the each required ECCS injection/spray subsystem actuates on can be manually operatedan actual or simulated automatic initiation signal.

24 months SR 3.5.2.6 Verify each valve credited for automatically isolating 24 months

RPV Water Inventory ControlECCS - Shutdown 3.5.2 Monticello 3.5.2-7 Amendment No. 146, 167, 189, 194, XXX SURVEILLANCE FREQUENCY a penetration flow path actuates to the isolation position on an actual or simulated isolation signal.

SR 3.5.2.7 Verify DRAIN TIME 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

RCIC System 3.5.3 Monticello 3.5.3-1 Amendment No. 146, XXX 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM (RCIC) 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY:

MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to the RCIC System.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable.

A.1 Verify by administrative means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System to OPERABLE status.

Immediately 14 days B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Reduce reactor steam dome pressure to 150 psig.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

PCIVs 3.6.1.3 Monticello 3.6.1.3-6 Amendment No. 146, 148, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and associated Completion Time of Condition A or B not met for PCIV(s) required to be OPERABLE during MODE 4 or 5.

G.1 Initiate action to suspend operations with a potential for draining the reactor vessel (OPDRVs).

OR G.12 Initiate action to restore valve(s) to OPERABLE status.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1

NOTE------------------------------

Not required to be met when the 18 inch primary containment purge and vent valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

Verify each 18 inch primary containment purge and vent valve is closed.

31 days

Secondary Containment 3.6.4.1 Monticello 3.6.4.1-1 Amendment No. 146, XXX 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment inoperable in MODE 1, 2, or 3.

A.1 Restore secondary containment to OPERABLE status.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion Time of Condition A not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours C. Secondary containment inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

C.1

NOTE--------------

LCO 3.0.3 is not applicable.

Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND C.2 Initiate action to suspend OPDRVs.

Immediately Immediately

SCIVs 3.6.4.2 Monticello 3.6.4.2-1 Amendment No. 146, XXX 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

NOTES----------------------------------------------------------

1.

Penetration flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration flow paths with one SCIV inoperable.

A.1 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

SCIVs 3.6.4.2 Monticello 3.6.4.2-3 Amendment No.146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition A or B not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

D.1

NOTE--------------

LCO 3.0.3 is not applicable.

Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND D.2 Initiate action to suspend OPDRVs.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1

NOTES-----------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment isolation manual valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

31 days SR 3.6.4.2.2 Verify the isolation time of each power operated, automatic SCIV is within limits.

92 days SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation position on an actual or simulated actuation signal.

24 months

SGT System 3.6.4.3 Monticello 3.6.4.3-1 Amendment No. 146, XXX 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem inoperable.

A.1 Restore SGT subsystem to OPERABLE status.

7 days B. Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours C. Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

NOTE-------------------

LCO 3.0.3 is not applicable.

C.1 Place OPERABLE SGT subsystem in operation.

OR C.2.1 Suspend movement of recently irradiated fuel assemblies in secondary containment.

AND Immediately Immediately

SGT System 3.6.4.3 Monticello 3.6.4.3-2 Amendment No. 146, 181, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.2.2 Initiate action to suspend OPDRVs.

Immediately D. Two SGT subsystems inoperable in MODE 1, 2, or 3.

D.1 Enter LCO 3.0.3.

Immediately E. Two SGT subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

E.1

NOTE--------------

LCO 3.0.3 is not applicable.

Suspend movement of recently irradiated fuel assemblies in secondary containment.

AND E.2 Initiate action to suspend OPDRVs.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for 15 continuous minutes.

31 days SR 3.6.4.3.2 Perform required SGT filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or simulated initiation signal.

24 months

CREF System 3.7.4 Monticello 3.7.4-1 Amendment No. 146, 160, 175, XXX 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Filtration (CREF) System LCO 3.7.4 Two CREF subsystems shall be OPERABLE.

NOTE--------------------------------------------

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.,

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREF subsystem inoperable for reasons other than Condition B.

A.1 Restore CREF subsystem to OPERABLE status.

7 days B. One or more CREF subsystems inoperable due to inoperable CRE boundary in MODE 1, 2, or 3.

B.1 Initiate action to implement mitigating actions.

AND B.2 Verify mitigating actions ensure CRE occupant exposures to radiological, chemical and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to OPERABLE status.

Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 90 days

CREF System 3.7.4 Monticello 3.7.4-2 Amendment No. 146, 160, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours D. Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

NOTE-------------------

LCO 3.0.3 is not applicable.

D.1 Place OPERABLE CREF subsystem in pressurization mode.

OR D.2.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND D.2.2 Initiate action to suspend OPDRVs.

Immediately Immediately Immediately E. Two CREF subsystems inoperable in MODE 1, 2, or 3 for reasons other than Condition B.

E.1 Enter LCO 3.0.3.

Immediately

CREF System 3.7.4 Monticello 3.7.4-3 Amendment No. 146, 160, 175, 181, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREF subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

OR One or more CREF subsystems inoperable due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

NOTE-------------------

LCO 3.0.3 is not applicable.

F.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND F.2 Initiate action to suspend OPDRVs.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate each CREF subsystem for 15 continuous minutes.

31 days SR 3.7.4.2 Perform required CREF filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.4.3 Verify each CREF subsystem actuates on an actual or simulated initiation signal.

24 months SR 3.7.4.4 Perform required CRE unfiltered air in-leakage testing in accordance with the Control Room Envelope Habitability Program.

In accordance with the Control Room Envelope Habitability Program

Control Room Ventilation System 3.7.5 Monticello 3.7.5-1 Amendment No. 146, 148, 154, XXX 3.7 PLANT SYSTEMS 3.7.5 Control Room Ventilation System LCO 3.7.5 Two control room ventilation subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment,.

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One control room ventilation subsystem inoperable.

A.1 Restore control room ventilation subsystem to OPERABLE status.

30 days B. Two control room ventilation subsystems inoperable.

B.1 Verify control room area temperature < 90°F.

AND B.2 Restore one control room ventilation subsystem to OPERABLE status.

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 72 hours C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours D. Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies in the secondary containment or during OPDRVs.

NOTE-------------------

LCO 3.0.3 is not applicable.

D.1 Place OPERABLE control room ventilation subsystem in operation.

OR Immediately

Control Room Ventilation System 3.7.5 Monticello 3.7.5-2 Amendment No. 146, 148, 154, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.2.1 Suspend movement of irradiated fuel assemblies in the secondary containment.

AND D.2.2 Initiate action to suspend OPDRVs.

Immediately Immediately E. Required Action and associated Completion Time of Condition B not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.

NOTE-------------------

LCO 3.0.3 is not applicable.

E.1 Suspend movement of irradiated fuel assemblies in the secondary containment.

AND E.2 Initiate actions to suspend OPDRVs.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each control room ventilation subsystem has the capability to remove the assumed heat load.

24 months

AC Sources - Shutdown 3.8.2 Monticello 3.8.2-2 Amendment No. 146, 148, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND A.2.3 Initiate action to suspend operations with a potential for draining the reactor vessel (OPDRVs).

AND A.2.34 Initiate action to restore required offsite power circuit to OPERABLE status.

Immediately Immediately Immediately Immediately B. One required EDG inoperable.

B.1 Suspend CORE ALTERATIONS.

AND B.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND B.3 Initiate action to suspend OPDRVs.

AND B.34 Initiate action to restore required EDG to OPERABLE status.

Immediately Immediately Immediately Immediately

AC Sources - Shutdown 3.8.2 Monticello 3.8.2-3 Amendment No. 146, XXX SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1

NOTES-----------------------------

1. The following SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.7, and SR 3.8.1.8 through SR 3.8.1.13.

2. SR 3.8.1.8 and SR 3.8.1.12 are not required to be met when associated ECCS subsystem(s) are not required to be OPERABLE per LCO 3.5.2, "RPV Water Inventory ControlECCS

- Shutdown."

For AC sources required to be OPERABLE the SRs of Specification 3.8.1, except SR 3.8.1.6, are applicable.

In accordance with applicable SRs

DC Sources - Shutdown 3.8.5 Monticello 3.8.5-1 Amendment No. 146, 148, XXX 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 Division 1 or Division 2 125 VDC electrical power subsystem shall be OPERABLE to support one division of the DC Electrical Power Distribution System required by LCO 3.8.8, "Distribution Systems -

Shutdown."

APPLICABILITY:

MODES 4 and 5, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Required DC electrical power subsystem inoperable.

A.1 Suspend CORE ALTERATIONS.

AND A.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND A.3 Initiate action to suspend operations with a potential for draining the reactor vessel.

AND A.34 Initiate action to restore required DC electrical power subsystem to OPERABLE status.

Immediately Immediately Immediately Immediately

Distribution Systems - Shutdown 3.8.8 Monticello 3.8.8-1 Amendment No. 146, 148, XXX 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems - Shutdown LCO 3.8.8 The necessary portions of the AC and DC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY:

MODES 4 and 5, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required AC or DC electrical power distribution subsystems inoperable.

A.1 Declare associated supported required feature(s) inoperable.

OR A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend handling of recently irradiated fuel assemblies in the secondary containment.

AND A.2.3 Initiate action to suspend operations with a potential for draining the reactor vessel.

AND Immediately Immediately Immediately Immediately

Distribution Systems - Shutdown 3.8.8 Monticello 3.8.8-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2.34 Initiate actions to restore required AC and DC electrical power distribution subsystems to OPERABLE status.

AND A.2.45 Declare associated required shutdown cooling subsystem(s) inoperable and not in operation.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and voltage to required AC and DC electrical power distribution subsystems.

7 days

ATTACHMENT 3 MONTICELLO NUCLEAR GENERATING PLANT Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" TECHNICAL SPECIFICATION PAGES (Clean Typed) 44 pages follow

Monticello i

Revision 1 TABLE OF CONTENTS Page Number 1.0 USE AND APPLICATION 1.1 Definitions............................................................................................................ 1.1-1 1.2 Logical Connectors.............................................................................................. 1.2-1 1.3 Completion Times................................................................................................ 1.3-1 1.4 Frequency........................................................................................................... 1.4-1 2.0 SAFETY LIMITS (SLs).............................................................................................. 2.0-1 2.1 Safety Limits 2.2 SL Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.......................... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY......................................... 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM).................................................................... 3.1.1-1 3.1.2 Reactivity Anomalies................................................................................... 3.1.2-1 3.1.3 Control Rod OPERABILITY........................................................................ 3.1.3-1 3.1.4 Control Rod Scram Times........................................................................... 3.1.4-1 3.1.5 Control Rod Scram Accumulators............................................................... 3.1.5-1 3.1.6 Rod Pattern Control.................................................................................... 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System......................................................... 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves............................. 3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)............................................................................................. 3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)........................................... 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optional)............................ 3.2.3-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation................................... 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation......................................... 3.3.1.2-1 3.3.2.1 Control Rod Block Instrumentation........................................................... 3.3.2.1-1 3.3.2.2 Feedwater Pump and Main Turbine High Water Level Trip Instrumentation.................................................................................. 3.3.2.2-1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation..................................... 3.3.3.1-1 3.3.3.2 Alternate Shutdown System..................................................................... 3.3.3.2-1 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation............................................................. 3.3.4.1-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation...................... 3.3.5.1-1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation.......... 3.3.5.2-1 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation3.3.5.3-1 3.3.6.1 Primary Containment Isolation Instrumentation........................................ 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation................................... 3.3.6.2-1 3.3.6.3 Low-Low Set (LLS) Instrumentation......................................................... 3.3.6.3-1 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrumentation.................................................................................. 3.3.7.1-1 3.3.7.2 Mechanical Vacuum Pump Isolation Instrumentation............................... 3.3.7.2-1 3.3.8.1 Loss of Power (LOP) Instrumentation...................................................... 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring..3.3.8.2-1

Monticello ii Revision 1 TABLE OF CONTENTS Page Number 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating.................................................................... 3.4.1-1 3.4.2 Jet Pumps................................................................................................... 3.4.2-1 3.4.3 Safety/Relief Valves (S/RVs)...................................................................... 3.4.3-1 3.4.4 RCS Operational LEAKAGE....................................................................... 3.4.4-1 3.4.5 RCS Leakage Detection Instrumentation.................................................... 3.4.5-1 3.4.6 RCS Specific Activity.................................................................................. 3.4.6-1 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown.............................................................................................. 3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown.............................................................................................. 3.4.8-1 3.4.9 RCS Pressure and Temperature (P/T) Limits.............................................. 3.4.9-1 3.4.10 Reactor Steam Dome Pressure................................................................ 3.4.10-1 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating....................................................................................... 3.5.1-1 3.5.2 RPV Water Inventory Control...................................................................... 3.5.2-1 3.5.3 RCIC System.............................................................................................. 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment............................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Lock.................................................................. 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs)........................................ 3.6.1.3-1 3.6.1.4 Drywell Air Temperature.......................................................................... 3.6.1.4-1 3.6.1.5 Low-Low Set (LLS) Valves....................................................................... 3.6.1.5-1 3.6.1.6 Reactor Building-to-Suppression Chamber Vacuum Breakers................. 3.6.1.6-1 3.6.1.7 Suppression Chamber-to-Drywell Vacuum Breakers............................... 3.6.1.7-1 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray.......................................... 3.6.1.8-1 3.6.2.1 Suppression Pool Average Temperature................................................. 3.6.2.1-1 3.6.2.2 Suppression Pool Water Level................................................................. 3.6.2.2-1 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling....................... 3.6.2.3-1 3.6.3.1 Primary Containment Oxygen Concentration........................................... 3.6.3.1-1 3.6.4.1 Secondary Containment........................................................................... 3.6.4.1-1 3.6.4.2 Secondary Containment Isolation Valves (SCIVs).................................... 3.6.4.2-1 3.6.4.3 Standby Gas Treatment (SGT) System.................................................... 3.6.4.3-1 3.7 PLANT SYSTEMS 3.7.1 Residual Heat Removal Service Water (RHRSW) System.......................... 3.7.1-1 3.7.2 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS)............................................................................................ 3.7.2-1 3.7.3 Emergency Diesel Generator Emergency Service Water (EDG-ESW) System.............................................................................. 3.7.3-1 3.7.4 Control Room Emergency Filtration (CREF) System................................... 3.7.4-1 3.7.5 Control Room Ventilation System............................................................... 3.7.5-1 3.7.6 Main Condenser Offgas.............................................................................. 3.7.6-1 3.7.7 Main Turbine Bypass System..................................................................... 3.7.7-1 3.7.8 Spent Fuel Storage Pool Water Level......................................................... 3.7.8-1

Definitions 1.1 Monticello 1.1-2 Amendment No. 146, 148, XXX 1.1 Definitions CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

The following exceptions are not considered to be CORE ALTERATIONS:

a.

Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and

b.

Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.3. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR)-11, "Limiting Values of Radionuclide Intake and Air Concentration Factors for Inhalation, Submersion and Ingestion," September 1988, and FGR-12, "External Exposure to Radionuclides in Air, Water and Soil,"

September 1993.

DRAIN TIME The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming:

a)

The water inventory above the TAF is divided by the limiting drain rate; b)

The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single

Definitions 1.1 Monticello 1.1-3 Amendment No. 194, XXX 1.1 Definitions DRAIN TIME (continued) human error), for all penetration flow paths below the TAF except:

1.

Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;

2.

Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation; or

3.

Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.

c)

The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory; d)

No additional draining events occur; and e)

Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-1 Amendment No. 146, XXX 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.5.1-1.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels inoperable.

A.1 Enter the Condition referenced in Table 3.3.5.1-1 for the channel.

Immediately B. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

B.1

NOTE-------------

1. Only applicable for Functions 1.a, 1.b, 2.a, 2.b, 2.f, 2.h, and 2.k.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

AND 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divisions

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.2

NOTE--------------

Only applicable for Functions 3.a and 3.b.

Declare High Pressure Coolant Injection (HPCI)

System inoperable.

AND B.3 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

C.1

NOTE-------------

1. Only applicable for Functions 1.c, 1.d, 1.e, 1.f, 2.c, 2.d, 2.e, 2.i, 2.j, 2.l, and 2.m.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

AND C.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divisions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-3 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

D.1

NOTE--------------

Only applicable if HPCI pump suction is not aligned to the suppression pool.

Declare HPCI System inoperable.

AND D.2.1 Place channel in trip.

OR D.2.2 Align the HPCI pump suction to the suppression pool.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours E. As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

E.1

NOTE-------------

1. Only applicable for Function 2.g.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

AND E.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for subsystems in both divisions 7 days

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-6 Amendment No. 146, 151, 170, 176, XXX Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

a.

Reactor Vessel Water Level - Low Low 1, 2, 3 4(a)

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Drywell Pressure -

High 1, 2, 3 4(a)

B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.8 2 psig

c.

Reactor Steam Dome Pressure -

Low (Injection Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 397 psig and 440 psig

d.

Reactor Steam Dome Pressure Permissive - Low (Pump Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 397 psig

e.

Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive) 1, 2, 3 2

C SR 3.3.5.1.7 SR 3.3.5.1.8 18 minutes (a) Also required to initiate the associated emergency diesel generator (EDG).

(b)

If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c)

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the Technical Requirements Manual (TRM).

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-7 Amendment No. 146, XXX Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

f.

Core Spray Pump Start - Time Delay Relay 1, 2, 3 1 per pump C

SR 3.3.5.1.7 SR 3.3.5.1.8 15.86 seconds

2. Low Pressure Coolant Injection (LPCI) System

a.

Reactor Vessel Water Level - Low Low 1, 2, 3, 4

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Drywell Pressure -

High 1, 2, 3 4

B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.8 2 psig

c.

Reactor Steam Dome Pressure -

Low (Injection Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 397 psig and 440 psig

d.

Reactor Steam Dome Pressure Permissive - Low (Pump Permissive) 1, 2, 3 2

C SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 397 psig (b)

If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c)

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-8 Correction letter of 10/12/06 Amendment No. 146, 149, 151, 161, 170 176, XXX Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

2. LPCI System

e.

Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive) 1, 2, 3 2

C SR 3.3.5.1.7 SR 3.3.5.1.8 18 minutes

f.

Low Pressure Coolant Injection Pump Start - Time Delay Relay 1, 2, 3, 4 per pump B

SR 3.3.5.1.7 SR 3.3.5.1.8 Pumps A, B 5.33 seconds Pumps C, D 10.59 seconds

g.

Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 1, 2, 3, 1 per pump E

SR 3.3.5.1.2 SR 3.3.5.1.7 SR 3.3.5.1.8 360 gpm and 745 gpm

h.

Reactor Steam Dome Pressure -

Low (Break Detection) 1, 2, 3 4

B SR 3.3.5.1.2 SR 3.3.5.1.7 SR 3.3.5.1.8 873.6 psig and 923.4 psig

i.

Recirculation Pump Differential Pressure

- High (Break Detection) 1, 2, 3 4 per pump C

SR 3.3.5.1.2 SR 3.3.5.1.7 SR 3.3.5.1.8 63.5 inches wc

j.

Recirculation Riser Differential Pressure

- High (Break Detection) 1, 2, 3 4

C SR 3.3.5.1.2 SR 3.3.5.1.7(b)(c)

SR 3.3.5.1.8 100.0 inches wc (b)

If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c)

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-9 Amendment No. 146, 149, 151, 161, XXX Table 3.3.5.1-1 (page 4 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

2. LPCI System

k.

Recirculation Steam Dome Pressure -

Time Delay Relay (Break Detection) 1, 2, 3 2

B SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.9 2.97 seconds

l.

Recirculation Pump Differential Pressure

- Time Delay Relay (Break Detection) 1, 2, 3 2

C SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.9 0.75 seconds

m. Recirculation Riser Differential Pressure

- Time Delay Relay (Break Detection) 1, 2, 3 2

C SR 3.3.5.1.7 SR 3.3.5.1.8 SR 3.3.5.1.9 0.75 seconds

3. High Pressure Coolant Injection (HPCI) System

a.

Reactor Vessel Water Level - Low Low 1, 2(d), 3(d) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Drywell Pressure -

High 1, 2(d), 3(d) 4 B

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.8 2 psig

c.

Reactor Vessel Water Level - High 1, 2(d), 3(d) 2 C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8 48 inches

d.

Condensate Storage Tank Level - Low 1, 2(d), 3(d) 2 D

SR 3.3.5.1.7 SR 3.3.5.1.8 29.3 inches

e.

Suppression Pool Water Level - High 1, 2(d), 3(d) 2 D

SR 3.3.5.1.5 SR 3.3.5.1.6 SR 3.3.5.1.8 3.0 inches

f.

High Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 1, 2(d), 3(d) 1 E

SR 3.3.5.1.5 SR 3.3.5.1.6 SR 3.3.5.1.8 362 gpm and 849 gpm (d)

With reactor steam dome pressure > 150 psig.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-10 Amendment No. 146, XXX Table 3.3.5.1-1 (page 5 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

4. Automatic Depressurization System (ADS) Trip System A

a.

Reactor Vessel Water Level - Low Low 1, 2(d), 3(d) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Automatic Depressurization System Initiation Timer 1, 2(d), 3(d) 1 G

SR 3.3.5.1.7 SR 3.3.5.1.8 120 seconds

c.

Core Spray Pump Discharge Pressure

- High 1, 2(d), 3(d) 2 G

SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 75 psig and 125 psig

d.

Low Pressure Coolant Injection Pump Discharge Pressure - High 1, 2(d), 3(d) 4 G

SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 75 psig and 125 psig

5. ADS Trip System B

a.

Reactor Vessel Water Level - Low Low 1, 2(d), 3(d) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.7 SR 3.3.5.1.8

-48 inches

b.

Automatic Depressurization System Initiation Timer 1, 2(d), 3(d) 1 G

SR 3.3.5.1.7 SR 3.3.5.1.8 120 seconds (b)

If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c)

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

(d)

With reactor steam dome pressure > 150 psig.

ECCS Instrumentation 3.3.5.1 Monticello 3.3.5.1-11 Amendment No. 146, XXX Table 3.3.5.1-1 (page 6 of 6)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5. ADS Trip System B

c.

Core Spray Pump Discharge Pressure

- High 1, 2(d), 3(d) 2 G

SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 75 psig and 125 psig

d.

Low Pressure Coolant Injection Pump Discharge Pressure - High 1, 2(d), 3(d) 4 G

SR 3.3.5.1.2 SR 3.3.5.1.4(b)(c)

SR 3.3.5.1.8 75 psig and 125 psig (b)

If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c)

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the nominal trip setpoint; otherwise, the channel shall be declared inoperable. The nominal trip setpoint and the methodology used to determine the as-found tolerance and the as-left tolerance are specified in the TRM.

(d)

With reactor steam dome pressure > 150 psig.

RPV Water Inventory Control Instrumentation 3.3.5.3 Monticello 3.3.5.3-1 Amendment No. XXX 3.3 INSTRUMENTATION 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.3 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.5.3-1.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels inoperable.

A.1 Enter the Condition referenced in Table 3.3.5.3-1 for the channel.

Immediately B. As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

B.1 Declare associated penetration flow path(s) incapable of automatic isolation.

AND B.2 Calculate DRAIN TIME.

Immediately Immediately C. As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

C.1 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

D.1 Restore channel to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

RPV Water Inventory Control Instrumentation 3.3.5.3 Monticello 3.3.5.3-2 Amendment No. XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and associated Completion Time of Condition C or D not met.

E.1 Declare associated low pressure ECCS injection/spray subsystem inoperable.

Immediately SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

Refer to Table 3.3.5.3-1 to determine which SRs apply for each ECCS Function.

SURVEILLANCE FREQUENCY SR 3.3.5.3.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.5.3.2 Perform CHANNEL FUNCTIONAL TEST.

92 days

RPV Water Inventory Control Instrumentation 3.3.5.3 Monticello 3.3.5.3-3 Amendment No. XXX Table 3.3.5.3-1 (page 1 of 1)

RPV Water Inventory Control Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Core Spray System

a.

Reactor Steam Dome Pressure - Low (Injection Permissive) 4, 5 2

C SR 3.3.5.3.2 397 psig and 440psig

2. Low Pressure Coolant Injection (LPCI)

System

a.

Reactor Steam Dome Pressure -

Low (Injection Permissive) 4, 5 2

C SR 3.3.5.3.2 397 psig and 440 psig

b.

Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 4, 5 1 per pump(a)

D SR 3.3.5.3.2 360 gpm and 745 gpm

3. RHR System Isolation

a.

Reactor Vessel Water Level - Low (b) 2 in one trip system B

SR 3.3.5.3.1 SR 3.3.5.3.2 7 inches

4. Reactor Water Cleanup (RWCU) System Isolation

a.

Reactor Vessel Water Level - Low Low (b) 2 in one trip system B

SR 3.3.5.3.1 SR 3.3.5.3.2

-48 inches (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "RPV Water Inventory Control."

(b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.

Primary Containment Isolation Instrumentation 3.3.6.1 Monticello 3.3.6.1-3 Amendment No. XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I.

As required by Required Action C.1 and referenced in Table 3.3.6.1-1.

I.1 Initiate action to restore channel to OPERABLE status.

Immediately SURVEILLANCE REQUIREMENTS

NOTES----------------------------------------------------------

1.

Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2.

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains primary containment isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.6.1.3 Calibrate the trip unit.

92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION.

92 days

Primary Containment Isolation Instrumentation 3.3.6.1 Monticello 3.3.6.1-7 Amendment No. 146, 164, XXX Table 3.3.6.1-1 (page 3 of 3)

Primary Containment Isolation Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION C.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5. RWCU System Isolation

d.

SLC System Initiation 1, 2, 3 1

H SR 3.3.6.1.6 NA

e.

Reactor Vessel Water Level - Low Low 1, 2, 3 2

F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6

-48 inches

6. Shutdown Cooling System Isolation

a.

Reactor Steam Dome Pressure -

High 1, 2, 3 2

F SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 81.8 psig

b.

Reactor Vessel Water Level - Low 3

2 I

SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 7 inches

7. Traversing Incore Probe System Isolation

a.

Reactor Vessel Water Level - Low 1, 2, 3 2

G SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 7 inches

b.

Drywell Pressure -

High 1, 2, 3 2

G SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 2.0 psig

Secondary Containment Isolation Instrumentation 3.3.6.2 Monticello 3.3.6.2-3 Amendment No. 146, XXX Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1. Reactor Vessel Water Level - Low Low 1, 2, 3 2

SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

-48 inches

2. Drywell Pressure - High 1, 2, 3 2

SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 2 psig

3. Reactor Building Ventilation Exhaust Radiation - High 1, 2, 3, (a) 2 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 100 mR/hr

4. Refueling Floor Radiation - High 1, 2, 3, (a) 2 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 100 mR/hr (a)

During movement of recently irradiated fuel assemblies in secondary containment.

CREF System Instrumentation 3.3.7.1 Monticello 3.3.7.1-3 Amendment No. 146, 148, XXX Table 3.3.7.1-1 (Page 1 of 1)

Control Room Emergency Filtration System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1, 2, 3 2

SR 3.3.7.1.1

- 48 inches Level - Low Low SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6

2. Drywell Pressure -

1, 2, 3 2

SR 3.3.7.1.2 2 psig High SR 3.3.7.1.4 SR 3.3.7.1.6

3. Reactor Building 1, 2, 3, (a) 2 SR 3.3.7.1.1 100 mR/hr Ventilation Exhaust SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.4 SR 3.3.7.1.6

4. Refueling Floor 1, 2, 3, (a) 2 SR 3.3.7.1.1 100 mR/hr Radiation - High SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.6 (a)

During movement of recently irradiated fuel assemblies in the secondary containment.

RPS Electric Power Monitoring 3.3.8.2 Monticello 3.3.8.2-1 Amendment No. 146, XXX 3.3 INSTRUMENTATION 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring LCO 3.3.8.2 Two RPS electric power monitoring assemblies shall be OPERABLE for each inservice RPS motor generator set or alternate power supply.

APPLICABILITY:

MODES 1, 2, and 3, MODES 4 and 5 with residual heat removal (RHR) shutdown cooling supply isolation valves open, MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both inservice power supplies with one electric power monitoring assembly inoperable.

A.1 Remove associated inservice power supply(s) from service.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. One or both inservice power supplies with both electric power monitoring assemblies inoperable.

B.1 Remove associated inservice power supply(s) from service.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

RPS Electric Power Monitoring 3.3.8.2 Monticello 3.3.8.2-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition A or B not met in MODE 4 or 5 with RHR shutdown cooling supply isolation valves open.

D.1 Initiate action to restore one electric power monitoring assembly to OPERABLE status for inservice power supply(s) supplying required instrumentation.

OR D.2 Initiate action to isolate the RHR Shutdown Cooling System.

Immediately Immediately E. Required Action and associated Completion Time of Condition A or B not met in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies.

E.1 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Immediately F. Required Action and associated Completion Time of Condition A or B not met during movement of recently irradiated fuel assemblies in the secondary containment.

F.1.1 Isolate the associated secondary containment penetration flow path(s).

OR F.1.2 Declare the associated secondary containment isolation valve(s) inoperable.

AND F.2.1 Place the associated standby gas treatment (SGT) subsystem(s) in operation.

OR Immediately Immediately Immediately

ECCS - Operating 3.5.1 Monticello 3.5.1-1 Amendment No. 146, XXX 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM (RCIC) 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of three safety/relief valves shall be OPERABLE.

NOTE--------------------------------------------

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling supply isolation interlock in MODE 3, if capable of being manually realigned and not otherwise inoperable.

APPLICABILITY:

MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One LPCI pump inoperable.

A.1 Restore LPCI pump to OPERABLE status.

30 days B. One LPCI subsystem inoperable for reasons other than Condition A.

OR One Core Spray subsystem inoperable.

B.1 Restore low pressure ECCS injection/spray subsystem to OPERABLE status.

7 days

RPV Water Inventory Control 3.5.2 Monticello 3.5.2-1 Amendment No. 146, XXX 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) SYSTEM 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control LCO 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AND One low pressure ECCS injection/spray subsystem shall be OPERABLE.

NOTE--------------------------------------------

A Low Pressure Coolant Injection (LPCI) subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

APPLICABILITY:

MODES 4 and 5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required ECCS injection/spray subsystem inoperable.

A.1 Restore required ECCS injection/spray subsystem to OPERABLE status.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion Time of Condition A not met.

B.1 Initiate action to establish a method of water injection capable of operating without offsite electrical power.

Immediately C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.1 Verify secondary containment boundary is capable of being established in less than the DRAIN TIME.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

RPV Water Inventory Control 3.5.2 Monticello 3.5.2-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.2 Verify each secondary containment penetration flow path is capable of being isolated in less than the DRAIN TIME.

AND C.3 Verify one standby gas treatment subsystem is capable of being placed in operation in less than the DRAIN TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours D. DRAIN TIME < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

D.1

NOTE ---------------

Required ECCS injection/spray subsystem or additional method of water injection shall be capable of operating without offsite electrical power.

Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AND D.2 Initiate action to restore secondary containment boundary.

AND Immediately Immediately

RPV Water Inventory Control 3.5.2 Monticello 3.5.2-3 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.3 Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room.

AND D.4 Initiate action to verify one standby gas treatment subsystem is capable of being placed in operation.

Immediately Immediately E. Required Action and associated Completion Time of Condition C or D not met.

OR DRAIN TIME < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

E.1 Initiate action to restore DRAIN TIME to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for the required ECCS injection/spray subsystem, the:

a. Suppression pool water level is -3 ft; or

b. Condensate storage tank(s) water level is 7 ft for one tank operation and 4 ft for two tank operation.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

RPV Water Inventory Control 3.5.2 Monticello 3.5.2-4 Amendment No. 146, 167, 189, 194, XXX SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

31 days SR 3.5.2.3

NOTE------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify, for the required ECCS injection/spray subsystem, each manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

31 days SR 3.5.2.4 Operate the required ECCS injection/spray subsystem for 10 minutes.

92 days SR 3.5.2.5

NOTE------------------------------

Vessel injection/spray may be excluded.

Verify the required ECCS injection/spray subsystem can be manually operated.

24 months SR 3.5.2.6 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal.

24 months SR 3.5.2.7 Verify DRAIN TIME 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

RCIC System 3.5.3 Monticello 3.5.3-1 Amendment No. 146, XXX 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY:

MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to the RCIC System.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable.

A.1 Verify by administrative means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System to OPERABLE status.

Immediately 14 days B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Reduce reactor steam dome pressure to 150 psig.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

PCIVs 3.6.1.3 Monticello 3.6.1.3-6 Amendment No. 146, 148, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and associated Completion Time of Condition A or B not met for PCIV(s) required to be OPERABLE during MODE 4 or 5.

G.1 Initiate action to restore valve(s) to OPERABLE status.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1

NOTE------------------------------

Not required to be met when the 18 inch primary containment purge and vent valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

Verify each 18 inch primary containment purge and vent valve is closed.

31 days

Secondary Containment 3.6.4.1 Monticello 3.6.4.1-1 Amendment No. 146, XXX 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment inoperable in MODE 1, 2, or 3.

A.1 Restore secondary containment to OPERABLE status.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Required Action and associated Completion Time of Condition A not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours C. Secondary containment inoperable during movement of recently irradiated fuel assemblies in the secondary containment.

C.1

NOTE--------------

LCO 3.0.3 is not applicable.

Suspend movement of recently irradiated fuel assemblies in the secondary containment.

Immediately

SCIVs 3.6.4.2 Monticello 3.6.4.2-1 Amendment No. 146, XXX 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS

NOTES----------------------------------------------------------

1.

Penetration flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration flow paths with one SCIV inoperable.

A.1 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

SCIVs 3.6.4.2 Monticello 3.6.4.2-3 Amendment No.146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition A or B not met during movement of recently irradiated fuel assemblies in the secondary containment.

D.1

NOTE--------------

LCO 3.0.3 is not applicable.

Suspend movement of recently irradiated fuel assemblies in the secondary containment.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1

NOTES-----------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment isolation manual valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

31 days SR 3.6.4.2.2 Verify the isolation time of each power operated, automatic SCIV is within limits.

92 days SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation position on an actual or simulated actuation signal.

24 months

SGT System 3.6.4.3 Monticello 3.6.4.3-1 Amendment No. 146, XXX 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem inoperable.

A.1 Restore SGT subsystem to OPERABLE status.

7 days B. Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours C. Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment.

NOTE-------------------

LCO 3.0.3 is not applicable.

C.1 Place OPERABLE SGT subsystem in operation.

OR C.2.

Suspend movement of recently irradiated fuel assemblies in secondary containment.

Immediately Immediately

SGT System 3.6.4.3 Monticello 3.6.4.3-2 Amendment No. 146, 181, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two SGT subsystems inoperable in MODE 1, 2, or 3.

D.1 Enter LCO 3.0.3.

Immediately E. Two SGT subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment.

E.1

NOTE--------------

LCO 3.0.3 is not applicable.

Suspend movement of recently irradiated fuel assemblies in secondary containment.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for 15 continuous minutes.

31 days SR 3.6.4.3.2 Perform required SGT filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or simulated initiation signal.

24 months

CREF System 3.7.4 Monticello 3.7.4-1 Amendment No. 146, 160, 175, XXX 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Filtration (CREF) System LCO 3.7.4 Two CREF subsystems shall be OPERABLE.

NOTE--------------------------------------------

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREF subsystem inoperable for reasons other than Condition B.

A.1 Restore CREF subsystem to OPERABLE status.

7 days B. One or more CREF subsystems inoperable due to inoperable CRE boundary in MODE 1, 2, or 3.

B.1 Initiate action to implement mitigating actions.

AND B.2 Verify mitigating actions ensure CRE occupant exposures to radiological, chemical and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to OPERABLE status.

Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 90 days

CREF System 3.7.4 Monticello 3.7.4-2 Amendment No. 146, 160, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours D. Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment.

NOTE-------------------

LCO 3.0.3 is not applicable.

D.1 Place OPERABLE CREF subsystem in pressurization mode.

OR D.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

Immediately Immediately E. Two CREF subsystems inoperable in MODE 1, 2, or 3 for reasons other than Condition B.

E.1 Enter LCO 3.0.3.

Immediately

CREF System 3.7.4 Monticello 3.7.4-3 Amendment No. 146, 160, 175, 181, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREF subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment.

OR One or more CREF subsystems inoperable due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment.

NOTE-------------------

LCO 3.0.3 is not applicable.

F.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate each CREF subsystem for 15 continuous minutes.

31 days SR 3.7.4.2 Perform required CREF filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.4.3 Verify each CREF subsystem actuates on an actual or simulated initiation signal.

24 months SR 3.7.4.4 Perform required CRE unfiltered air in-leakage testing in accordance with the Control Room Envelope Habitability Program.

In accordance with the Control Room Envelope Habitability Program

Control Room Ventilation System 3.7.5 Monticello 3.7.5-1 Amendment No. 146, 148, 154, XXX 3.7 PLANT SYSTEMS 3.7.5 Control Room Ventilation System LCO 3.7.5 Two control room ventilation subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One control room ventilation subsystem inoperable.

A.1 Restore control room ventilation subsystem to OPERABLE status.

30 days B. Two control room ventilation subsystems inoperable.

B.1 Verify control room area temperature < 90°F.

AND B.2 Restore one control room ventilation subsystem to OPERABLE status.

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 72 hours C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours D. Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies in the secondary containment.

NOTE-------------------

LCO 3.0.3 is not applicable.

D.1 Place OPERABLE control room ventilation subsystem in operation.

OR Immediately

Control Room Ventilation System 3.7.5 Monticello 3.7.5-2 Amendment No. 146, 148, 154, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.2 Suspend movement of irradiated fuel assemblies in the secondary containment.

Immediately E. Required Action and associated Completion Time of Condition B not met during movement of recently irradiated fuel assemblies in the secondary containment.

NOTE-------------------

LCO 3.0.3 is not applicable.

E.1 Suspend movement of irradiated fuel assemblies in the secondary containment.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each control room ventilation subsystem has the capability to remove the assumed heat load.

24 months

AC Sources - Shutdown 3.8.2 Monticello 3.8.2-2 Amendment No. 146, 148, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND A.2.3 Initiate action to restore required offsite power circuit to OPERABLE status.

Immediately Immediately Immediately B. One required EDG inoperable.

B.1 Suspend CORE ALTERATIONS.

AND B.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND B.3 Initiate action to restore required EDG to OPERABLE status.

Immediately Immediately Immediately

AC Sources - Shutdown 3.8.2 Monticello 3.8.2-3 Amendment No. 146, XXX SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1

NOTES-----------------------------

1. The following SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.7, and SR 3.8.1.8 through SR 3.8.1.13.

2. SR 3.8.1.8 and SR 3.8.1.12 are not required to be met when associated ECCS subsystem(s) are not required to be OPERABLE per LCO 3.5.2, "RPV Water Inventory Control."

For AC sources required to be OPERABLE the SRs of Specification 3.8.1, except SR 3.8.1.6, are applicable.

In accordance with applicable SRs

DC Sources - Shutdown 3.8.5 Monticello 3.8.5-1 Amendment No. 146, 148, XXX 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 Division 1 or Division 2 125 VDC electrical power subsystem shall be OPERABLE to support one division of the DC Electrical Power Distribution System required by LCO 3.8.8, "Distribution Systems -

Shutdown."

APPLICABILITY:

MODES 4 and 5, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Required DC electrical power subsystem inoperable.

A.1 Suspend CORE ALTERATIONS.

AND A.2 Suspend movement of recently irradiated fuel assemblies in the secondary containment.

AND A.3 Initiate action to restore required DC electrical power subsystem to OPERABLE status.

Immediately Immediately Immediately

Distribution Systems - Shutdown 3.8.8 Monticello 3.8.8-1 Amendment No. 146, 148, XXX 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems - Shutdown LCO 3.8.8 The necessary portions of the AC and DC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY:

MODES 4 and 5, During movement of recently irradiated fuel assemblies in the secondary containment.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required AC or DC electrical power distribution subsystems inoperable.

A.1 Declare associated supported required feature(s) inoperable.

OR A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend handling of recently irradiated fuel assemblies in the secondary containment.

AND Immediately Immediately Immediately

Distribution Systems - Shutdown 3.8.8 Monticello 3.8.8-2 Amendment No. 146, XXX ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2.3 Initiate actions to restore required AC and DC electrical power distribution subsystems to OPERABLE status.

AND A.2.4 Declare associated required shutdown cooling subsystem(s) inoperable and not in operation.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and voltage to required AC and DC electrical power distribution subsystems.

7 days

ATTACHMENT 4 MONTICELLO NUCLEAR GENERATING PLANT Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (Mark-Ups) 79 pages follow

Monticello i

Revision No. 45 TABLE OF CONTENTS Page Number B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs......................................................................................... B 2.1.1-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL................................................ B 2.1.2-1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY...................... B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY................................... B 3.0-13 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)................................................................ B 3.1.1-1 B 3.1.2 Reactivity Anomalies............................................................................... B 3.1.2-1 B 3.1.3 Control Rod OPERABILITY.................................................................... B 3.1.3-1 B 3.1.4 Control Rod Scram Times....................................................................... B 3.1.4-1 B 3.1.5 Control Rod Scram Accumulators........................................................... B 3.1.5-1 B 3.1.6 Rod Pattern Control................................................................................ B 3.1.6-1 B 3.1.7 Standby Liquid Control (SLC) System..................................................... B 3.1.7-1 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves......................... B 3.1.8-1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)......................................................................................... B 3.2.1-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)....................................... B 3.2.2-1 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optional)........................ B 3.2.3-1 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation............................... B 3.3.1.1-1 B 3.3.1.2 Source Range Monitor (SRM) Instrumentation..................................... B 3.3.1.2-1 B 3.3.2.1 Control Rod Block Instrumentation....................................................... B 3.3.2.1-1 B 3.3.2.2 Feedwater Pump and Main Turbine High Water Level Trip Instrumentation.............................................................................. B 3.3.2.2-1 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation................................. B 3.3.3.1-1 B 3.3.3.2 Alternate Shutdown System................................................................. B 3.3.3.2-1 B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation......................................................... B 3.3.4.1-1 B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation.................. B 3.3.5.1-1 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation.......... B 3.3.5.2-1 B 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation.................................................................................... B 3.3.5.2-1 B 3.3.6.1 Primary Containment Isolation Instrumentation.................................... B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation............................... B 3.3.6.2-1 B 3.3.6.3 Low-Low Set (LLS) Instrumentation..................................................... B 3.3.6.3-1 B 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrumentation.............................................................................. B 3.3.7.1-1 B 3.3.7.2 Mechanical Pump Isolation Instrumentation......................................... B 3.3.7.2-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation.................................................. B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring............... B 3.3.8.2-1

Monticello ii Revision No. 45 TABLE OF CONTENTS Page Number B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating................................................................ B 3.4.1-1 B 3.4.2 Jet Pumps............................................................................................... B 3.4.2-1 B 3.4.3 Safety/Relief Valves (S/RVs).................................................................. B 3.4.3-1 B 3.4.4 RCS Operational LEAKAGE................................................................... B 3.4.4-1 B 3.4.5 RCS Leakage Detection Instrumentation................................................ B 3.4.5-1 B 3.4.6 RCS Specific Activity.............................................................................. B 3.4.6-1 B 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown.......................................................................................... B 3.4.7-1 B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown.......................................................................................... B 3.4.8-1 B 3.4.9 RCS Pressure and Temperature (P/T) Limits.......................................... B 3.4.9-1 B 3.4.10 Reactor Steam Dome Pressure....................................................... B 3.4.10-1 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating................................................................................... B 3.5.1-1 B 3.5.2 RPV Water Inventory ControlECCS - Shutdown..................................... B 3.5.2-1 B 3.5.3 RCIC System.......................................................................................... B 3.5.3-1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment........................................................................... B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Lock.............................................................. B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs).................................... B 3.6.1.3-1 B 3.6.1.4 Drywell Air Temperature...................................................................... B 3.6.1.4-1 B 3.6.1.5 Low-Low Set (LLS) Valves................................................................... B 3.6.1.5-1 B 3.6.1.6 Reactor Building-to-Suppression Chamber Vacuum Breakers............. B 3.6.1.6-1 B 3.6.1.7 Suppression Chamber-to-Drywell Vacuum Breakers........................... B 3.6.1.7-1 B 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray...................................... B 3.6.1.8-1 B 3.6.2.1 Suppression Pool Average Temperature............................................. B 3.6.2.1-1 B 3.6.2.2 Suppression Pool Water Level............................................................. B 3.6.2.2-1 B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling................... B 3.6.2.3-1 B 3.6.3.1 Primary Containment Oxygen Concentration...................................... B 3.6.3.1-1 B 3.6.4.1 Secondary Containment....................................................................... B 3.6.4.1-1 B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)................................ B 3.6.4.2-1 B 3.6.4.3 Standby Gas Treatment (SGT) System................................................ B 3.6.4.3-1 B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Water (RHRSW) System...................... B 3.7.1-1 B 3.7.2 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS)........................................................................................ B 3.7.2-1 B 3.7.3 Emergency Diesel Generator Emergency Service Water (EDG-ESW) System.......................................................................... B 3.7.3-1 B 3.7.4 Control Room Emergency Filtration (CREF) System............................... B 3.7.4-1 B 3.7.5 Control Room Ventilation System........................................................... B 3.7.5-1 B 3.7.6 Main Condenser Offgas.......................................................................... B 3.7.6-1 B 3.7.7 Main Turbine Bypass System................................................................. B 3.7.7-1 B 3.7.8 Spent Fuel Storage Pool Water Level..................................................... B 3.7.8-1

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-8 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Table 3.3.5.1-1 is modified by atwo footnotes which. Footnote (a) is added to clarify that the associated functions are required to be OPERABLE in MODES 4 and 5 only when their supported ECCS are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown.

Footnote (b) is added to show that certain ECCS instrumentation Functions also perform EDG initiation.

Allowable Values are specified for each ECCS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations.

The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between 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 if the setpoint is found conservative to the Allowable Value during surveillance testing, the instrumentation would

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-9 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) have provided the required trip function by the time the process reached the analytic limit for the applicable events.

In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or EDG) initiation to mitigate the consequences of a design basis transient or accident. To ensure reliable ECCS and EDG function, a combination of Functions is required to provide primary and secondary initiation signals.

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

Core Spray and Low Pressure Coolant Injection Systems 1.a, 2.a. Reactor Vessel Water Level - Low Low Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. The low pressure ECCS and associated EDGs are initiated at Low Low to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. The Reactor Vessel Water Level - Low Low is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in Reference 2. In addition, the Reactor Vessel Water Level - Low Low Function is directly assumed in the analysis of the recirculation line break (Ref. 1). The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS),

ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low 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.

The Reactor Vessel Water Level - Low Low Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling.

Four channels of CS Reactor Vessel Water Level - Low Low Function are only required to be OPERABLE when CS and the EDGs are required to be OPERABLE to ensure that no single instrument failure can preclude CS and EDG initiation. Four channels of the LPCI Reactor Vessel Water Level - Low Low Function are only required to be OPERABLE when LPCI is required to be OPERABLE to ensure that no single instrument failure can preclude LPCI initiation. Per Footnote (a) to Table 3.3.5.1-1, these

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-10 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

ECCS Functions are only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems; LCO 3.8.1, "AC Sources -

Operating," and LCO 3.8.2, "AC Sources - Shutdown," for Applicability Bases for the EDGs.

1.b, 2.b. Drywell Pressure - High High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated EDGs are initiated upon receipt of the Drywell Pressure - High Function in order to minimize the possibility of fuel damage. The Drywell Pressure -

High Function, along with the Reactor Water Level - Low Low Function, is directly assumed in the analysis of the recirculation line break (Ref. 1).

The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

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 be indicative of a LOCA inside primary containment.

The Drywell Pressure - High Function is required to be OPERABLE when the ECCS or EDG is required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure - High Functions are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude ECCS and EDG initiation. In MODES 4 and 5, the Drywell Pressure - High Functions are not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure - High setpoint. Refer to LCO 3.5.1 for Applicability Bases for the low pressure ECCS subsystems and to LCO 3.8.1 for Applicability Bases for the EDGs.

1.c, 2.c. Reactor Steam Dome Pressure - Low (Injection Permissive)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. The Reactor Steam Dome Pressure - Low (Injection Permissive) is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in Reference 2. In addition, the Reactor Steam Dome Pressure - Low (Injection Permissive) Function is directly assumed in the analysis of the

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-11 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) recirculation line break (Ref. 1). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

The Reactor Steam Dome Pressure - Low (Injection Permissive) signals are initiated from two pressure switches (shared by both CS and LPCI) that sense the reactor dome pressure.

The Allowable Value is low enough to prevent overpressurizing the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.

Two channels of CS Reactor Steam Dome Pressure - Low (Injection Permissive) Function are only required to be OPERABLE when CS is required to be OPERABLE to ensure that no single instrument failure can preclude CS initiation. Two channels of the LPCI Reactor Steam Dome Pressure - Low (Injection Permissive) Function are only required to be OPERABLE when LPCI is required to be OPERABLE to ensure that no single instrument failure can preclude LPCI initiation. Per Footnote (a) to Table 3.3.5.1-1, these ECCS Functions are only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

1.d, 2.d. Reactor Steam Dome Pressure - Low (Pump Permissive)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. These channels delay CS and LPCI pump starts on Reactor Vessel Water Level - Low Low until reactor steam dome pressure is below the setpoint. This ensures that, prior to starting the pumps of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure.

The Reactor Steam Dome Pressure - Low (Pump Permissive) is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. In addition, the Reactor Steam Dome Pressure - Low Function is directly assumed in the analysis of the recirculation line break (Ref. 2).

The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

The Reactor Steam Dome Pressure - Low signals are initiated from two pressure switches (shared by both CS and LPCI) that sense the reactor dome pressure.

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-12 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Allowable Value is high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.

Two channels of CS Reactor Steam Dome Pressure - Low (Pump Permissive) Function are only required to be OPERABLE when the CS is required to be OPERABLE to ensure that no single instrument failure can preclude CS initiation. Two channels of LPCI Reactor Steam Dome Pressure - Low (Pump Permissive) Function are only required to be OPERABLE when the LPCI is required to be OPERABLE to ensure that no single instrument failure can preclude LPCI initiation. Per Footnote (a) to Table 3.3.5.1-1, these ECCS Functions are only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

1.e, 2.e. Reactor Steam Dome Pressure - Bypass Timer (Pump Permissive)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. The Bypass Timer channels allow the CS and LPCI pumps to start on Reactor Vessel Water Level - Low Low after the time delay times out, even if the reactor steam dome pressure is above its permissive setpoint. This ensures that, starting the pumps of the low pressure ECCS subsystems will occur on a Reactor Vessel Water Level - Low Low signal after an 18 minute time delay (Refs. 7 and 8). The Reactor Steam Dome Pressure - Time Delay (Pump Permissive) is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

The Reactor Steam Dome Pressure - Bypass Timer (Pump Permissive) signals are initiated from four time delay relays.

The Allowable Value is long enough to provide sufficient time for the operator to inhibit any unnecessary ADS actuation, yet short enough to limit the peak cladding temperature to less than 2200°F.

Two channels of CS Reactor Steam Dome Pressure - Bypass Timer (Pump Permissive) Function are only required to be OPERABLE when the CS is required to be OPERABLE to ensure that no single instrument failure can preclude CS initiation. Two channels of LPCI Reactor Steam Dome Pressure - Bypass Timer (Pump Permissive) Function are only required to be OPERABLE when the LPCI is required to be OPERABLE

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-13 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) to ensure that no single instrument failure can preclude LPCI initiation.

Per Footnote (a) to Table 3.3.5.1-1, these ECCS Functions are only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

1.f, 2.f. Core Spray and Low Pressure Coolant Injection Pump Start -

Time Delay Relay The purpose of the time delay relays is to stagger the start of the CS and LPCI pumps that are in each of Divisions 1 and 2, thus limiting the starting transients on the 4.16 kV essential buses. The CS and LPCI Pump Start - Time Delay Relays are assumed to be OPERABLE in the accident and transient analyses requiring ECCS initiation. That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources.

There are two CS Pump Start - Time Delay Relay channels, one in each of the CS pump start logic circuits. While each CS pump time delay relay is dedicated to a single pump start logic, a single failure of a CS Pump Start - Time Delay Relay channel could result in the failure of the three low pressure ECCS pumps, powered from the same 4.16 kV essential bus, to perform their intended function (e.g., as in the case where two ECCS pumps on one 4.16 kV essential bus start simultaneously due to an inoperable time delay relay). This still leaves three of the six low pressure ECCS pumps OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation).

Sixteen Low Pressure Coolant Injection Pump Start - Time Delay Relay channels, four in each of the LPCI pump start logic circuits, are required to be OPERABLE to ensure that no single instrument failure can preclude the associated LPCI pump start. The Allowable Values for the CS and LPCI Pump Start - Time Delay Relays are chosen short enough so that ECCS operation is not degraded.

Each CS and LPCI Pump Start - Time Delay Relay Function is required to be OPERABLE only when the associated ECCS subsystem is required to be OPERABLE. Per Footnote (a) to Table 3.3.5.1-1, these ECCS Functions are only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the LPCI subsystems.

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-14 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.g. Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass)

The minimum flow instruments are provided to protect the associated LPCI pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The LPCI Pump Discharge Flow - Low (Bypass) Function is assumed to be OPERABLE and capable of closing the minimum flow valves to ensure that the LPCI flows assumed during the transients and accidents analyzed in References 1 and 2 are met. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

One flow switch per LPCI pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each switch causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the pump start. This delay can reduce the reactor vessel inventory loss (to the suppression pool) during the startup of the RHR pump while aligned in the shutdown cooling mode, since it provides time (prior to opening the minimum flow valve) to manually increase RHR flow above the minimum flow closure setpoint. The LPCI Pump Discharge Flow - Low (Bypass) Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.

Each channel of LPCI Pump Discharge Flow - Low (Bypass) Function (four LPCI channels) is only required to be OPERABLE when the associated LPCI pump is required to be OPERABLE to ensure that no single instrument failure can preclude the LPCI function. Per Footnote (a) to Table 3.3.5.1-1, this LPCI Function is only required to be OPERABLE in MODES 4 and 5 whenever the associated LPCI pump is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the LPCI subsystems.

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-15 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.h, 2.k. Reactor Steam Dome Pressure - Low (Break Detection) and Reactor Steam Dome Pressure - Time Delay Relay (Break Detection)

The purpose of the Reactor Steam Dome Pressure - Low (Break Detection) and Reactor Steam Dome Pressure - Time Delay Relay (Break Detection) Functions are to optimize the LPCI Loop Select Logic sensitivity if the logic previously actuated recirculation pump trips. This is accomplished by preventing the logic from continuing on to the unbroken loop selection activity until reactor steam dome pressure has dropped below a specified value. These Functions are only required to be OPERABLE for the DBA LOCA analysis, i.e., if the break location is in the recirculation system suction piping (Ref. 2). For a DBA LOCA, the analysis assumes that the LPCI Loop Select Logic successfully identifies and directs LPCI flow to the unbroken recirculation loop so that core reflooding is accomplished in time to ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. For other LOCA events, (i.e., non-DBA recirculation system pipe breaks), or other RPV pipe breaks the success of the Loop Select Logic is less critical than for the DBA.

Reactor Steam Dome Pressure - Low (Break Detection) signals are initiated from four pressure switches that sense the reactor steam dome pressure. Reactor Steam Dome Pressure - Time Delay Relay (Break Detection) signals are initiated from two time delay relays.

The Reactor Steam Dome Pressure - Low (Break Detection) Allowable Value is chosen to allow for coastdown of any recirculation pump which has just tripped, thus optimizing the sensitivity of the LPCI Loop Select Logic while ensuring that LPCI injection is not delayed. The Reactor Steam Dome Pressure - Time Delay Relay (Break Detection) Allowable Value is chosen to allow momentum effects to establish the maximum differential pressure for break detection.

Four channels of the Reactor Steam Dome Pressure - Low (Break Detection) Function and two channels of the Reactor Steam Dome Pressure - Time Delay Relay (Break Detection) Function are only required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single failure can prevent the LPCI Loop Select Logic from successfully selecting the unbroken recirculation loop for LPCI injection. These Functions are not required to be OPERABLE in MODES 4 and 5 because, in those MODES, the loop for selection is controlled by plant operating procedures, which ensure an OPERABLE LPCI flow path.

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-16 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.i, 2.l. Recirculation Pump Differential Pressure - High (Break Detection) and Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection)

Recirculation pump differential pressure signals are used by the LPCI Loop Select Logic to determine if either recirculation pump is running. If either pump is not running, i.e., single loop operation, the logic, after a short time delay, sends a trip signal to both recirculation pumps. This is necessary to eliminate the possibility of small pipe breaks being masked by a running recirculation pump. These Functions are only required to be OPERABLE for the DBA LOCA analysis, i.e., if the break location is in the recirculation system suction piping (Ref. 2). For a DBA LOCA, the analysis assumes that the LPCI Loop Select Logic successfully identifies and directs LPCI flow to the unbroken recirculation loop so that core reflooding is accomplished in time to ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. For other LOCA events (i.e., non-DBA recirculation system pipe breaks or other RPV pipe breaks), the success of the Loop Select Logic is less critical than for the DBA.

Recirculation Pump Differential Pressure - High (Break Detection) signals are initiated from eight differential pressure switches, four of which sense the pressure differential between the suction and discharge of each recirculation pump. Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection) signals are initiated by two time delay relays.

The Recirculation Pump Differential Pressure - High (Break Detection)

Allowable Value is chosen to be as low as possible, while still maintaining the ability to differentiate between a running and non-running recirculation pump. Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection) Allowable Value is chosen to allow enough time to determine the status of the operating conditions of the recirculation pumps.

Eight channels of the Recirculation Pump Differential Pressure - High (Break Detection) Function and two channels of the Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection) Function are only required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single failure can prevent the LPCI Loop Select Logic from successfully determining if either recirculation pump is running. This Function is not required to be OPERABLE in MODES 4 and 5 because, in those MODES, the loop for selection is controlled by plant operating procedures, which ensure an OPERABLE LPCI flow path.

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-18 Revision No. 38 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) selecting the unbroken recirculation loop for LPCI injection. This Function is not required to be OPERABLE in MODES 4 and 5 because, in those MODES, the loop for selection is controlled by plant operating procedures, which ensure an OPERABLE LPCI flow path.

HPCI System 3.a. Reactor Vessel Water Level - Low Low Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCI System is initiated at Low Low to maintain level above the top of the active fuel. The Reactor Vessel Water Level -

Low Low is one of the Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in Reference 2.

Additionally, the Reactor Vessel Water Level - Low Low Function associated with HPCI along with the Drywell Pressure - High Function is directly assumed in the analysis of the recirculation line break (Ref. 1).

The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low 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.

The Reactor Vessel Water Level - Low Low Allowable Value is high enough such that for complete loss of feedwater flow when the reactor vessel is isolated, the Reactor Core Isolation Cooling (RCIC) System flow with HPCI assumed to fail will be sufficient to avoid injection of low pressure ECCS.

Four channels of Reactor Vessel Water Level - Low Low Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation.

Refer to LCO 3.5.1 for HPCI Applicability Bases.

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-25 Revision No. 38 BASES ACTIONS (continued) and untripped such that both trip systems lose initiation capability; (e) two or more Function 2.f channels are inoperable and untripped such that one or more pumps in both LPCI subsystems lose initiation (i.e., time delay) capability; (f) two or more Function 2.h channels are inoperable and untripped such that both trip systems lose initiation capability; or (g) two Function 2.k channels are inoperable and untripped. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system of low pressure ECCS and EDGs to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and EDGs being concurrently declared inoperable.

For Required Action B.2, redundant automatic initiation capability is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system (a trip system in this case is defined as channels associated with the parallel level in the logic arrangement).

In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.3 is not appropriate and the feature(s) associated with the inoperable, untripped channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1 to Required Action B.1), Required Action B.1 is only applicable in MODES 1, 2, and 3.

In MODES 4 and 5, the specific initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action B.3) is allowed during MODES 4 and 5. There is no similar Note provided for Required Action B.2 since HPCI instrumentation is not required in MODES 4 and 5; thus, a Note is not necessary. Notes are also provided (the Note 2 to Required Action B.1 and the Note to Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable.

This ensures that the proper loss of initiation capability check is performed.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action B.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same Function as described in the paragraph above. For Required Action B.2, the Completion Time

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-27 Revision No. 38 BASES ACTIONS (continued) of the associated system to be declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g.,

both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being concurrently declared inoperable. For these Functions the affected portions are the associated low pressure ECCS pumps.

In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1), Required Action C.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of automatic initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action C.2) is allowed during MODES 4 and 5.

The Note 2 states that Required Action C.1 is only applicable for Functions 1.c, 1.d, 1.e, 1.f, 2.c, 2.d, 2.e, 2.i, 2.j, 2.l, and 2.m. Required Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 3 and considered acceptable for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed by Required Action C.2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.

Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 3) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its

ECCS Instrumentation B 3.3.5.1 Monticello B 3.3.5.1-29 Revision No. 38 BASES ACTIONS (continued) and D.2.2 (e.g., as in the case where shifting the suction source could drain down the HPCI suction piping), Condition H must be entered and its Required Action taken.

E.1 and E.2 Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) Function results in redundant automatic initiation capability being lost for the feature(s). For Required Action E.1, the features would be those that are initiated by Function 2.g (i.e., LPCI). Redundant automatic initiation capability is lost if one or more Function 2.g channels associated with pumps in LPCI subsystem A and one or more Function 2.g channels associated with pumps in LPCI subsystem B are inoperable. Since each inoperable channel would have Required Action E.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected LPCI pump to be declared inoperable. However, since channels for more than one LPCI pump are inoperable, and the Completion Times started concurrently for the channels of the LPCI pumps, this results in the affected ECCS pumps being concurrently declared inoperable.

In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is not appropriate and the subsystem associated with each inoperable channel must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1 to Required Action E.1),

Required Action E.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 7 days (as allowed by Required Action E.2) is allowed during MODES 4 and 5. A Note is also provided (the Note 2 to Required Action E.1) to delineate that Required Action E.1 is only applicable to the LPCI Function. Required Action E.1 is not applicable to HPCI Function 3.f since the loss of one channel results in a loss of the Function (one-out-of-one logic). This loss was considered during the development of Reference 3 and considered acceptable for the 7 days allowed by Required Action E.2. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-1 Revision 0 B 3.3 INSTRUMENTATION B 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation BASES BACKGROUND The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF.

If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.4 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.

Technical Specifications are required by 10 CFR 50.36 to include limiting safety system settings (LSSS) for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded." The Analytical Limit is the limit of the process variable at which a safety action is initiated to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur. The actual settings for the automatic isolation channels are the same as those established for the same functions in MODES 1, 2, and 3 in LCO 3.3.5.1, "Emergency Core Cooling System (ECCS)

Instrumentation," or LCO 3.3.6.1, "Primary Containment Isolation instrumentation".

With the unit in MODE 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses.

RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.4 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur. Under the definition of DRAIN TIME, some penetration flow paths may be excluded from the DRAIN TIME calculation if they will be isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-2 Revision 0 BASES BACKGROUND (continued)

The purpose of the RPV Water Inventory Control Instrumentation is to support the requirements of LCO 3.5.2, RPV Water Inventory Control, and the definition of DRAIN TIME. There are functions that are required for manual operation of the ECCS injection/spray subsystem required to be OPERABLE by LCO 3.5.2 and other functions that support automatic isolation of Residual Heat Removal subsystem and Reactor Water Cleanup system penetration flow path(s) on low RPV water level.

The RPV Water Inventory Control Instrumentation supports operation of core spray (CS) and low pressure coolant injection (LPCI). The equipment involved with each of these systems is described in the Bases for LCO 3.5.2.

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is not required SAFETY to mitigate any events or accidents evaluated in the safety analyses. RPV ANALYSES, LCO, water inventory control is required in MODES 4 and 5 to protect and APPLICABILITY Safety Limit 2.1.1.4 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur.

A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is postulated in which a single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g.,

seismic event, loss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can be manually initiated to maintain adequate reactor vessel water level.

As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety.

Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Permissive and interlock setpoints are generally considered as nominal values without regard to measurement accuracy.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-3 Revision 0 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

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

Core Spray and Low Pressure Coolant Injection Systems 1.a, 2.a. Reactor Steam Dome Pressure - Low (Injection Permissive)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystems. This function ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. While it is assured during MODES 4 and 5 that the reactor steam dome pressure will be below the ECCS maximum design pressure, the Reactor Steam Dome Pressure -

Low (Injection Permissive) signals are assumed to be OPERABLE and capable of permitting initiation of the ECCS.

The Reactor Steam Dome Pressure - Low (Injection Permissive) signals are initiated from two pressure switches (shared by both CS and LPCI) that sense the reactor dome pressure.

The Allowable Value is low enough to prevent overpressurizing the equipment in the low pressure ECCS.

The two channels of Reactor Steam Dome Pressure - Low (Injection Permissive) Function are required to be OPERABLE in MODES 4 and 5 when ECCS is required to be OPERABLE by LCO 3.5.2.

2.b. Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass)

The minimum flow instruments are provided to protect the associated LPCI pump from overheating when the pump is operating and the associated injection valve is not fully open. The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the pump start. This delay can reduce the reactor vessel inventory loss (to the suppression pool) during the startup of the RHR pump while aligned in the shutdown cooling mode, since it provides time (prior to opening the minimum flow valve) to manually increase RHR flow above the minimum flow closure setpoint.

One flow switch per LPCI pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each switch causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-4 Revision 0 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) for 10 seconds after the pump start. The time delay is provided to limit reactor vessel inventory loss during the startup of the Residual Heat Removal (RHR) shutdown cooling mode.

The LPCI Pump Discharge Flow - Low (Bypass) Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.

One channel of the LPCI Pump Discharge Flow - Low (Bypass) Function is required to be OPERABLE in MODES 4 and 5 when the associated LPCI pump is required to be OPERABLE by LCO 3.5.2 to ensure the pump is capable of injecting into the Reactor Pressure Vessel when manually operated.

RHR System Isolation 3.a - Reactor Vessel Water Level - Low The definition of Drain Time allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low Function associated with RHR System isolation may be credited for automatic isolation of penetration flow paths associated with the RHR System.

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. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low Function are available, only two channels (all in the same trip system) are required to be OPERABLE to ensure automatic isolation of one of the two isolation valves.

The Reactor Vessel Water Level - Low Allowable Value was chosen to be the same as the Primary Containment Isolation Instrumentation Reactor Vessel Water Level - Low Allowable Value (LCO 3.3.6.1), since the capability to cool the fuel may be threatened.

The Reactor Vessel Water Level - Low Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.

This Function isolates the Group 2 drywell and sump isolation valves.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-5 Revision 0 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Reactor Water Cleanup (RWCU) System Isolation 4.a - Reactor Vessel Water level - Low Low The definition of Drain Time allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low Low Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.

Reactor Vessel Water Level - Low Low signals are initiated from four differential pressure 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. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low Low Function are available, only two channels (all in the same trip system) are required to be OPERABLE.

The Reactor Vessel Water Level - Low Low Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level - Low Low Allowable Value (LCO 3.3.5.1), since the capability to cool the fuel may be threatened.

The Reactor Vessel Water Level - Low Low Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.

This Function isolates the Group 3 valves.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-6 Revision 0 BASES ACTIONS A Note has been provided to modify the ACTIONS related to RPV Water Inventory Control 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 continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPV Water Inventory Control instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable RPV Water Inventory Control instrumentation channel.

A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.3-1. The applicable Condition specified in the Table is Function or other specified condition dependent. Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.

B.1 and B.2 RHR System Isolation, Reactor Vessel Water Level - Low and Reactor Water Cleanup System Isolation, Reactor Vessel Water Level - Low Low functions are applicable when automatic isolation of the associated penetration flow path is credited in calculating Drain Time. If the instrumentation is inoperable, Required Action B.1 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation. Required Action B.2 directs calculation of DRAIN TIME. The calculation cannot credit automatic isolation of the affected penetration flow paths.

C.1 Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. If the permissive is inoperable, manual initiation of ECCS is prohibited. Therefore, the permissive must be placed in the trip condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the permissive in the trip condition, manual operation may be performed. Prior to placing the permissive in the tripped condition, the operator can take manual control of the pump and the injection valve to inject water into the RPV.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is intended to allow the operator time to evaluate any discovered inoperabilities and to place the channel in trip.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-7 Revision 0 BASES ACTIONS (continued)

D.1 If a LPCI Pump Discharge Flow - Low Bypass function is inoperable, there is a risk that the associated low pressure ECCS pump could overheat when the pump is operating and the associated injection valve is not fully open. In this condition, the operator can take manual control of the pump and the injection valve to ensure the pump does not overheat.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities. The Completion Time is appropriate given the ability to manually start the ECCS pumps and open the injection valves and to manually ensure the pump does not overheat.

E.1 With the Required Action and associated Completion Time of Condition C or D not met, the associated low pressure ECCS injection/spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-8 Revision 0 BASES SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RPV Water REQUIREMENTS Inventory Control instrumentation Function are found in the SRs column of Table 3.3.5.3-1.

SR 3.3.5.3.1 Performance of the CHANNEL CHECK 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 the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST.

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 of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 3.3.5.3.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.

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

The Frequency of 92 days is based upon operating experience that demonstrates channel failure is rare.

RPV Water Inventory Control Instrumentation B 3.3.5.3 Monticello B 3.3.5.3-9 Revision 0 BASES REFERENCES

1.

Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup,"

November 1984.

2.

Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.

3.

Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F), " August 1992.

4.

NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.

5.

Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.

Primary Containment Isolation Instrumentation B 3.3.6.1 Monticello B 3.3.6.1-15 Revision No. 34 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization.

This Function isolates the Group 2 RHR shutdown cooling supply isolation valves.

6.b. Reactor Vessel Water Level - Low Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water Level - Low Function associated with RHR Shutdown Cooling System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the recirculation and MSL. The RHR Shutdown Cooling System isolation on low RPV water level supports actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System.

Reactor Vessel Water Level - Low signals are initiated from four differential pressure 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 (two channels per trip system) of the Reactor Vessel Water Level - Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (a) to Table 3.3.6.1-1), only one channel per trip system (with an isolation signal available to one shutdown cooling pump supply isolation valve) of the Reactor Vessel Water Level - Low Function is required to be OPERABLE in MODES 4 and 5, provided RHR Shutdown Cooling System integrity is maintained. System integrity is maintained provided the piping is intact and no maintenance is being performed that has the potential for draining the reactor vessel through the system.

The Reactor Vessel Water Level - Low Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level - Low Allowable Value (LCO 3.3.1.1), since the capability to cool the fuel may be threatened.

The Reactor Vessel Water Level - Low Function is only required to be OPERABLE in MODES 3, 4, and 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2, another isolation (i.e., Reactor Steam Dome Pressure - High) and

Primary Containment Isolation Instrumentation B 3.3.6.1 Monticello B 3.3.6.1-21 Revision No. 34 BASES ACTIONS (continued)

I.1 and I.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated.

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Primary REQUIREMENTS Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1.

The Surveillances are modified by a Note to indicate that when a channel (a channel that is directed to two trip systems is considered to be one channel) is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains primary containment isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path(s) when necessary.

SR 3.3.6.1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 the instrument channels could be an indication of excessive instrument drift in one of the channels or of 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.

Secondary Containment Isolation Instrumentation B 3.3.6.2 Monticello B 3.3.6.2-4 Revision No. 42 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Reactor Vessel Water Level - Low Low Allowable Value was chosen to be the same as the High Pressure Coolant Injection/Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level - Low Low Allowable Value (LCO 3.3.5.1, "Emergency Core Cooling System (ECCS)

Instrumentation," and LCO 3.3.5.2, "Reactor Core Isolation Cooling (RCIC) System Instrumentation"), since this could indicate that the capability to cool the fuel is being threatened.

The Reactor Vessel Water Level - Low Low Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs) to ensure that control room operator and offsite dose limits are not exceeded if core damage occurs.

2. Drywell Pressure - High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Drywell Pressure - High Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals to ensure that any offsite releases are within the limits calculated in the safety analysis.

High drywell pressure signals are initiated from pressure switches that sense the pressure in the drywell. Four channels of Drywell Pressure -

High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.

The Allowable Value was chosen to be the same as the Reactor Protection System (RPS) Drywell Pressure - High Function Allowable Value (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation")

since this is indicative of a loss of coolant accident (LOCA).

The Drywell Pressure - High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.

CREF System Instrumentation B 3.3.7.1 Monticello B 3.3.7.1-4 Revision No. 42 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Vessel Water Level - Low Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low Low Allowable Value was chosen to be the same as the High Pressure Coolant Injection/Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level - Low Low Allowable Value (LCO 3.3.5.1, "Emergency Core Cooling System (ECCS)

Instrumentation," and LCO 3.3.5.2, "Reactor Core Isolation Cooling (RCIC) System Instrumentation"), since this could indicate that the capability to cool the fuel is being threatened.

The Reactor Vessel Water Level - Low Low Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs) to ensure that control room dose limits are not exceeded if core damage occurs.

2. Drywell Pressure - High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). The CREF System is initiated to maintain the habitability of the control room envelope. The Drywell Pressure - High Function is one of the Functions assumed to be OPERABLE and capable of providing an initiation signal to ensure that control room doses are within the limits calculated in the safety analysis.

High drywell pressure signals are initiated from pressure switches that sense the pressure in the drywell. Four channels of Drywell Pressure -

High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.

The Allowable Value was chosen to be the same as the Reactor Protection System (RPS) Drywell Pressure - High Function Allowable Value (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation")

since this is indicative of a loss of coolant accident (LOCA).

The Drywell Pressure - High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4

CREF System Instrumentation B 3.3.7.1 Monticello B 3.3.7.1-5 Revision No. 42 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.

3, 4. Reactor Building Ventilation Exhaust and Refueling Floor Radiation

- High High reactor building ventilation exhaust radiation or refuel floor radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident. When Reactor Building Ventilation Exhaust Radiation - High or Refueling Floor Radiation - High is detected the CREF System is initiated. These actions are required to mitigate the consequences of the LOCA or FHA involving recently irradiated fuel by limiting the control room doses to less than the limits calculated in the safety analysis (Refs. 1 and 4).

The Reactor Building Ventilation Exhaust Radiation - High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building. The Refueling Floor Radiation - High signals are initiated for radiation detectors that are located to monitor the environment of the refuel floor area. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Building Ventilation Exhaust Radiation - High Function and two channels of Refueling Floor Radiation - High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.

The Reactor Building Ventilation Exhaust Radiation - High and Refueling Floor Radiation - High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel un-covering or dropped fuel assemblies) must be provided to ensure that control room dose limits are not exceeded. Due to radioactive decay, these Functions are only required to initiate the CREF System during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

RPS Electric Power Monitoring B 3.3.8.2 Monticello B 3.3.8.2-3 Revision No. 01 BASES LCO (continued)

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 if the 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 Allowable Values for the instrument settings are based on the RPS buses providing 57 Hz and 116 V +/- 10% to all equipment. The Allowable Values are within the ratings of the RPS bus powered components and ensure their protection from sustained abnormal power.

APPLICABILITY The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS bus powered components from the inservice MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass 1E source supplying power to the RPS bus can occur as a result of any random single failure, the OPERABILITY of the RPS electric power monitoring assemblies is required when the RPS bus powered components are required to be OPERABLE. This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES 1, 2, and 3, in MODES 4 and 5 with both residual heat removal (RHR) shutdown cooling supply isolation valves open, in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, during movement of recently irradiated fuel assemblies in the secondary containment, and during operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS A.1 If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inservice power supply is inoperable, the OPERABLE assembly will still provide protection to the RPS bus powered components under degraded voltage or frequency conditions. However, the reliability and redundancy of the RPS Electric Power Monitoring System is reduced, and only a limited time (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) is allowed to restore the inoperable assembly to OPERABLE status. If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply(s) must be removed from service (Required Action A.1). This

RPS Electric Power Monitoring B 3.3.8.2 Monticello B 3.3.8.2-5 Revision No. 01 BASES ACTIONS (continued) bus loads (e.g., scram of control rods) is not required. The plant shutdown is accomplished by placing the plant in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1 and D.2 If any Required Action and associated Completion Time of Condition A or B are not met in MODE 4 or 5 with both RHR shutdown cooling supply isolation valves open, action must be immediately initiated to either restore one electric power monitoring assembly to OPERABLE status for the inservice power source supplying the required instrumentation powered from the RPS bus (Required Action D.1) or to isolate the RHR Shutdown Cooling System (Required Action D.2). Required Action D.1 is provided because the RHR Shutdown Cooling System may be needed to provide core cooling. All actions must continue until the applicable Required Actions are completed.

E.1 If any Required Action and associated Completion Time of Condition A or B are not met in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, the operator must immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Required Action E.1 results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g., scram of control rods) is not required.

F.1.1, F.1.2, F.2.1, F.2.2, F.3.1, and F.3.2 If any Required Action and associated Completion Time of Condition A or B are not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, the ability to isolate the secondary containment, start the Standby Gas Treatment (SGT) System, and start Control Room Emergency Filtration (CREF) System cannot be ensured. Therefore, actions must be immediately performed to ensure the ability to maintain the secondary containment isolation, SGT System, and CREF System functions. Isolating the affected penetration flow path(s) and starting the associated SGT subsystem(s) and CREF subsystem(s) (Required Actions F.1.1, F.2.1, and F.3.1) performs the

ECCS - Operating 3.5.1 Monticello B 3.5.1-1 Revision No. 47 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS.

Although no credit is taken in the safety analyses for the condensate storage tanks (CSTs), they are capable of providing a source of water for the HPCI, LPCI, and CS Systems.

On receipt of an initiation signal, ECCS pumps automatically start and the system aligns and the pumps inject water, taken either from the CSTs or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps.

Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed. The HPCI pump discharge pressure almost immediately exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core. If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI and CS. In this event, the ADS timed sequence would be allowed to time out and open the selected safety/relief valves (S/RVs) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.

Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool is circulated through a heat exchanger cooled by the RHR Service Water System. Depending on the location and size of the break, portions of the ECCS may be ineffective; however, the overall design is effective in cooling the core regardless of the size or location of the piping break.

The combined operation of all ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipment.

ECCS - Operating 3.5.1 Monticello B 3.5.1-6 Revision No. 47 BASES LCO (continued)

As noted, LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR shutdown cooling supply isolation interlock in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.

APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, when reactor steam dome pressure is 150 psig, ADS and HPCI are not required to be OPERABLE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. RECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "RPV Water Inventory ControECCS -

Shutdownl."

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI subsystem. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 If one LPCI pump is inoperable, the inoperable pump must be restored to OPERABLE status within 30 days. In this condition, the remaining OPERABLE pumps provide adequate core cooling during a LOCA.

However, overall LPCI reliability is reduced, because a single failure in one of the remaining OPERABLE LPCI subsystems, concurrent with a LOCA, may result in the LPCI subsystems not being able to perform their intended safety function. The 30 day Completion Time is based on a reliability study cited in Reference 11 that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-1 Revision No. 47 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM EMERGENCY CORE COOLING SYSTEM (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 RPV Water Inventory Control ECCS - Shutdown BASES BACKGROUND The Reactor Pressure Vessel (RPV) contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF. If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.4 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.A description of the Core Spray (CS) System and the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCS - Operating."

APPLICABLE With the unit in MODES 4 or 5, RPV water inventory control is not The ECCS performance is evaluated for the entire spectrum of break SAFETY required to mitigate any events or accidents evaluated in the safetysizes for a postulated loss of coolant accident (LOCA). The long term ANALYSES analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.4 and the fuel cladding barrier to prevent the release of radioactive material to the environment should an unexpected draining event occur.

A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is considered in which single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g.,

seismic event, loss of normal power, single human error)cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one low pressure ECCS injection/spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. It is reasonable to assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. To provide redundancy, a minimum of two low pressure ECCS injection/spray subsystems are required to be OPERABLE in MODES 4 and 5.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-2 Revision No. 47 As discussed in References 3, 4, 5, 6, 7 and 8 operating experience has shown RPV water inventory to be significant to public health and safety.

Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

The low pressure ECCS subsystems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The RPV water level must be controlled in MODES 4 and 5 to ensure that if an unexpected draining event should occur, the reactor coolant water level remains above the top of the active irradiated fuel as required by Safety Limit 2.1.1.4.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-3 Revision No. 47 BASES LCO (continued)

The Limiting Condition for Operation (LCO) requires the DRAIN TIME of RPV water inventory to the TAF to be 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A DRAIN TIME of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate an unexpected draining of reactor coolant. An event that could cause loss of RPV water inventory and result in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1.1.4 and can be managed as part of normal plant operation.

Two One low pressure ECCS injection/spray subsystems are is required to be OPERABLE and capable of being manually started to provide defense-in-depth should an unexpected draining event occur.. The A low pressure ECCS injection/spray subsystems consists of either onetwo Core Spray (CS) subsystems or and twoone Low Pressure Coolant Injection (LPCI) subsystems. Each CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or one or two both condensate storage tanks (CSTs) to the reactor pressure vessel (RPV). Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or from one or both CSTs to the RPV. A single LPCI pump is required per subsystem because of similar injection capacity in relation to a CS subsystem. In addition, in MODES 4 and 5 the RHR System cross-tie valve is not required to be open. Management of gas voids is important to ECCS injection/spray subsystem OPERABILITY (Ref. 2).

The LCO is modified by a Note which allows a required As noted, one LPCI subsystem to may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable.

Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor.

Because of the restrictions on DRAIN TIME, sufficient time will be available following an unexpected draining event to manually align and initiate LPCI subsystem operation to maintain RPV water inventory prior to the RPV water level reaching the TAF.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-4 Revision No. 47 BASES LCO (continued) operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Because of low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery.

APPLICABILITY RPV water inventory control is required in MODES 4 and 5.

Requirements on water inventory control in other MODES are contained in LCOs in Section 3.3, Instrumentation, and other LCOs in Section 3.5, ECCS, RCIC, and RPV Water Inventory Control. RPV water inventory control is required to protect Safety Limit 2.1.1.4 which is applicable whenever irradiated fuel is in the reactor vessel.

OPERABILITY of the low pressure ECCS injection/spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1. ECCS subsystems are not required to be OPERABLE during MODE 5 with the spent fuel storage pool gates removed and the water level maintained at 21 ft 11 inches above the RPV flange. This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown.

The Automatic Depressurization System is not required to be OPERABLE during MODES 4 and 5 because the RPV pressure is 150 psig, and the CS System and the LPCI subsystems can provide core cooling without any depressurization of the primary system.

The High Pressure Coolant Injection System is not required to be OPERABLE during MODES 4 and 5 since the low pressure ECCS injection/spray subsystems can provide sufficient flow to the vessel.

ACTIONS A.1 and B.1 If any onethe required low pressure ECCS injection/spray subsystem is inoperable, it the inoperable subsystem must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In this Condition, the LCO controls on DRAIN TIME minimize the possibility that an unexpected draining event could necessitate the use of the ECCS injection/spray subsystem, however the defense-in-depth provided by the ECCS injection/spray subsystem is lost.

the remaining OPERABLE subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown.

However, overall system reliability is reduced because a single failure in

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-5 Revision No. 47 the remaining OPERABLE subsystem concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for restoring the required low pressure ECCS injection/spray subsystem to OPERABLE status is based on engineering judgment that considersed the LCO controls on DRAIN TIME remaining available subsystem and the low probability of an unexpected draining vessel draindown event that would result in loss of RPV water inventory.

WithIf the inoperable ECCS injection/spray subsystem is not restored to OPERABLE status within the required Completion Time, action must be initiated immediately initiated to establish a method of water injection capable of operating without offsite electrical power. The method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The method of water injection may be manually initiated and may consist of one or more systems or subsystems, and must be able to access water inventory capable of maintaining the RPV water level above the TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . If recirculation of injected water would occur, it may be credited in determining the necessary water volume.

C.1, C.2, and C.3 With the DRAIN TIME less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months but greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , compensatory measures should be taken to ensure the ability to implement mitigating actions should an unexpected draining event occur.

Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

The secondary containment provides a controlled volume in which fission products can be contained, diluted, and processed prior to release to the environment. Required Action C.1 requires verification of the capability to BASES ACTIONS (continued) establish the secondary containment boundary in less than the DRAIN TIME. The required verification confirms actions to establish the secondary containment boundary are preplanned and necessary materials are available. The secondary containment boundary is considered established when one Standby Gas Treatment (SGT) subsystem is capable of maintaining a negative pressure in the secondary containment with respect to the environment. suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-6 Revision No. 47 potential for fission product release. Actions must continue until OPDRVs are suspended.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-7 Revision No. 47 BASES ACTIONS (continued)

Verification that the secondary containment boundary can be established must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment. Secondary containment penetration flow paths form a part of the secondary containment boundary. Required Action C.2 requires verification of the capability to isolate each secondary containment penetration flow path in less than the DRAIN TIME. The required verification confirms actions to isolate the secondary containment penetration flow paths are preplanned and necessary materials are available. Power operated valves are not required to receive automatic isolation signals if they can be closed manually within the required time.

Verification that the secondary containment penetration flow paths can be isolated must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases. Required Action C.3 requires verification of the capability to place one SGT subsystem in operation in less than the DRAIN TIME. The required verification confirms actions to place a SGT subsystem in operation are preplanned and necessary materials are available. Verification that a SGT subsystem can be placed in operation must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

D.1, D.2, D.3, and D.4 With the DRAIN TIME less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , mitigating actions are implemented in case an unexpected draining event should occur. Note that if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action E.1 is also applicable.

Required Action D.1 requires immediate action to establish an additional method of water injection augmenting the ECCS injection/spray subsystem required by the LCO. The additional method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity BASES ACTIONS (continued) should an unexpected draining event occur. The Note to Required Action D.1 states that either the ECCS injection/spray subsystem or the

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-8 Revision No. 47 additional method of water injection must be capable of operating without offsite electrical power. The additional method of water injection may be manually initiated and may consist of one or more systems or subsystems. The additional method of water injection must be able to C.1, C.2, D.1, D.2, and D.3 With both of the required ECCS injection/spray subsystems inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must immediately be initiated to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. One ECCS injection/spray subsystem must also be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If at least one low pressure ECCS injection/spray subsystem is not restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE; one standby gas treatment subsystem is OPERABLE; and secondary containment isolation capability is available in each associated penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability). These administrative controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. OPERABILITY may be verified by an administrative check, or by examining logs or other information, to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-9 Revision No. 47 access water inventory capable of being injected to maintain the RPV water level above the TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The additional method of water injection and the ECCS injection/spray subsystem may share all or part of the same water sources. If recirculation of injected water would occur, it may be credited in determining the required water volume.

Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

The secondary containment provides a control volume in which fission products can be contained, diluted, and processed prior to release to the environment. Required Action D.2 requires that actions be immediately initiated to establish the secondary containment boundary. With the secondary containment boundary established, one SGT subsystem is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

The secondary containment penetrations form a part of the secondary containment boundary. Required Action D.3 requires that actions be immediately initiated to verify that each secondary containment penetration flow path is isolated or to verify that it can be manually isolated from the control room.

One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases. Required Action D.4 requires that actions be immediately initiated to verify that at least one SGT subsystem is capable of being placed in operation. The required verification is an administrative activity and does not require manipulation or testing of equipment.

E.1 If the Required Actions and associated Completion times of Conditions C or D are not met or if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , actions must be initiated immediately to restore the DRAIN TIME to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In this condition, there may be insufficient time to respond to an unexpected BASES ACTIONS (continued) draining event to prevent the RPV water inventory from reaching the TAF.

Note that Required Actions D.1, D.2, D.3, and D.4 are also applicable when DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-10 Revision No. 47 The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time to restore at least one low pressure ECCS injection/spray subsystem to OPERABLE status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS The minimum water level of -3 ft required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the CS subsystem System and or LPCI subsystem pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, allthe required ECCS injection/spray subsystems isare inoperable unless they areit is aligned to an OPERABLE CST.

When suppression pool level is < -3 ft, the CS andor LPCI subsystem s areis considered OPERABLE only if theyit can take suction from the CST(s), and the CST(s) water level is sufficient to provide the required NPSH and vortex prevention for the CS pump or LPCI pump. Therefore, a verification that either the suppression pool water level is -3 ft or that the required low pressure ECCS injection/spray subsystem s areis aligned to take suction from the CST(s) and the CST(s) contain 58,000 gallons of water, equivalent to 4 ft in both CSTs when they are cross-tied (normal configuration) and 7 ft in one CST when they are not cross-tied, ensures that the required low pressure ECCS injection/spray subsystems can supply at least 50,000 available gallons of makeup water to the RPV. The low pressure ECCS injection/spray suction is uncovered at the 2366 gallon level. However, as noted, only one required low pressure ECCS injection/spray subsystem may take credit for the CST option during OPDRVs. During OPDRVs, the volume in the CST(s) may not provide adequate makeup if the RPV were completely drained.

Therefore, only one low pressure ECCS injection/spray subsystem is allowed to use the CST(s). This ensures the other required ECCS subsystem has adequate makeup volume.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of these SRs was developed considering operating experience related to suppression pool water level and CST water level variations and instrument drift during the applicable MODES.

Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.

SR 3.5.2.2 The ECCS injection/spray subsystem flow path piping and components have the potential to develop voids and pockets of entrained gases.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-11 Revision No. 47 Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required ECCS injection/spray subsystem and may also prevent a water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

BASES SURVEILLANCE REQUIREMENTS (continued)

Selection of ECCS injection/spray subsystem locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The required ECCS injection/spray subsystem is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the ECCS injection/spray subsystem is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met.

Accumulated gas should be eliminated or brought within the acceptance criteria limits.

ECCS injection/spray subsystem locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-12 Revision No. 47 sufficient to assure system OPERABILITY during the Surveillance interval.

The 31 day Frequency is based on the gradual nature of void buildup in the ECCS injection/spray subsystem piping, the procedural controls governing system operation, and operating experience.

SR 3.5.2.2, SR 3.5.2.4, and SR 3.5.2.5 The Bases provided for SR 3.5.1.1, SR 3.5.1.7, and SR 3.5.1.10 are applicable to SR 3.5.2.2, SR 3.5.2.4, and SR 3.5.2.5, respectively.

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.3 Verifying the correct alignment for manual, power operated, and automatic valves in the required ECCS subsystem flow paths provides assurance that the proper flow paths will be availableexist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an BASES SURVEILLANCE REQUIREMENTS (continued) initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.

The Surveillance is modified by a Note which exempts system vent flow paths opened under administrative control. The administrative control should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication with the operators in the control room. This individual will have a method to rapidly close the system vent flow path if directed (Ref. 2).

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-13 Revision No. 47 SR 3.5.2.4 Verifying that the required ECCS injection/spray subsystem can be manually started and operate for at least 10 minutes demonstrates that the subsystem is available to mitigate a draining event. Testing the ECCS injection/spray subsystem through the recirculation line is necessary to avoid overfilling the refueling cavity. The minimum operating time of 10 minutes was based on engineering judgment. The performance frequency of 92 days is consistent with similar at-power testing required by SR 3.5.1.7.

SR 3.5.2.5 The required ECCS subsystem shall be capable of being manually operated. This Surveillance verifies that the required CS or LPCI subsystem (including the associated pump and valve(s)) can be manually operated to provide additional RPV Water Inventory, if needed.

The 24 month Frequency is based on the need to perform the 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 SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

SR 3.5.2.6 Verifying that each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated RPV water level isolation signal is required to prevent RPV water inventory from dropping below the TAF should an unexpected draining event occur.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-14 Revision No. 47 SR 3.5.2.7 This Surveillance verifies that the DRAIN TIME of RPV water inventory to the TAF is 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The period of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate draining of reactor coolant. Loss of RPV water inventory that would result in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1.1.4 and can be managed as part of normal plant operation.

The definition of DRAIN TIME states that realistic cross-sectional areas and drain rates are used in the calculation. A realistic drain rate may be determined using a single, step-wise, or integrated calculation considering the changing RPV water level during a draining event. For a Control Rod RPV penetration flow path with the Control Rod Drive Mechanism removed and not replaced with a blank flange, the realistic cross-sectional area is based on the control rod blade seated in the control rod guide tube. If the control rod blade will be raised from the BASES SURVEILLANCE REQUIREMENTS (continued) penetration to adjust or verify seating of the blade, the exposed cross-sectional area of the RPV penetration flow path is used.

The definition of DRAIN TIME excludes from the calculation those penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths. A blank flange or other bolted device must be connected with a sufficient number of bolts to prevent draining in the event of an Operating Basis Earthquake. Normal or expected leakage from closed systems or past isolation devices is permitted. Determination that a system is intact and closed or isolated must consider the status of branch lines and ongoing plant maintenance and testing activities.

The Residual Heat Removal (RHR) Shutdown Cooling System is only considered an intact closed system when misalignment issues (Reference 8) have been precluded by functional valve interlocks or by isolation devices, such that redirection of RPV water out of an RHR subsystem is precluded. Further, RHR Shutdown Cooling System is only considered an intact closed system if its controls have not been transferred to the Alternate Shutdown System, which disables the interlocks and isolation signals.

The exclusion of penetration flow paths from the determination of DRAIN TIME must consider the potential effects of a single operator error or

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2-15 Revision No. 47 initiating event on items supporting maintenance and testing (rigging, scaffolding, temporary shielding, piping plugs, snubber removal, freeze seals, etc.). If failure of such items could result and would cause a draining event from a closed system or between the RPV and the isolation device, the penetration flow path may not be excluded from the DRAIN TIME calculation.

Surveillance Requirement 3.0.1 requires SRs to be met between performances. Therefore, any changes in plant conditions that would change the DRAIN TIME requires that a new DRAIN TIME be determined.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of indications of RPV water level available to the operator.

RPV Water Inventory Control ECCS - Shutdown B 3.5.2 Monticello B 3.5.2 Last Revision No. 47 BASES REFERENCES

1.

USAR, Section 14.7.2.3.6.

2.

Amendment No. 189, Monticello Nuclear Generating Plant -

Issuance of Amendment Re: Technical Specifications Task Force Standard Technical Specifications Change Traveler TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation, June 21, 2016 (ADAMS Accession No. ML16125A165)

3. Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup,"

November 1984.

4. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.

5. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f), " August 1992.

6. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.

7. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.

8. General Electric Service Information Letter No. 388, "RHR Valve Misalignment During Shutdown Cooling Operation for BWR 3/4/5/6,"

February 1983.

RCIC System B 3.5.3 Monticello B 3.5.3-1 Revision No. 47 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.3 RCIC System BASES BACKGROUND The RCIC System is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level. Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference 1 are satisfied.

The RCIC System (Ref. 1) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping is provided from the condensate storage tanks (CSTs) and the suppression pool. Pump suction is normally aligned to the CSTs to minimize injection of suppression pool water into the RPV. However, if the water supply is low in any CST, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System.

The steam supply to the turbine is piped from a main steam line upstream of the associated inboard main steam line isolation valve.

The RCIC System is designed to provide core cooling for a wide range of reactor pressures (165 psia to 1135 psia). Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

The RCIC pump is provided with a minimum flow bypass line, which discharges to the suppression pool. The valve in this line automatically opens to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge piping is kept full of water. The RCIC System is normally aligned to the CSTs.

The height of water in the CSTs is sufficient to maintain the piping full of water up to the first isolation valve in the discharge piping. The relative

RCIC System B 3.5.3 Monticello B 3.5.3-2 Revision No. 47 BASES BACKGROUND (continued) height of the feedwater line connection for RCIC is such that the water in the feedwater lines keeps the remaining portion of the RCIC discharge line full of water. Therefore, RCIC does not require a "keep fill" system.

APPLICABLE The function of the RCIC System is to respond to transient events by SAFETY providing makeup coolant to the reactor. The RCIC System is not an ANALYSES Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. The RCIC System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The OPERABILITY of the RCIC System provides adequate core cooling such that actuation of any of the low pressure ECCS subsystems is not required in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC System has sufficient capacity for maintaining RPV inventory during an isolation event. Management of gas voids is important to RCIC System OPERABILITY (Ref. 3).

APPLICABILITY The RCIC System is required to be OPERABLE during MODE 1, and MODES 2 and 3 with reactor steam dome pressure > 150 psig, since RCIC is the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure 150 psig, the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV. Inand in MODES 4 and 5, RCIC is not required to be OPERABLE since RPV water inventory control is required by LCO 3.5.2, RPV Water Level Inventory Control.the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable RCIC System. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable RCIC System and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 and A.2 If the RCIC System is inoperable during MODE 1, or MODE 2 or 3 with reactor steam dome pressure > 150 psig, and the HPCI System is immediately verified to be OPERABLE, the RCIC System must be

PCIVs B 3.6.1.3 Monticello B 3.6.1.3-3 Revision No. 41 BASES LCO PCIVs form a part of the primary containment boundary. The PCIV safety function is related to establishing the primary containment boundary during a DBA and minimizing the loss of reactor coolant inventory. The valves covered by this LCO are listed with their associated primary containment penetrations in Reference 1.

Although not listed in Reference 1, some manual valves are also part of the containment boundary and must be controlled as PCIVs. For example, manual valves for leak rate test connections, vent paths, and drain paths inboard and between PCIVs are part of the containment boundary.

The power operated, automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal.

These valves are listed with their associated stroke times in Reference 7.

The 18 inch purge and vent valves must be maintained blocked to prevent full opening. While the reactor building-to-suppression chamber vacuum breakers isolate primary containment penetrations, they are excluded from this Specification. Controls on their isolation function are adequately addressed in LCO 3.6.1.6, "Reactor Building-to-Suppression Chamber Vacuum Breakers."

The normally closed manual PCIVs are considered OPERABLE when valves are closed or open in accordance with appropriate administrative controls, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves and devices are those listed in Reference 1.

Purge and vent valves with resilient seals must meet leakage rate requirements consistent with Type C testing requirements. Other PCIV leakage rates are addressed by LCO 3.6.1.1, "Primary Containment," as Type B or C testing.

This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, most PCIVs are not required to be OPERABLE in MODES 4 and 5. Certain valves (i.e.,

residual heat removal (RHR) shutdown cooling supply isolation valves),

however, are required to be OPERABLE when to prevent inadvertent reactor vessel draindown. These valves are those whose the associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary

PCIVs B 3.6.1.3 Monticello B 3.6.1.3-9 Revision No. 41 BASES ACTIONS (continued)

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration, the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown, and the relative importance of MSIV or main steam pathway leakage to the overall containment function.

F.1 and F.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

G.1 and G.2 If any Required Action and associated Completion Time cannot be met for PCIV(s) required OPERABLE in MODE 4 or 5, the unit must be placed in a condition in which the LCO does not apply. Action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. If suspending an OPDRV would result in closing the RHR shutdown cooling supply isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valve(s) to OPERABLE status. This allows RHR shutdown cooling to remain in service while actions are being taken to restore the valve.

SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR ensures that the 18 inch primary containment purge and vent valves are closed as required or, if open, open for an allowable reason. If a purge or vent valve is open in violation of this SR, the valve is considered inoperable. If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of limits. The SR is modified by a Note stating that the SR is not required to be met when the purge and vent valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

Suppression Pool Water Level B 3.6.2.2 Monticello B 3.6.2.2-2 Revision No. 0 BASES LCO A limit that suppression pool water level be - 4.0 inches and

+ 3.0 inches is required to ensure that the primary containment conditions assumed for the safety analyses are met. Either the high or low water level limits were used in the safety analyses, depending upon which is more conservative for a particular calculation.

APPLICABILITY In MODES 1, 2, and 3, a DBA would cause significant loads on the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. The requirements for maintaining suppression pool water level within limits in MODE 4 or 5 is addressed in LCO 3.5.2, "RPV Water Inventory ControlECCS -

Shutdown."

ACTIONS A.1 With suppression pool water level outside the limits, the conditions assumed for the safety analyses are not met. If water level is below the minimum level, the pressure suppression function still exists as long as downcomer lines are covered, HPCI and RCIC turbine exhausts are covered, and S/RV quenchers are covered. If suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and the capability of the Drywell Spray System. Therefore, continued operation for a limited time is allowed. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient to restore suppression pool water level to within limits. Also, it takes into account the low probability of an event impacting the suppression pool water level occurring during this interval.

B.1 and B.2 If suppression pool water level cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Secondary Containment B 3.6.4.1 Monticello B 3.6.4.1-2 Revision No. 42 BASES LCO An OPERABLE secondary containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be diluted and processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

B.1 and B.2 If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Secondary Containment B 3.6.4.1 Monticello B 3.6.4.1-3 Revision No. 42 BASES ACTIONS (continued)

C.1 and C.2 Movement of recently irradiated fuel assemblies in the secondary containment and OPDRVs can be postulated to cause significant fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release.

Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR was developed based on operating experience related to secondary containment vacuum variations during the applicable MODES and the low probability of a DBA occurring.

Furthermore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal secondary containment vacuum condition.

SR 3.6.4.1.2 and SR 3.6.4.1.3 Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary

SCIVs B 3.6.4.2 Monticello B 3.6.4.2-2 Revision No. 42 BASES APPLICABLE that they can be treated by the SGT System prior to discharge to the SAFETY environment.

ANALYSES (continued)

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO SCIVs form a part of the secondary containment boundary. The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated, automatic isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke times, are listed in Reference 3.

The normally closed manual SCIVs are considered OPERABLE when the valves are closed and blind flanges in place, or open under administrative controls. These passive isolation valves or devices are listed in Reference 3.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment.

Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES.

Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3. Due to radioactive decay, SCIVs are only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated individual, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.

The second Note provides clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide

SCIVs B 3.6.4.2 Monticello B 3.6.4.2-4 Revision No. 42 BASES ACTIONS (continued) locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low.

B.1 With two SCIVs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low.

The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves. This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations.

C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1 and D.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

SGT System B 3.6.4.3 Monticello B 3.6.4.3-2 Revision No. 42 BASES BACKGROUND (continued)

The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. The SGT System is initiated by Reactor Vessel Water Level - Low Low, Drywell Pressure - High, Reactor Building Ventilation Exhaust Radiation - High, and Refueling Floor Radiation -

High signals. Following initiation, the SGT subsystem A starts and both the inlet and outlet dampers of the reactor building ventilation ducts are isolated. A failure of the SGT subsystem A to start within the required time delay will initiate the automatic start and alignment of SGT subsystem B. Automatic valves provide for isolation of each SGT subsystem. Each subsystem can draw air to remove radioactive decay heat from the charcoal adsorber.

APPLICABLE The design basis for the SGT System is to mitigate the consequences of SAFETY a loss of coolant accident and fuel handling accidents involving handling ANALYSES recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) (Refs. 3 and 4). For all events analyzed, the SGT System is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment.

The SGT System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the SGT System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

SGT System B 3.6.4.3 Monticello B 3.6.4.3-3 Revision No. 42 BASES ACTIONS A.1 With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT System and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and, C.2.1, and C.2.2 During movement of recently irradiated fuel assemblies, in the secondary containment or during OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation will occur, and that any other failure would be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

SGT System B 3.6.4.3 Monticello B 3.6.4.3-4 Revision No. 42 BASES ACTIONS (continued)

The Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

D.1 If both SGTS subsystems are inoperable in MODE 1, 2, or 3, the SGT system may not be capable of supporting the required radioactivity release control function. Therefore, actions are required to enter LCO 3.0.3 immediately.

E.1 and E.2 When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.

Also, if applicable, action must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem for 15 continuous minutes (Ref. 5) ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.

CREF System B 3.7.4 Monticello B 3.7.4-3 Revision No. 42 BASES LCO (continued)

b.

Low efficiency filter, HEPA filters, and charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functions; and Note: The in-line heater remains In the system, but the heater function requirement was removed by Amendment 181.

c.

Ductwork and dampers are OPERABLE, and air circulation can be maintained.

In order for the CREF subsystems to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs, and that CRE occupants are protected from hazardous chemicals and smoke.

The LCO is modified by a Note allowing the CRE boundary to be opened intermittently under administrative controls. This Note only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the CRE. This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for main control room envelope isolation is indicated.

APPLICABILITY In MODES 1, 2, and 3, the CREF System must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA LOCA, since the DBA LOCA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA LOCA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREF System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a.

During operations with a potential for draining the reactor vessel (OPDRVs); and

b.

Dduring movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the CREF System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

CREF System B 3.7.4 Monticello B 3.7.4-5 Revision No. 42 BASES ACTIONS (continued)

C.1 and C.2 In MODE 1, 2, or 3, if the inoperable CREF subsystem or the CRE boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

D.1, and D.2.1, and D.2.2 The Required Actions of Condition D are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, if the inoperable CREF subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CREF subsystem may be placed in the pressurization mode. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the CRE boundary. This places the unit in a condition that minimizes the accident risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions must continue until the OPDRVs are suspended.

CREF System B 3.7.4 Monticello B 3.7.4-6 Revision No. 42 BASES ACTIONS (continued)

E.1 If both CREF subsystems are inoperable in MODE 1, 2, or 3 for reasons other than an inoperable CRE boundary (i.e., Condition B), the CREF System may not be capable of performing the intended function and the unit is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

F.1 and F.2 The Required Actions of Condition F are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, with two CREF subsystems inoperable, or with one or more CREF subsystems inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the CRE boundary. This places the unit in a condition that minimizes the accident risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. If applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

Control Room Ventilation System B.3.7.5 Monticello B 3.7.5-2 Revision No. 42 BASES APPLICABLE envelope, including consideration of equipment heat loads and personnel SAFETY occupancy requirements to ensure equipment OPERABILITY.

ANALYSES (continued)

The Control Room Ventilation System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant subsystems of the Control Room Ventilation System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem.

Total system failure could result in the equipment operating temperature exceeding limits.

The Control Room Ventilation System is considered OPERABLE when the individual components necessary to maintain the control room envelope temperature are OPERABLE in both subsystems. These components include the cooling coils, fans, compressors, ductwork, dampers, and associated instrumentation and controls.

As described in LCO 3.7.4, an OPERABLE exhaust/recirculation fan (V-ERF-14A or B) and air handling unit (V-EAC-14A or B) (excluding the condenser unit), from the same subsystem, are required to provide flow to support OPERABILITY of the associated CREF subsystem.

APPLICABILITY In MODE 1, 2, or 3, the Control Room Ventilation System must be OPERABLE to ensure that the control room envelope temperature will not exceed equipment OPERABILITY limits following control room envelope boundary isolation.

In MODES 4 and 5, the probability and consequences of a DBA LOCA (Ref. 3) are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room Ventilation System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a.

During operations with a potential for draining the reactor vessel (OPDRVs); and Dduring movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the Control Room Ventilation System is only required to be OPERABLE during fuel handling involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Control Room Ventilation System B.3.7.5 Monticello B 3.7.5-4 Revision No. 42 BASES ACTIONS (continued)

D.1, and D.2.1, and D.2.2 The Required Actions of Condition D are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies, although not feasible, while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, if Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE control room ventilation subsystem may be placed immediately in operation.

This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

E.1 and E.2 The Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies, although not feasible, while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, if Required Actions B.1 and B.2 cannot be met within the required Completion Times, action must be taken to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

Control Room Ventilation System B.3.7.5 Monticello B 3.7.5 Last Revision No. 42 BASES ACTIONS (continued)

If applicable, handling of recently irradiated fuel in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room envelope heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation.

The 24 month Frequency is appropriate since significant degradation of the Control Room Ventilation System is not expected over this time period.

REFERENCES

1.

USAR, Section 6.7.

2.

Amendment No. 154, Issuance of Amendment Re: Two Inoperable Control Room Ventilation Subsystems Using the Guidance of TSTF-477, dated January 23, 2008.

3.

USAR, Section 14.7.2.

4.

USAR, Section 14.7.6.

AC Sources - Shutdown B 3.8.2 Monticello B 3.8.2-1 Revision No. 27 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources - Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources - Operating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 and 5 SAFETY and during movement of recently irradiated fuel assemblies in the ANALYSES secondary containment ensures that:

a.

The facility can be maintained in the shutdown or refueling condition for extended periods;

b.

Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c.

Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving handling recently irradiated fuel. Due to radioactive decay, AC electrical power is only required to be OPERABLE during fuel handling involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

In general, when the unit is shut down the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS.

This allowance is in recognition that certain testing and maintenance activities must be conducted, provided an acceptable level of risk is not exceeded. During MODES 4 and 5, performance of a significant number of required testing and maintenance activities is also required. In

AC Sources - Shutdown B 3.8.2 Monticello B 3.8.2-2 Revision No. 27 BASES APPLICABLE SAFETY ANALYSES (continued)

MODES 4 and 5, the activities are generally planned and administratively controlled. Relaxations from typical MODES 1, 2, and 3 LCO requirements are acceptable during shutdown MODES, based on:

a.

The fact that time in an outage is limited. This is a risk prudent goal as well as a utility economic consideration.

b.

Requiring appropriate compensatory measures for certain conditions.

These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operation MODE analyses, or both.

c.

Prudent utility consideration of the risk associated with multiple activities that could affect multiple systems.

d.

Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODES 1, 2, and 3 OPERABILITY requirements) with systems assumed to function during an event.

In the event of an accident during shutdown, this LCO ensures the capability of supporting systems necessary for avoiding immediate difficulty, assuming either a loss of all offsite power or a loss of all onsite (emergency diesel generator (EDG)) power.

AC Sources - Shutdown satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO One offsite circuit capable of supplying the onsite Class 1E power distribution subsystem(s) of LCO 3.8.8, "Distribution Systems -

Shutdown," ensures that all required loads are powered from offsite power. An OPERABLE EDG, associated with a Distribution System 4.16 kV essential bus required OPERABLE by LCO 3.8.8, ensures that a diverse power source is available for providing electrical power support assuming a loss of the offsite circuit. Together, OPERABILITY of the required offsite circuit and EDG ensures the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents involving handling recently irradiated fuel and inadvertent reactor vessel draindown). Automatic initiation of the required EDG during shutdown conditions is specified in LCO 3.3.5.1, "ECCS Instrumentation," and LCO 3.3.8.1, "LOP Instrumentation."

The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective 4.16 kV essential bus, and of accepting required loads during an accident. The primary AC electrical power distribution subsystem for each division

AC Sources - Shutdown B 3.8.2 Monticello B 3.8.2-4 Revision No. 27 BASES APPLICABILITY The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment to provide assurance that:

a.

Systems that provide core cooling providing adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b.

Systems needed to mitigate a fuel handling accident involving handling recently irradiated fuel are available;

c.

Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d.

Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

AC power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.1.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement should be accounted for in MODE 1, 2, or 3, although not feasible, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies, although not feasible, while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3, while in MODE 1, 2, or 3 would require the unit to be shutdown unnecessarily.

A.1 An offsite circuit is considered inoperable if it is not available to one required division. If two or more 4.16 kV essential buses are required per LCO 3.8.8, one division with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, and recently irradiated fuel movement, and operations with a potential for draining the reactor vessel. By the allowance of the option to declare required features inoperable with no offsite power available, appropriate restrictions can be implemented in accordance with the affected required feature(s) LCOs' ACTIONS.

A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, B.3, and B.34 With the offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable. Since this option may

AC Sources - Shutdown B 3.8.2 Monticello B 3.8.2-5 Revision No. 27 BASES ACTIONS (continued) involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With the required EDG inoperable, the minimum required diversity of AC power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, and movement of recently irradiated fuel assemblies in the secondary containment, and activities that could result in inadvertent draining of the reactor vessel.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC source and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power source should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in de-energization. Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any required 4.16 kV essential bus, ACTIONS for LCO 3.8.8 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized. LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division.

SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, and 3. SR 3.8.1.6 is not required to be met since only one offsite circuit is required to be OPERABLE. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

This SR is modified by two Notes. The reason for Note 1 is to preclude requiring the OPERABLE required EDG from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and to preclude deenergizing a required 4.16 kV essential bus or disconnecting a required offsite circuit during performance of SRs. With limited AC sources available, a single event could compromise both the required circuit and the EDG. It is the intent that these SRs must still be capable of being met, but actual performance

AC Sources - Shutdown B 3.8.2 Monticello B 3.8.2 Last Revision No. 27 BASES SURVEILLANCE REQUIREMENTS (continued) is not required during periods when the EDG and offsite circuit is required to be OPERABLE. Note 2 states that SRs 3.8.1.8 and 3.8.1.12 are not required to be met when its associated ECCS subsystem(s) are not required to be OPERABLE. These SRs demonstrate the EDG response to an ECCS signal (either alone or in conjunction with a loss-of-power signal). This is consistent with the ECCS instrumentation requirements that do not require the ECCS signals when the ECCS System is not required to be OPERABLE. per LCO 3.5.2, "ECCS - Shutdown."

REFERENCES

1.

USAR, Section 8.2

DC Sources - Shutdown B 3.8.5 Monticello B 3.8.5-1 Revision No. 4 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources - Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources - Operating."

APPLICABLE The initial conditions of Design Basis Accident and transient analyses in SAFETY USAR, Chapter 14 (Ref. 1), assume that Engineered Safety Feature ANALYSES systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the emergency diesel generators (EDGs), emergency auxiliaries, and control and switching during all MODES of operation.

The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of recently irradiated fuel assemblies ensures that:

a.

The facility can be maintained in the shutdown or refueling condition for extended periods;

b.

Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c.

Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving handling recently irradiated fuel. Due to radioactive decay, DC electrical sources are only required to be OPERABLE during fuel handling involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

In general, when the unit is shut down, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or all onsite power is not required.

The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence being significantly reduced or eliminated, and in minimal

DC Sources - Shutdown B 3.8.5 Monticello B 3.8.5-2 Revision No. 4 BASES APPLICABLE SAFETY ANALYSES (continued) consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

The shutdown Technical Specification requirements are designed to ensure that the unit has the capability to mitigate the consequences of certain postulated accidents. Worst case DBAs which are analyzed for operating MODES are generally viewed not to be a significant concern during shutdown MODES due to the lower energies involved. The Technical Specifications therefore require a lesser complement of electrical equipment to be available during shutdown than is required during operating MODES. More recent work completed on the potential risks associated with shutdown, however, have found significant risk associated with certain shutdown evolutions. As a result, in addition to the requirements established in the Technical Specifications, the industry has adopted NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," as an Industry initiative to manage shutdown tasks and associated electrical support to maintain risk at an acceptable low level. This may require the availability of additional equipment beyond that required by the shutdown Technical Specifications.

The DC Sources - Shutdown satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The Division 1 or Division 2 125 VDC electrical power subsystem consisting of one 125 V battery, one battery charger, and the corresponding control equipment and interconnecting cabling is required to be OPERABLE to support one division of the DC electrical power distribution subsystem(s) required OPERABLE by LCO 3.8.8, "Distribution Systems - Shutdown." This requirement ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents involving handling recently irradiated fuel and inadvertent reactor vessel draindown).

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment provide assurance that:

a.

Required features to provide core coolingadequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b.

Required features needed to mitigate a fuel handling accident involving handling recently irradiated fuel are available;

DC Sources - Shutdown B 3.8.5 Monticello B 3.8.5-3 Revision No. 4 BASES APPLICABILITY (continued)

c.

Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d.

Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement should be accounted for in MODE 1, 2, or 3, although not feasible, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, although not feasible, the fuel movement is independent of reactor operations. Entering LCO 3.0.3, while in MODE 1, 2, or 3 would require the unit to be shutdown unnecessarily.

A.1, A.2, and A.3, and A.4 If the required Division 1 or Division 2 125 VDC electrical power subsystem is inoperable, the minimum required DC power sources are not available. Therefore, suspension of CORE ALTERATIONS and, movement of recently irradiated fuel assemblies in the secondary containment, and any activities that could result in inadvertent draining of the reactor vessel is required.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystem and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystem should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

Distribution Systems - Shutdown B 3.8.8 Monticello B 3.8.8-1 Revision No. 4 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems - Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution system is provided in the Bases for LCO 3.8.7, "Distribution Systems - Operating."

APPLICABLE The initial conditions of Design Basis Accident and transient analyses in SAFETY USAR, Chapter 14 (Ref. 1), assume Engineered Safety Feature (ESF)

ANALYSES systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Emergency Core Cooling Systems and Reactor Core Isolation Cooling System, and containment design limits are not exceeded.

The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5, and during movement of recently irradiated fuel assemblies in the secondary containment ensures that:

a.

The facility can be maintained in the shutdown or refueling condition for extended periods;

b.

Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c.

Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving handling recently irradiated fuel. Due to radioactive decay, AC and DC electrical power distribution subsystems are only required to be OPERABLE during fuel handling involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

The Distribution Systems - Shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support required features. This LCO explicitly requires energization of the portions of the electrical distribution system necessary to support OPERABILITY of Technical Specifications required systems,

Distribution Systems - Shutdown B 3.8.8 Monticello B 3.8.8-2 Revision No. 4 BASES LCO (continued) equipment, and components - both specifically addressed by their own LCO, and implicitly required by the definition of OPERABILITY.

Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents involving handling recently irradiated fuel and inadvertent reactor vessel draindown).

APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment provide assurance that:

a.

Systems that provide core cooling to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b.

Systems needed to mitigate a fuel handling accident involving handling recently irradiated fuel are available;

c.

Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d.

Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC and DC electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8.7.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement should be accounted for in MODE 1, 2, or 3, although not feasible, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, although not feasible, the fuel movement is independent of reactor operations. Entering LCO 3.0.3, while in MODE 1, 2, or 3 would require the unit to be shutdown unnecessarily.

A.1, A.2.1, A.2.2, A.2.3, A.2.4, and A.2.45 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting

Distribution Systems - Shutdown B 3.8.8 Monticello B 3.8.8-3 Revision No. 4 BASES ACTIONS (continued) sufficient required features to allow continuation of CORE ALTERATIONS and, recently irradiated fuel movement, and operations with a potential for draining the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances this option may involve undesired administrative efforts.

Therefore, the allowance for sufficiently conservative actions is made, (i.e., to suspend CORE ALTERATIONS and, movement of recently irradiated fuel assemblies in the secondary containment, and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.

Notwithstanding performance of the above conservative Required Actions, a required residual heat removal-shutdown cooling (RHR-SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS would not be entered. Therefore, Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable, which results in taking the appropriate RHR-SDC ACTIONS.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.

SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystems are functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The 7 day Frequency takes into account the redundant capability of the electrical power distribution subsystems, as well as other indications available in the control room that alert the operator to subsystem malfunctions.

REFERENCES

1.

USAR, Chapter 14.

Inservice Leak and Hydrostatic Testing Operation B 3.10.1 Monticello B 3.10.1-2 Revision No. 28 BASES BACKGROUND (continued)

The hydrostatic and/or RCS system leakage tests require increasing pressure to approximately 1000 psig. Since scram time testing required by SR 3.1.4.1 and SR 3.1.4.4 requires reactor steam dome pressure 800 psig.

Other testing may be performed in conjunction with the allowances for inservice leak or hydrostatic tests and control rod scram time tests.

APPLICABLE Allowing the reactor to be considered in MODE 4, when the reactor SAFETY coolant temperature is > 212°F, during, or as a consequence of, ANALYSES hydrostatic or leak testing, or as a consequence of control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test, effectively provides an exception to MODE 3 requirements, including OPERABILITY of primary containment and the full complement of redundant Emergency Core Cooling Systems. Since the tests are performed nearly water solid, at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be very low.

Under these conditions, the potential for failed fuel and a subsequent increase in coolant activity above the LCO 3.4.6, "RCS Specific Activity,"

limits are minimized. In addition, the secondary containment will be OPERABLE, in accordance with this Special Operations LCO, and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. Furthermore, the specific activity of the reactor coolant is assumed to be 0.02 µCi/gm DOSE EQUIVALENT I-131. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment described in Reference 2. Therefore, these requirements will conservatively limit radiation releases to the environment.

In the unlikely event of a largeany primary system leak that could result in draining of the RPV, the reactor vessel would rapidly depressurize, allowing the low pressure core cooling systems to operate. The make-up capability of the low pressure coolant injection and core spray subsystems, as required in MODE 4 by LCO 3.5.2, "RPV Water Inventory ControlECCS -

Shutdown," would be more than adequate to keep the RPV water level above the TAFcore flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred.

For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the reactor coolant specific activity limit and secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences

v t e Letter Sequence Request
EPID:L-2017-LLA-0360, TSTF-542 (Approved, Closed)
Initiation Request Acceptance Supplement, Supplement Administration Questions 9 May 2018 Answers Results Approval Other: L-MT-18-031, Project Plan Progress Update Toward Restoring 10 CFR 72 Compliance to Dry Shielded Canisters Designated 11 Through 16, ML19052A142
MONTHYEAR2018-05-09 [Table View]

L-MT-17-058, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control

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L-MT-17-058, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control

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1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 REGULATORY ANALYSIS

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