Text

Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.
	ML18057B125
Person / Time
Site:	Quad Cities
Issue date:	02/26/2018
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-18-006
	Download: ML18057B125 (215)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-18-006 10 CFR 50.90 February 26, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (EGC) is submitting a request for an amendment to the Technical Specifications (TS) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2.

The proposed change replaces existing 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 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel (TAF). 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.

The proposed change has been reviewed and recommended for approval by the QCNPS Plant Operations Review Committee in accordance with the EGC Quality Assurance Program.

Approval of the proposed amendment is requested by February 26, 2019, in order to support the QCNPS, Unit 1 Refueling Outage that is planned for March 2019 (i.e., Q1R25). Once approved, the amendment shall be implemented prior to initial entry into Mode 4 in preparation for refueling activities during Q1R25.

EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b).

February 26, 2018 U.S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 26th day of February 2018.

Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments: 1. Description and Assessment

2. Proposed Technical Specifications Changes (Mark-Up)

3. Revised Technical Specifications Pages

4. Proposed Technical Specifications Bases Changes (Mark-Up) cc: NRC Regional Administrator, Region 111 NRC Senior Resident Inspector- Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

Quad Cities Nuclear Power Station, Units 1 and 2, Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control"

1.0 DESCRIPTION

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation 2.2 Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration

4.0 ENVIRONMENTAL CONSIDERATION

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

Exelon Generation Company, LLC (EGC) proposes a change to the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 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 2.1.1.3.

Safety Limit 2.1.1.3 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 EGC has reviewed the safety evaluation provided to the Technical Specifications Task Force on December 20, 2016, as well as the information provided in TSTF-542. EGC has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the NRC are applicable to QCNPS, Units 1 and 2 and justify this amendment for the incorporation of the changes to the QCNPS TS.

The following QCNPS, Units 1 and 2 TS reference or are related to OPDRVs and are affected by the proposed change:

1.1, Definitions 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation 3.3.6.1, Primary Containment Isolation Instrumentation 3.3.6.2, Secondary Containment Isolation Instrumentation 3.3.7.1, Control Room Emergency Ventilation (CREV) System Instrumentation 3.5.1, ECCS - Operating 3.5.2, ECCS - Shutdown 3.5.3, RCIC System 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 Ventilation (CREV) System 3.7.5, Control Room Emergency Ventilation Air Conditioning (AC) System 3.8.2, AC Sources - Shutdown 3.8.5, DC Sources - Shutdown 3.8.8, Distribution Systems - Shutdown 2.2 Variations EGC is proposing the following variations from the TS changes described in the TSTF-542 or the applicable parts of the NRCs safety evaluation. These variations do not affect the applicability of TSTF-542 or the NRC's safety evaluation to the proposed license amendment.

Page 1 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT 2.2.1. In a few instances, the QCNPS TS utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-542 was based. These differences are administrative and do not affect the applicability of TSTF-542 to the QCNPS, Units 1 and 2 TS. Specifically, the titles for the following QCNPS, Units 1 and 2 TS vary from the STS discussed in TSTF-542:

3.3.7.1, Control Room Emergency Ventilation (CREV) System Instrumentation 3.7.4, Control Room Emergency Ventilation (CREV) System 3.7.5, Control Room Emergency Ventilation Air Conditioning (AC) System 3.8.8, Distribution Systems - Shutdown 2.2.2. The TSTF-542 Traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). QCNPS, Units 1 and 2 were not licensed to the 10 CFR 50, Appendix A, GDC. The QCNPS, Units 1 and 2 Updated Final Safety Analysis Report (UFSAR),

Section 3.1, "Conformance with NRC General Design Criteria," contains an evaluation of the design basis of QCNPS with respect to the first draft of the 70 proposed "General Design Criteria for Nuclear Power Plant Construction Permits" issued by the Atomic Energy Commission in July 1967. This difference does not alter the conclusion that the proposed change is applicable to QCNPS.

2.2.3. The QCNPS TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement Frequencies for Technical Specifications 3.3.5.2 and 3.5.2 are "In accordance with the Surveillance Frequency Control Program."

2.2.4. A previously identified issue related to TSTF-542 applies to this request to adopt the TSTF for QCNPS, Units 1 and 2. Specifically, in accordance with TSTF-542, TS Table 3.3.5.2-1, Function 1.a (Core Spray System Reactor Steam Dome Pressure - Low (Injection Permissive), and Function 2.a (Low Pressure Coolant Injection (LPCI) Reactor Steam Dome Pressure - Low (Injection Permissive) are required in Modes 4 and 5.

Prior to TSTF-542, the analogous Functions 1.c and 2.c in TS Table 3.3.5.1-1 had a Mode 4 and 5 applicability modified by a footnote specifying that these functions were only required when the associated emergency core cooling system (ECCS) subsystem(s) were required to be operable per limiting condition for operation (LCO) 3.5.2, "ECCS Shutdown." The footnote was inadvertently omitted from Table 3.3.5.2-1 Functions 1.a and 2.a in TSTF-542. Without the footnote, the Reactor Steam Dome Pressure - Low functions would be required to be operable for all low pressure ECCS subsystems, regardless of whether they are credited for meeting LCO 3.5.2. Requiring the functions for all ECCS subsystems is unnecessary. In Modes 4 and 5 with the reactor steam dome at atmospheric pressure, these functions only serve to satisfy permissives for opening low pressure ECCS injection valves for manual actuation.

Accordingly, a variation is proposed to affix Footnote (a) (i.e., "Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, Reactor Pressure Vessel Water Inventory Control") to the "Required Channels Per Function" column of Functions 1.a and 2.a of TS Table 3.3.5.2-1.

Page 2 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT 2.2.5. TSTF-542 inadvertently omitted the TS Bases markup corresponding to the deletion of the TS 3.3.6.1, "Primary Containment Isolation Instrumentation," action to isolate the residual heat removal shutdown cooling system (i.e., currently TS 3.3.6.1 Required Action I.2 for QCNPS, Units 1 and 2). The corresponding TS Bases changes are marked up in Attachment 4 to align with this proposed TS change.

2.2.6. The QCNPS TS contain the following requirements that differ from the Standard Technical Specifications on which TSTF-542 was based, but are encompassed in the TSTF-542 justification:

a. There are QCNPS specific instrumentation functions that differ from the Standard Technical Specifications (STS). Changes to these instrumentation functions are justified by the discussion in Section 3.4.1 of the TSTF-542 justification. QCNPS TS Table 3.3.5.1-1, Functions 1.e and 2.e describe pump start time delay relays for the core spray (CS) and low pressure coolant injection (LPCI) pumps. The purpose of these time delays is to stagger the automatic start of CS and LPCI pumps that are in Division 1 and 2, thus limiting the starting transients on the 4.16 kV emergency buses. This staggering is unnecessary for manual operation; therefore, these Functions can be removed from the TS because the required ECCS subsystem is proposed to be started by manual operation.

b. QCNPS, Units 1 and 2 do not currently have the capability to perform Channel Checks for proposed Table 3.3.5.2-1 Functions 1.a, "Reactor Steam Dome PressureLow (Permissive), 1.b, "Core Spray Pump Discharge FlowLow (Bypass)," 2.a, "Reactor Steam Dome PressureLow (Permissive)," and 2.b, "Low Pressure Coolant Injection Pump Discharge FlowLow (Bypass)." The current QCNPS, Units 1 and 2 TS do not include Channel Checks for these functions; therefore, no Channel Check Surveillance Requirement (SR) was added for these functions.

c. QCNPS, Units 1 and 2 each have two LPCI loops with two pumps in each loop (i.e.,

four LPCI pumps per unit). The QCNPS TS currently require one operable channel for each LPCI loop for the "Low Pressure Coolant Injection Pump Discharge Flow Low (Bypass)," Function. EGC proposes to maintain the one channel per loop for Technical Specification 3.3.5.2, Function 2.b in lieu of one channel per pump as described in the STS, since at QCNPS, there is only one flow transmitter for each loop that monitors the flow of both pumps in that loop (i.e., two LPCI loop flow transmitters per unit).

d. The current QCNPS, Units 1 and 2 TS do not include a manual initiation logic function for the CS or LPCI subsystems. Therefore, since this function does not exist at QCNPS, manual initiation functions for LPCI and CS are not being included in Technical Specification 3.3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control," Table 3.3.5.2-1. Additionally, since the manual initiation functions are not included in Table 3.3.5.2-1, the associated Logic System Functional Test would likewise not be required for TS 3.3.5.2; therefore, TS 3.3.5.2 as proposed for QCNPS, Units 1 and 2 does not include a Logic System Functional Test SR.

Page 3 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

e. As an alternative, EGC proposes that Technical Specification 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control," include an SR to verify that the QCNPS, Units 1 and 2 LPCI and CS subsystem can be manually operated (i.e.,

proposed SR 3.5.2.7 shown in Figure 1 below).

SR 3.5.2 . 1 ------------------NOTt-----------------

Vessel injection/spray may be excluded .

Verify the required ECCS injection/spray In accordance subsystem can be manually operated . with the Surve i 7 7ance Frequency Control Program Figure 1: Proposed SR 3.5.2. 7 The manual operation of the LPCI and CS 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 Main 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 Technical Specification 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. Consequently, there is no need for manual initiation logic to actuate the required subsystem components.

Since the LPCI and CS subsystems can be placed in service using manual means in a short period of time (i.e., within the timeframes assumed in the development of TSTF-542), using controls and indications that are readily available in the Main Control Room, manual operation of the required subsystem would be an equivalent alternative to system initiation via manual initiation logic.

f. The QCNPS design provides for isolation of the reactor water cleanup (RWCU) system at Reactor Vessel Water Level-Low versus Reactor Vessel Water Level-Low Low; therefore, proposed Table 3.3.5.2-1 reflects this RWCU isolation difference.

g. The automatic isolation on Reactor Vessel Water Level-Low functions for the RWCU and residual heat removal shutdown cooling (RHR SOC) systems at QCNPS, Units 1 and 2 varies slightly from the system described in the STS. These functions receive input from four reactor vessel water level channels. Each channel inputs into one of four trip strings, and two trip strings make up a trip system. The trip systems are aligned in a parallel configuration, so both trip systems must trip in order to cause Page 4 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT an isolation of the RWCU or RHR SDC system valves. Any channel will trip its associated trip string and trip system. Therefore, both trip systems with one trip string in each trip system is required to provide for automatic RWCU and RHR SDC system isolation. Proposed Table 3.3.5.2-1 has been revised to reflect the QCNPS, Units 1 and 2 requirement for one operable channel in each Reactor Vessel Water LevelLow isolation trip system for RWCU and RHR SDC.

h. The QCNPS Control Room Emergency Ventilation (CREV) and the Control Room Emergency Ventilation Air Conditioning (AC) systems (i.e., Technical Specifications 3.7.4 and 3.7.5, respectively) provide Control Room habitability functions. Changes to the TS controls on these systems is justified by the discussion in Section 3.4.3 of the TSTF-542 justification. Specifically, these QCNPS specific systems provide similar Control Room habitability functions as those described in the STS, and changes to these Technical Specifications are similarly justified.

2.2.7. The QCNPS TS do not currently contain a Note applicable to LCO 3.5.2 that allows a Low Pressure Coolant Injection (LPCI) subsystem to be considered operable when aligned for decay heat removal. Instead, this Note is currently associated with existing SR 3.5.2.3 (i.e., proposed SR 3.5.2.4). The proposed QCNPS, Units 1 and 2 LCO 3.5.2 includes this Note to align with the STS. This is a minor variation, as the purpose of the Note is the same as the one described in the STS and the Note is applicable to QCNPS, Units 1 and 2.

2.2.8. At QCNPS, verification of Suppression Pool and Contaminated Condensate Storage Tank volumes is contained in a single SR (i.e., existing SR 3.5.2.1 and proposed SR 3.5.2.2) versus in separate SRs as found in the STS. This does not affect the applicability of TSTF-542.

2.2.9. EGC proposes to delete QCNPS, Units 1 and 2 TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs);" Condition F and all of its associated Required Actions as shown in Figure 2 below. The Applicability for TS 3.6.1.3 is Modes 1, 2, and 3, and When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation." This change is justified since OPDRV requirements have been deleted, and Mode 4 and 5 PCIV requirements have been relocated from TS 3.3.6.1 and 3.6.1.3 to the proposed TS 3.3.5.2. Thus, there are no longer any PCIVs required to be operable by TS 3.6.1.3 during Mode 4 or 5. These requirements are addressed by the proposed TS 3.3.5.2 in their entirety. Following the removal of OPDRV and relocation of Mode 4 and 5 requirements as discussed above, this Condition and associated Actions in TS 3.6.1.3 would never be applicable; therefore, are no longer necessary.

Page 5 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT PCI Vs 3 .6 . 1.3 AbTIQNa CQ~IQ I TI Q~I CQMPbUIQ~I TB<I[

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I Figure 2: Proposed Variation That Deletes Technical Specification 3.6.1.3 Condition F

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Exelon Generation Company, LLC (EGC) requests adoption of Technical Specifications Task Force Traveler (TSTF)-542 "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications (STS), into the Quad Cities Nuclear Power Station, Units 1 and 2 Technical 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 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) 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 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate Page 6 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT 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 verification of a second method of water injection and additional confirmations that secondary 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 2.1.1.3. 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 Page 7 of 8

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT 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 safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel 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, EGC 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 CONSIDERATION

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.

Page 8 of 8

Quad Cities Nuclear Power Station, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 2 - PROPOSED TECHNICAL SPECIFICATIONS CHANGES (MARK-UP)

TOC Page i 3.3.5.2-3 3.6.1.3-4 TOC Page ii 3.3.5.3-1 3.6.1.3-5 1.1-1 3.3.5.3-2 3.6.4.1-1 1.1-2 3.3.5.3-3 3.6.4.2-1 1.1-3 3.3.5.3-4 3.6.4.2-3 1.1-4 3.3.6.1-3 3.6.4.3-1 1.1-5 3.3.6.1-8 3.6.4.3-2 1.1-6 3.3.6.2-5 3.7.4-1 3.3.5.1-2 3.3.7.1-5 3.7.4-2 3.3.5.1-3 3.5.1-1 3.7.5-1 3.3.5.1-5 3.5.2-1 3.8.2-3 3.3.5.1-10 3.5.2-2 3.8.2-4 3.3.5.1-11 3.5.2-3 3.8.2-5 3.3.5.1-12 3.5.2-4 3.8.5-2 3.3.5.1-13 3.5.2-5 3.8.8-2 3.3.5.2-1 3.5.2-6 3.3.5.2-2 3.5.3-1

TABLE OF CONTENTS 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) 2.1 SLs......................................................2.0-1 2.2 SL Violations............................................2.0-1 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) ......................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 3.3.1.3 Oscillation Power Range Monitor (OPRM) Instrumentation...3.3.1.3-1 3.3.2.1 Control Rod Block Instrumentation........................3.3.2.1-1 3.3.2.2 Feedwater System 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 3.3.4.1 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 Pressure Vessel (RPV) Water Inventory Control Instrumentation.......................................3.3.5.2-1 3.3.5.32 Reactor Core Isolation Cooling (RCIC) System Instrumentation..................................... 3.3.5.32-1 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 Relief Valve Instrumentation.............................3.3.6.3-1 (continued)

Quad Cities 1 and 2 i Amendment No. 227/222

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation........................................3.3.7.1-1 3.3.7.2 Mechanical Vacuum Pump Trip 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 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 and Relief Valves ................................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 SystemHot Shutdown....................................3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling SystemCold 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 SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.1 ECCSOperating...........................................3.5.1-1 3.5.2 RPV Water Inventory ControlECCSShutdown.................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 Pressure. .......................................3.6.1.4-1 3.6.1.5 Drywell Air Temperature..................................3.6.1.5-1 3.6.1.6 Low Set Relief Valves....................................3.6.1.6-1 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers...............................................3.6.1.7-1 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers...........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 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray.......3.6.2.4-1 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure.....3.6.2.5-1 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 (continued)

Quad Cities 1 and 2 ii Amendment No. 199/195

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific HEAT GENERATION RATE planar height and is equal to the sum of the (APLHGR) LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

(continued)

Quad Cities 1 and 2 1.1-1 Amendment No. 199/195

Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

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 REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. 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 the inhalation committed dose conversion factors in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

(continued)

Quad Cities 1 and 2 1.1-2 Amendment No. 233/229

Definitions 1.1 1.1 Definitions (continued)

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 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 is 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.

(continued)

Quad Cities 1 and 2 1.1-3 Amendment No. 266/261

Definitions 1.1 1.1 Definitions DRAIN TIME c. The penetration flow paths required to be (continued) 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).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and (continued)

Quad Cities 1 and 2 1.1-4 Amendment No. 202/198

Definitions 1.1 1.1 Definitions LEAKAGE d. Pressure Boundary LEAKAGE (continued)

LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

(continued)

Quad Cities 1 and 2 1.1-5 Amendment No. 254/249

Definitions 1.1 1.1 Definitions (continued)

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2957 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from the opening of the sensor contact until the TIME opening of the trip actuator. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is > 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME shall be RESPONSE TIME that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

Quad Cities 1 and 2 1.1-6 Amendment No. 248/243

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. As required by B.1 --------NOTES--------

Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.

2. Only applicable for Functions 1.a, 1.b, 2.a, 2.b, 2.d, and 2.j.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND B.2 --------NOTE---------

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

Declare High Pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Coolant Injection discovery of (HPCI) System loss of HPCI inoperable. initiation capability AND B.3 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

(continued)

Quad Cities 1 and 2 3.3.5.1-2 Amendment No. 199/195

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 --------NOTES--------

Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.

2. Only applicable for Functions 1.c, 1.e, 2.c, 2.e, 2.g, 2.h, 2.i, and 2.k.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND C.2 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

(continued)

Quad Cities 1 and 2 3.3.5.1-3 Amendment No. 199/195

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 --------NOTES--------

Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.

2. Only applicable for Functions 1.d and 2.f.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. subsystems in both divisions AND E.2 Restore channel to 7 days OPERABLE status.

(continued)

Quad Cities 1 and 2 3.3.5.1-5 Amendment No. 199/195

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 4)

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

1. Core Spray System

a. Reactor Vessel Water 1, 2, 3, B SR 3.3.5.1.1 -55.2 inches 4(ab)

LevelLow Low SR 3.3.5.1.2 4(a), 5(a) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell PressureHigh 1, 2, 3 4(ab) B SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 SR 3.3.5.1.7

c. Reactor Steam Dome 1, 2, 3 2 C SR 3.3.5.1.2 306 psig and PressureLow SR 3.3.5.1.4 342 psig (Permissive) SR 3.3.5.1.7 4(a),5(a) 2 B SR 3.3.5.1.2 306 psig and SR 3.3.5.1.4 342 psig SR 3.3.5.1.7

d. Core Spray Pump 1, 2, 3, 1 per pump E SR 3.3.5.1.2 577 gpm Discharge FlowLow SR 3.3.5.1.3 and (Bypass) 4(a), 5(a) SR 3.3.5.1.6 830 gpm SR 3.3.5.1.7

e. Core Spray Pump 1, 2, 3 1 per pump C SR 3.3.5.1.6 11.4 seconds Start-Time Delay SR 3.3.5.1.7 Relay 4(a), 5(a)

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Vessel Water 1, 2, 3, 4 B SR 3.3.5.1.1 -55.2 inches LevelLow Low SR 3.3.5.1.2 4(a), 5(a) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell PressureHigh 1, 2, 3 4 B SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 SR 3.3.5.1.7

c. Reactor Steam Dome 1, 2, 3 2 C SR 3.3.5.1.2 306 psig and Pressure Low SR 3.3.5.1.4 342 psig (Permissive) SR 3.3.5.1.7 4(a), 5(a) 2 B SR 3.3.5.1.2 306 psig and SR 3.3.5.1.4 342 psig SR 3.3.5.1.7 (continued)

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

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

Quad Cities 1 and 2 3.3.5.1-10 Amendment No. 204/200

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 4)

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

2. LPCI System (continued)

d. Reactor Steam Dome 1, 2, 3 4 B SR 3.3.5.1.2 868 psig and PressureLow (Break SR 3.3.5.1.5 891 psig Detection) SR 3.3.5.1.7

e. Low Pressure Coolant 1, 2, 3, 1 per pump C SR 3.3.5.1.6 6.7 seconds Injection Pump SR 3.3.5.1.7 Start-Time Delay 4(a),5(a)

Relay Pumps B and D

f. Low Pressure 1, 2, 3, 1 per loop E SR 3.3.5.1.2 2526 gpm Coolant Injection Pump SR 3.3.5.1.3 Discharge FlowLow 4(a), 5(a) SR 3.3.5.1.6 (Bypass) SR 3.3.5.1.7

g. Recirculation Pump 1, 2, 3 4 per pump C SR 3.3.5.1.2 2.3 psid Differential SR 3.3.5.1.6 Pressure-High (Break SR 3.3.5.1.7 Detection)

h. Recirculation Riser 1, 2, 3 4 C SR 3.3.5.1.2 2.15 psid Differential SR 3.3.5.1.6 PressureHigh (Break SR 3.3.5.1.7 Detection)

i. Recirculation Pump 1, 2, 3 2 C SR 3.3.5.1.6 0.82 seconds Differential Pressure SR 3.3.5.1.7 Time DelayRelay (Break Detection)

j. Reactor Steam Dome 1, 2, 3 2 B SR 3.3.5.1.6 2.26 seconds Pressure Time Delay SR 3.3.5.1.7 Relay (Break Detection)

k. Recirculation Riser 1, 2, 3 2 C SR 3.3.5.1.6 0.82 seconds Differential Pressure SR 3.3.5.1.7 Time DelayRelay (Break Detection)

(continued)

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

Quad Cities 1 and 2 3.3.5.1-11 Amendment No. 204/200

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 4)

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

3. High Pressure Coolant Injection (HPCI) System

a. Reactor Vessel Water 1, 4 B SR 3.3.5.1.1 -55.2 inches Level -Low Low SR 3.3.5.1.2 2(bc), 3(bc) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell Pressure High 1, 4 B SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 2(bc), 3(bc) SR 3.3.5.1.7

c. Reactor Vessel Water 1, 2 C SR 3.3.5.1.2 50.34 inches Level High SR 3.3.5.1.3 2(bc), 3(bc) SR 3.3.5.1.6 SR 3.3.5.1.7

d. Contaminated 1, 2 D SR 3.3.5.1.2 598 ft Condensate Storage SR 3.3.5.1.6 1 inch Tank (CCST) Level Low 2(bc), 3(bc) SR 3.3.5.1.7

e. Suppression Pool Water 1, 2 D SR 3.3.5.1.2 15 ft Level High SR 3.3.5.1.6 11.25 inches 2(bc), 3(bc) SR 3.3.5.1.7

f. High Pressure Coolant 1, 1 E SR 3.3.5.1.2 634 gpm Injection Pump SR 3.3.5.1.4 Discharge Flow Low 2(bc), 3(bc) SR 3.3.5.1.7 (Bypass) 1, 1 C SR 3.3.5.1.7 NA

g. Manual Initiation 2(bc), 3(bc)

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 -55.2 inches LevelLow Low SR 3.3.5.1.2 2(bc), 3(bc) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell PressureHigh 1, 2 F SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 2(bc), 3(bc) SR 3.3.5.1.7

c. Automatic 1, 1 G SR 3.3.5.1.6 119 seconds Depressurization SR 3.3.5.1.7 System Initiation 2(bc), 3(bc)

Timer (continued)

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

Quad Cities 1 and 2 3.3.5.1-12 Amendment No. 204/200

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 4)

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

4. ADS Trip System A (continued)

d. Core Spray Pump 1, 2 G SR 3.3.5.1.2 101.9 psig Discharge SR 3.3.5.1.4 and Pressure-High 2(bc), 3(bc) SR 3.3.5.1.7 148.1 psig

e. Low Pressure Coolant 1, 4 G SR 3.3.5.1.2 101.6 psig Injection Pump SR 3.3.5.1.4 and Discharge 2(bc), 3(bc) SR 3.3.5.1.7 148.4 psig Pressure-High

f. Automatic 1, 1 G SR 3.3.5.1.6 530 seconds Depressurization SR 3.3.5.1.7 System Low Low Water 2(bc), 3(bc)

Level Actuation Timer

5. ADS Trip System B

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 -55.2 inches LevelLow Low SR 3.3.5.1.2 2(bc), 3(bc) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7 1, SR 3.3.5.1.2

b. Drywell PressureHigh 2 F 2.43 psig SR 3.3.5.1.4 2(bc), 3(bc) SR 3.3.5.1.7

c. Automatic 1, 1 G SR 3.3.5.1.6 119 seconds Depressurization SR 3.3.5.1.7 System Initiation 2(bc), 3(bc)

Timer

d. Core Spray Pump 1, 2 G SR 3.3.5.1.2 101.9 psig Discharge SR 3.3.5.1.4 and PressureHigh 2(bc), 3(bc) SR 3.3.5.1.7 148.1 psig

e. Low Pressure Coolant 1, 4 G SR 3.3.5.1.2 101.6 psig Injection Pump SR 3.3.5.1.4 and Discharge 2(bc), 3(bc) SR 3.3.5.1.7 148.4 psig PressureHigh

f. Automatic 1, 1 G SR 3.3.5.1.6 530 seconds Depressurization SR 3.3.5.1.7 System Low Low Water 2(bc), 3(bc)

Level Actuation Timer (bc) With reactor steam dome pressure > 150 psig.

Quad Cities 1 and 2 3.3.5.1-13 Amendment No. 204/200

RPV Water Inventory Control Instrumentation 3.3.5.2 3.3 INSTRUMENTATION 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.2 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.2-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.2-1 for the channel.

B. As required by B.1 Declare associated Immediately Required Action A.1 penetration flow and referenced in path(s) incapable of Table 3.3.5.2-1. automatic isolation.

AND B.2 Calculate DRAIN TIME. Immediately C. As required by C.1 Place channel in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action A.1 and referenced in Table 3.3.5.2-1.

D. As required by D.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.1 OPERABLE status.

and referenced in Table 3.3.5.2-1.

(continued)

Quad Cities 1 and 2 3.3.5.2-1 Amendment No. 185/180

RPV Water Inventory Control Instrumentation 3.3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Declare associated low Immediately associated Completion pressure ECCS Time of Condition C or injection/spray D not met. subsystem inoperable.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.5.2-1 to determine which SRs apply for each ECCS Function.

SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.5.2-2 Amendment No. 237/230

RPV Water Inventory Control Instrumentation 3.3.5.2 Table 3.3.5.2-1 (Page 1 of 1)

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

1. Core Spray System

a. Reactor Steam Dome 4, 5 2 (a) C SR 3.3.5.2.2 342 psig PressureLow (Permissive)

b. Core Spray Pump 4, 5 1 per D SR 3.3.5.2.2 577 gpm Discharge FlowLow pump (a) and (Bypass) 830 gpm

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Steam Dome 4, 5 2 (a) C SR 3.3.5.2.2 342 psig PressureLow (Permissive)

b. Low Pressure Coolant 4, 5 1 per D SR 3.3.5.2.2 2526 gpm Injection Pump loop (a)

Discharge FlowLow (Bypass)

3. RHR Shutdown Cooling System (SDC) Isolation

a. Reactor Vessel Water (b) 1 per trip B SR 3.3.5.2.1 3.8 inches LevelLow system SR 3.3.5.2.2

4. Reactor Water Cleanup (RWCU) System Isolation

a. Reactor Vessel Water (b) 1 per trip B SR 3.3.5.2.1 3.8 inches LevelLow system SR 3.3.5.2.2 (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.

Quad Cities 1 and 2 3.3.5.2-3 Amendment No. 237/230

RCIC System Instrumentation 3.3.5.32 3.3 INSTRUMENTATION 3.3.5.32 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.32 The RCIC System instrumentation for each Function in Table 3.3.5.32-1 shall be OPERABLE.

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

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Enter the Condition Immediately channels inoperable. referenced in Table 3.3.5.32-1 for the channel.

B. As required by B.1 Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.32-1. initiation capability AND B.2 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

(continued)

Quad Cities 1 and 2 3.3.5.32-1 Amendment No. 199/195

RCIC System Instrumentation 3.3.5.32 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.1 OPERABLE status.

and referenced in Table 3.3.5.32-1.

D. As required by D.1 --------NOTE---------

Required Action A.1 Only applicable if and referenced in RCIC pump suction is Table 3.3.5.32-1. not aligned to the suppression pool.

Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable. discovery of loss of RCIC initiation capability AND D.2.1 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

OR D.2.2 Align RCIC pump 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months suction to the suppression pool.

E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable.

Time of Condition B, C, or D not met.

Quad Cities 1 and 2 3.3.5.32-2 Amendment No. 199/195

RCIC System Instrumentation 3.3.5.32 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.5.32-1 to determine which SRs apply for each RCIC 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 as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 5; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1, 3, and 4 provided the associated Function maintains RCIC initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.32.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.2 CALIBRATE the trip unit. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.3 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.4 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.5.32-3 Amendment No. 248/243

RCIC System Instrumentation 3.3.5.32 Table 3.3.5.32-1 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation CONDITIONS REQUIRED REFERENCED CHANNELS FROM REQUIRED SURVEILLANCE ALLOWABLE FUNCTION PER FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 4 B SR 3.3.5.32.1 -55.2 inches LevelLow Low SR 3.3.5.32.2 SR 3.3.5.32.3 SR 3.3.5.32.4 SR 3.3.5.32.5

2. Reactor Vessel Water 2 C SR 3.3.5.32.1 50.34 inches LevelHigh SR 3.3.5.32.2 SR 3.3.5.32.3 SR 3.3.5.32.4 SR 3.3.5.32.5

3. Contaminated Condensate 2 D SR 3.3.5.32.3 598 ft 1 inch Storage Tank (CCST) SR 3.3.5.32.4 Level-Low SR 3.3.5.32.5

4. Suppression Pool Water 2 D SR 3.3.5.32.3 15 ft LevelHigh SR 3.3.5.32.4 11.25 inches SR 3.3.5.32.5

5. Manual Initiation 1 C SR 3.3.5.32.5 NA Quad Cities 1 and 2 3.3.5.32-4 Amendment No. 204/200

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time for Condition F AND not met.

G.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR As required by Required Action C.1 and referenced in Table 3.3.6.1-1.

H. As required by H.1 Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action C.1 standby liquid and referenced in control subsystem Table 3.3.6.1-1. (SLC) inoperable.

OR H.2 Isolate the Reactor 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Water Cleanup System.

I. As required by I.1 Initiate action to Immediately Required Action C.1 restore channel to and referenced in OPERABLE status.

Table 3.3.6.1-1.

OR I.2 Initiate action to Immediately isolate the Residual Heat Removal (RHR)

Shutdown Cooling System.

Quad Cities 1 and 2 3.3.6.1-3 Amendment No. 199/195

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 3)

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

5. Reactor Water Cleanup System Isolation

a. SLC System Initiation 1,2,3 1 H SR 3.3.6.1.7 NA

b. Reactor Vessel Water 1,2,3 2 F SR 3.3.6.1.1 3.8 inches LevelLow SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

6. RHR Shutdown Cooling System Isolation

a. Reactor Vessel 1,2,3 2 F SR 3.3.6.1.2 130 psig PressureHigh SR 3.3.6.1.4 SR 3.3.6.1.7

b. Reactor Vessel Water 3,4,5 2(b) I SR 3.3.6.1.1 3.8 inches LevelLow SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (b) In MODES 4 and 5, provided RHR Shutdown Cooling System integrity is maintained, only one channel per trip system with an isolation signal available to one shutdown cooling pump suction isolation valve is required.

Quad Cities 1 and 2 3.3.6.1-8 Amendment No. 248/243

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

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

1. Reactor Vessel Water 1,2,3, 2 SR 3.3.6.2.1 3.8 inches LevelLow (a) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Drywell PressureHigh 1,2,3 2 SR 3.3.6.2.2 2.43 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1,2,3, 2 SR 3.3.6.2.1 9 mR/hr RadiationHigh (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1,2,3, 2 SR 3.3.6.2.1 100 mR/hr Radiation-High (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During operations with a potential for draining the reactor vessel.

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

Quad Cities 1 and 2 3.3.6.2-5 Amendment No. 248/243

CREV System Isolation Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Ventilation (CREV) System Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 C SR 3.3.7.1.1 3.8 inches LevelLow (a) 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 PressureHigh 1,2,3 2 C SR 3.3.7.1.2 2.43 psig SR 3.3.7.1.4 SR 3.3.7.1.6

3. Main Steam Line 1,2,3 2 per MSL B SR 3.3.7.1.1 248.1 psid(bc)

Flow-High SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6

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

5. Reactor Building 1,2,3, 2 B SR 3.3.7.1.1 9 mR/hr Ventilation Exhaust SR 3.3.7.1.2 RadiationHigh (a),(b) SR 3.3.7.1.4 SR 3.3.7.1.6 (a) During operations with a potential for draining the reactor vessel.

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

(bc) Function 3 is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint.

Quad Cities 1 and 2 3.3.7.1-5 Amendment No. 248/243

ECCSOperating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.1 ECCSOperating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.

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 Low Pressure A.1 Restore LPCI pump to 30 days Coolant Injection OPERABLE status.

(LPCI) pump inoperable.

B. One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition subsystem to OPERABLE A. status.

OR One Core Spray subsystem inoperable.

C. One LPCI pump in each C.1 Restore one LPCI pump 7 days subsystem inoperable. to OPERABLE status.

D. Required Action and --------------NOTE------------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.

B, or C not met. ------------------------------

D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)

Quad Cities 1 and 2 3.5.1-1 Amendment No. 245/240

RPV Water Inventory ControlECCSShutdown 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.2 RPV Water Inventory ControlECCSShutdown 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 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.,

MODE 5, except with the spent fuel storage pool gates removed and water level 23 ft over the top of the reactor pressure vessel flange.

ACTIONS Quad Cities 1 and 2 3.5.2-1 Amendment No. 199/195

RPV Water Inventory ControlECCSShutdown 3.5.2 CONDITION REQUIRED ACTION COMPLETION TIME A. One rRequired ECCS A.1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperable. subsystem to OPERABLE status.

B. Required Action and B.1 Initiate action to Immediately associated Completion establish a method of Time of Condition A water injection not met. capable of operating without offsite electrical power.suspend operations with a potential for draining the reactor vessel (OPDRVs).

(continued)

Quad Cities 1 and 2 3.5.2-2 Amendment No. 199/195

RPV Water Inventory ControlECCSShutdown 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.1 Verify secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . containment boundary is capable of being established in less than the DRAIN TIME.

AND C.2 Verify each secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months containment penetration flow path is capable of being isolated in less than the DRAIN TIME.

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

(continued)

Quad Cities 1 and 2 3.5.2-3 Amendment No. 199/195

RPV Water Inventory ControlECCSShutdown 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. DRAIN TIME < 8 D.1 ------- NOTE --------

hours.Required Required ECCS Action C.2 and injection/spray associated Completion subsystem or Time not met. additional method of water injection shall be capable of operating without offsite electrical power.

D.1 Initiate action Immediately 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 .restore secondary containment to OPERABLE status.

AND D.2 Initiate action to Immediately establish secondary containment boundary.restore one standby gas treatment subsystem to OPERABLE status.

AND (continued)

Quad Cities 1 and 2 3.5.2-4 Amendment No. 199/195

RPV Water Inventory ControlECCSShutdown 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.3 Initiate action to Immediately restore isolation capability in each required isolate each secondary containment penetration flow path notor verify it can be manually isolated from the control room.

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

E. Required Action and E.1 Initiate action to Immediately associated Completion restore DRAIN TIME to Time of Condition C or 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D not met.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify DRAIN TIME 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.5.2-5 Amendment No. 199/195

RPV Water Inventory ControlECCSShutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.21 Verify, for theeach required ECCS In accordance injection/ spray subsystem, the: with the Surveillance

a. Suppression pool water level is Frequency 8.5 ft; or Control Program

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

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

Contaminated condensate storage tank(s) water volume is 140,000 available gallons.

SR 3.5.2.32 Verify, for theeach required ECCS In accordance injection/spray subsystem, locations with the susceptible to gas accumulation are Surveillance sufficiently filled with water. Frequency Control Program SR 3.5.2.43 --------------------NOTES------------------

1. One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

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

Verify for theeach required ECCS In accordance injection/spray subsystem each manual, with the power operated, and automatic valve in the Surveillance flow path, that is not locked, sealed, or Frequency otherwise secured in position, is in the Control Program correct position.

(continued)

Quad Cities 1 and 2 3.5.2-6 Amendment No. 257/252

RPV Water Inventory ControlECCSShutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each required ECCS pump develops the In accordance specified flow rate against a test line with the pressure corresponding to the specified INSERVICE reactor pressure. TESTING TEST LINE PROGRAMInservic PRESSURE e Testing NO. CORRESPONDING Program OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS 4500 gpm 1 90 psig LPCI 4500 gpm 1 20 psig SR 3.5.2.5 Operate the required ECCS injection/spray In accordance subsystem through the recirculation line with the for 10 minutes. Surveillance Frequency Control Program SR 3.5.2.6 Verify each valve credited for In accordance automatically isolating a penetration flow with the path actuates to the isolation position on Surveillance an actual or simulated isolation signal. Frequency Control Program SR 3.5.2.75 -------------------NOTE--------------------

Vessel injection/spray may be excluded.

Verify theeach required ECCS In accordance injection/spray subsystem can be manually with the operated.actuates on an actual or simulated Surveillance automatic initiation signal. Frequency Control Program Quad Cities 1 and 2 3.5.2-7 Amendment No. 266/261

RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (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 RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System A.1 Verify by Immediately inoperable. administrative means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System 14 days to OPERABLE status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Quad Cities 1 and 2 3.5.3-1 Amendment No. 245/240

PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 --------NOTES--------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is isolated.

D. MSIV leakage rate D.1 Restore leakage rate 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months not within limit. to within limit.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, C, or D not met in MODE 1, 2, or 3. E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

Quad Cities 1 and 2 3.6.1.3-4 Amendment No. 199/195

PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Initiate action to Immediately associated Completion suspend operations Time of Condition A, with a potential for B, C, or D not met for draining the reactor PCIV(s) required to be vessel (OPDRVs).

OPERABLE during MODE 4 or 5. OR F.2 Initiate action to Immediately restore valve(s) to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 ------------------NOTE-------------------

Not required to be met when the 18 inch primary containment vent and purge 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, provided the drywell vent and purge valves and their associated suppression chamber vent and purge valves are not open simultaneously.

Verify each 18 inch primary containment In accordance vent and purge valve, except for the with the torus purge valve, is closed. Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.6.1.3-5 Amendment No. 248/243

Secondary Containment 3.6.4.1 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 A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to 2, or 3. OPERABLE status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met. -----------------------------

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Secondary containment C.1 --------NOTE---------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel ---------------------

assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

AND C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.1-1 Amendment No. 245/240

SCIVs 3.6.4.2 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 A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable. one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

Quad Cities 1 and 2 3.6.4.2-1 Amendment No. 233/229

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Required Action and D.1 --------NOTE---------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during ---------------------

movement of recently irradiated fuel Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND D.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.2-3 Amendment No. 233/229

SGT System 3.6.4.3 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 A.1 Restore SGT 7 days inoperable. subsystem to OPERABLE status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, -----------------------------

or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Required Action and ------------NOTE------------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A ----------------------------

not met during movement of recently C.1 Place OPERABLE SGT Immediately irradiated fuel subsystem in assemblies in the operation.

secondary containment or during OPDRVs. OR (continued)

Quad Cities 1 and 2 3.6.4.3-1 Amendment No. 245/240

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND C.2.2 Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems D.1 Restore one SGT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable in MODE 1, subsystem to 2, or 3. OPERABLE status.

E. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition D when entering MODE 3.

not met. -----------------------------

E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F. Two SGT subsystems F.1 --------NOTE--------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel --------------------

assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in secondary containment.

AND F.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.3-2 Amendment No. 245/240

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

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

The main 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. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3 for to OPERABLE status.

reasons other than Condition C.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, -----------------------------

or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. CREV system C.1 Initiate action to Immediately inoperable due to implement mitigating inoperable CRE actions.

boundary in MODE 1, 2, or 3. AND C.2 Verify mitigating 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits AND (continued)

Quad Cities 1 and 2 3.7.4-1 Amendment No. 245/240

CREV System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Restore CRE boundary 90 days to OPERABLE status D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C AND not met in MODE 1, 2, or 3. D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. CREV System inoperable ------------NOTE-------------

during movement of LCO 3.0.3 is not applicable.

recently irradiated -----------------------------

fuel assemblies in the secondary containment E.1 Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in OR the secondary containment.

CREV System inoperable due to an inoperable AND CRE boundary during movement of recently E.2 Initiate action to Immediately irradiated fuel suspend OPDRVs.

assemblies in the secondary containment or during OPDRVs.

Quad Cities 1 and 2 3.7.4-2 Amendment No. 245/240

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System 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. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1, AC System to OPERABLE 2, or 3. status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, -----------------------------

or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Control Room Emergency ------------NOTE-------------

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during -----------------------------

movement of recently irradiated fuel C.1 Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.7.5-1 Amendment No. 245/240

AC SourcesShutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

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

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

(continued)

Quad Cities 1 and 2 3.8.2-3 Amendment No. 233/229

AC SourcesShutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

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

AND B.3 Initiate action to Immediately suspend OPDRVs.

AND B.34 Initiate action to Immediately restore required DG to OPERABLE status.

Quad Cities 1 and 2 3.8.2-4 Amendment No. 233/229

AC SourcesShutdown 3.8.2 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 3.8.1.10 through SR 3.8.1.12,and SR 3.8.1.14 through SR 3.8.1.19.

2. SR 3.8.1.13 and SR 3.8.1.19 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 ControlECCSShutdown."

For AC sources required to be OPERABLE the In accordance SRs of Specification 3.8.1, except with applicable SR 3.8.1.9, SR 3.8.1.20, and SR 3.8.1.21, SRs are applicable.

Quad Cities 1 and 2 3.8.2-5 Amendment No. 199/195

DC SourcesShutdown 3.8.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.

AND A.2.34 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 -------------------NOTE--------------------

The following SRs are not required to be performed for the 250 VDC electrical power subsystem: SR 3.8.4.6, SR 3.8.4.7, and SR 3.8.4.8.

For DC electrical power subsystems required In accordance to be OPERABLE the following SRs are with applicable applicable: SRs SR 3.8.4.1, SR 3.8.4.2, SR 3.8.4.3, SR 3.8.4.4, SR 3.8.4.5, SR 3.8.4.6, SR 3.8.4.7, and SR 3.8.4.8.

Quad Cities 1 and 2 3.8.5-2 Amendment No. 199/195

Distribution Systems Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

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

AND A.2.34 Initiate actions to Immediately restore required AC and DC electrical power distribution subsystems to OPERABLE status.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and In accordance voltage to required AC and DC electrical with the power distribution subsystems. Surveillance Frequency Control Program Quad Cities 1 and 2 3.8.8-2 Amendment No. 248/243

Quad Cities Nuclear Power Station, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 3 - REVISED TECHNICAL SPECIFICATIONS PAGES TOC Page i 3.3.5.2-3 3.6.1.3-4 TOC Page ii 3.3.5.3-1 3.6.1.3-5 1.1-1 3.3.5.3-2 3.6.4.1-1 1.1-2 3.3.5.3-3 3.6.4.2-1 1.1-3 3.3.5.3-4 3.6.4.2-3 1.1-4 3.3.6.1-3 3.6.4.3-1 1.1-5 3.3.6.1-8 3.6.4.3-2 1.1-6 3.3.6.2-5 3.7.4-1 3.3.5.1-2 3.3.7.1-5 3.7.4-2 3.3.5.1-3 3.5.1-1 3.7.5-1 3.3.5.1-5 3.5.2-1 3.8.2-3 3.3.5.1-10 3.5.2-2 3.8.2-4 3.3.5.1-11 3.5.2-3 3.8.2-5 3.3.5.1-12 3.5.2-4 3.8.5-2 3.3.5.1-13 3.5.2-5 3.8.8-2 3.3.5.2-1 3.5.2-6 3.3.5.2-2 3.5.3-1

TABLE OF CONTENTS 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) 2.1 SLs......................................................2.0-1 2.2 SL Violations............................................2.0-1 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) ......................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 3.3.1.3 Oscillation Power Range Monitor (OPRM) Instrumentation...3.3.1.3-1 3.3.2.1 Control Rod Block Instrumentation........................3.3.2.1-1 3.3.2.2 Feedwater System 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 3.3.4.1 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 Pressure Vessel (RPV) Water Inventory Control Instrumentation.......................................3.3.5.2-1 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation.......................................3.3.5.3-1 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 Relief Valve Instrumentation.............................3.3.6.3-1 (continued)

Quad Cities 1 and 2 i Amendment No. 227/222

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation........................................3.3.7.1-1 3.3.7.2 Mechanical Vacuum Pump Trip 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 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 and Relief Valves ................................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 SystemHot Shutdown....................................3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling SystemCold 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 SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.1 ECCSOperating...........................................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 Pressure. .......................................3.6.1.4-1 3.6.1.5 Drywell Air Temperature..................................3.6.1.5-1 3.6.1.6 Low Set Relief Valves....................................3.6.1.6-1 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers...............................................3.6.1.7-1 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers...........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 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray.......3.6.2.4-1 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure.....3.6.2.5-1 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 (continued)

Quad Cities 1 and 2 ii Amendment No. 199/195

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific HEAT GENERATION RATE planar height and is equal to the sum of the (APLHGR) LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

(continued)

Quad Cities 1 and 2 1.1-1 Amendment No. 199/195

Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

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 REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. 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 the inhalation committed dose conversion factors in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

(continued)

Quad Cities 1 and 2 1.1-2 Amendment No. 233/229

Definitions 1.1 1.1 Definitions (continued)

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 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 is 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.

(continued)

Quad Cities 1 and 2 1.1-3 Amendment No. 266/261

Definitions 1.1 1.1 Definitions DRAIN TIME c. The penetration flow paths required to be (continued) 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).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and (continued)

Quad Cities 1 and 2 1.1-4 Amendment No. 202/198

Definitions 1.1 1.1 Definitions LEAKAGE d. Pressure Boundary LEAKAGE (continued)

LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

(continued)

Quad Cities 1 and 2 1.1-5 Amendment No. 254/249

Definitions 1.1 1.1 Definitions (continued)

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2957 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from the opening of the sensor contact until the TIME opening of the trip actuator. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is > 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME shall be RESPONSE TIME that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

Quad Cities 1 and 2 1.1-6 Amendment No. 248/243

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. As required by B.1 --------NOTE--------

Required Action A.1 Only applicable for and referenced in Functions 1.a, 1.b, Table 3.3.5.1-1. 2.a, 2.b, 2.d, and 2.j.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND B.2 --------NOTE---------

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

Declare High Pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Coolant Injection discovery of (HPCI) System loss of HPCI inoperable. initiation capability AND B.3 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

(continued)

Quad Cities 1 and 2 3.3.5.1-2 Amendment No. 199/195

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 --------NOTE--------

Required Action A.1 Only applicable for and referenced in Functions 1.c, 1.e, Table 3.3.5.1-1. 2.c, 2.e, 2.g, 2.h, 2.i, and 2.k.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND C.2 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

(continued)

Quad Cities 1 and 2 3.3.5.1-3 Amendment No. 199/195

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 --------NOTE--------

Required Action A.1 Only applicable for and referenced in Functions 1.d Table 3.3.5.1-1. and 2.f.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. subsystems in both divisions AND E.2 Restore channel to 7 days OPERABLE status.

(continued)

Quad Cities 1 and 2 3.3.5.1-5 Amendment No. 199/195

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 4)

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

1. Core Spray System

a. Reactor Vessel Water 1, 2, 3 B SR 3.3.5.1.1 -55.2 inches 4(a)

LevelLow Low SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell PressureHigh 1, 2, 3 4(a) B SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 SR 3.3.5.1.7

c. Reactor Steam Dome 1, 2, 3 2 C SR 3.3.5.1.2 306 psig and PressureLow SR 3.3.5.1.4 342 psig (Permissive) SR 3.3.5.1.7

d. Core Spray Pump 1, 2, 3 1 per pump E SR 3.3.5.1.2 577 gpm Discharge FlowLow SR 3.3.5.1.3 and (Bypass) SR 3.3.5.1.6 830 gpm SR 3.3.5.1.7

e. Core Spray Pump 1, 2, 3 1 per pump C SR 3.3.5.1.6 11.4 seconds Start-Time Delay SR 3.3.5.1.7 Relay

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Vessel Water 1, 2, 3 4 B SR 3.3.5.1.1 -55.2 inches LevelLow Low SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell PressureHigh 1, 2, 3 4 B SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 SR 3.3.5.1.7

c. Reactor Steam Dome 1, 2, 3 2 C SR 3.3.5.1.2 306 psig and Pressure Low SR 3.3.5.1.4 342 psig (Permissive) SR 3.3.5.1.7 (continued)

(a) Also required to initiate the associated diesel generator (DG).

Quad Cities 1 and 2 3.3.5.1-10 Amendment No. 204/200

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 4)

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

2. LPCI System (continued)

d. Reactor Steam Dome 1, 2, 3 4 B SR 3.3.5.1.2 868 psig and PressureLow (Break SR 3.3.5.1.5 891 psig Detection) SR 3.3.5.1.7

e. Low Pressure Coolant 1, 2, 3 1 per pump C SR 3.3.5.1.6 6.7 seconds Injection Pump SR 3.3.5.1.7 Start-Time Delay Relay Pumps B and D

f. Low Pressure 1, 2, 3 1 per loop E SR 3.3.5.1.2 2526 gpm Coolant Injection Pump SR 3.3.5.1.3 Discharge FlowLow SR 3.3.5.1.6 (Bypass) SR 3.3.5.1.7

g. Recirculation Pump 1, 2, 3 4 per pump C SR 3.3.5.1.2 2.3 psid Differential SR 3.3.5.1.6 Pressure-High (Break SR 3.3.5.1.7 Detection)

h. Recirculation Riser 1, 2, 3 4 C SR 3.3.5.1.2 2.15 psid Differential SR 3.3.5.1.6 PressureHigh (Break SR 3.3.5.1.7 Detection)

i. Recirculation Pump 1, 2, 3 2 C SR 3.3.5.1.6 0.82 seconds Differential Pressure SR 3.3.5.1.7 Time DelayRelay (Break Detection)

j. Reactor Steam Dome 1, 2, 3 2 B SR 3.3.5.1.6 2.26 seconds Pressure Time Delay SR 3.3.5.1.7 Relay (Break Detection)

k. Recirculation Riser 1, 2, 3 2 C SR 3.3.5.1.6 0.82 seconds Differential Pressure SR 3.3.5.1.7 Time DelayRelay (Break Detection)

(continued)

Quad Cities 1 and 2 3.3.5.1-11 Amendment No. 204/200

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 4)

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

3. High Pressure Coolant Injection (HPCI) System

a. Reactor Vessel Water 1, 4 B SR 3.3.5.1.1 -55.2 inches Level -Low Low SR 3.3.5.1.2 2(b), 3(b) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell Pressure High 1, 4 B SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 2(b), 3(b) SR 3.3.5.1.7

c. Reactor Vessel Water 1, 2 C SR 3.3.5.1.2 50.34 inches Level High SR 3.3.5.1.3 2(b), 3(b) SR 3.3.5.1.6 SR 3.3.5.1.7

d. Contaminated 1, 2 D SR 3.3.5.1.2 598 ft Condensate Storage SR 3.3.5.1.6 1 inch Tank (CCST) Level Low 2(b), 3(b) SR 3.3.5.1.7

e. Suppression Pool Water 1, 2 D SR 3.3.5.1.2 15 ft Level High SR 3.3.5.1.6 11.25 inches 2(b), 3(b) SR 3.3.5.1.7

f. High Pressure Coolant 1, 1 E SR 3.3.5.1.2 634 gpm Injection Pump SR 3.3.5.1.4 Discharge Flow Low 2(b), 3(b) SR 3.3.5.1.7 (Bypass) 1, 1 C SR 3.3.5.1.7 NA

g. Manual Initiation 2(b), 3(b)

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 -55.2 inches LevelLow Low SR 3.3.5.1.2 2(b), 3(b) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7

b. Drywell PressureHigh 1, 2 F SR 3.3.5.1.2 2.43 psig SR 3.3.5.1.4 2(b), 3(b) SR 3.3.5.1.7

c. Automatic 1, 1 G SR 3.3.5.1.6 119 seconds Depressurization SR 3.3.5.1.7 System Initiation 2(b), 3(b)

Timer (continued)

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

Quad Cities 1 and 2 3.3.5.1-12 Amendment No. 204/200

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 4)

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

4. ADS Trip System A (continued)

d. Core Spray Pump 1, 2 G SR 3.3.5.1.2 101.9 psig Discharge SR 3.3.5.1.4 and Pressure-High 2(b), 3(b) SR 3.3.5.1.7 148.1 psig

e. Low Pressure Coolant 1, 4 G SR 3.3.5.1.2 101.6 psig Injection Pump SR 3.3.5.1.4 and Discharge 2(b), 3(b) SR 3.3.5.1.7 148.4 psig Pressure-High

f. Automatic 1, 1 G SR 3.3.5.1.6 530 seconds Depressurization SR 3.3.5.1.7 System Low Low Water 2(b), 3(b)

Level Actuation Timer

5. ADS Trip System B

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 -55.2 inches LevelLow Low SR 3.3.5.1.2 2(b), 3(b) SR 3.3.5.1.3 SR 3.3.5.1.6 SR 3.3.5.1.7 1, SR 3.3.5.1.2

b. Drywell PressureHigh 2 F 2.43 psig SR 3.3.5.1.4 2(b), 3(b) SR 3.3.5.1.7

c. Automatic 1, 1 G SR 3.3.5.1.6 119 seconds Depressurization SR 3.3.5.1.7 System Initiation 2(b), 3(b)

Timer

d. Core Spray Pump 1, 2 G SR 3.3.5.1.2 101.9 psig Discharge SR 3.3.5.1.4 and PressureHigh 2(b), 3(b) SR 3.3.5.1.7 148.1 psig

e. Low Pressure Coolant 1, 4 G SR 3.3.5.1.2 101.6 psig Injection Pump SR 3.3.5.1.4 and Discharge 2(b), 3(b) SR 3.3.5.1.7 148.4 psig PressureHigh

f. Automatic 1, 1 G SR 3.3.5.1.6 530 seconds Depressurization SR 3.3.5.1.7 System Low Low Water 2(b), 3(b)

Level Actuation Timer (b) With reactor steam dome pressure > 150 psig.

Quad Cities 1 and 2 3.3.5.1-13 Amendment No. 204/200

RPV Water Inventory Control Instrumentation 3.3.5.2 3.3 INSTRUMENTATION 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.2 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.2-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.2-1 for the channel.

B. As required by B.1 Declare associated Immediately Required Action A.1 penetration flow and referenced in path(s) incapable of Table 3.3.5.2-1. automatic isolation.

AND B.2 Calculate DRAIN TIME. Immediately C. As required by C.1 Place channel in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action A.1 and referenced in Table 3.3.5.2-1.

D. As required by D.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.1 OPERABLE status.

and referenced in Table 3.3.5.2-1.

(continued)

Quad Cities 1 and 2 3.3.5.2-1 Amendment No. 185/180

RPV Water Inventory Control Instrumentation 3.3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Declare associated low Immediately associated Completion pressure ECCS Time of Condition C or injection/spray D not met. subsystem inoperable.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.5.2-1 to determine which SRs apply for each ECCS Function.

SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.5.2-2 Amendment No. 237/230

RPV Water Inventory Control Instrumentation 3.3.5.2 Table 3.3.5.2-1 (Page 1 of 1)

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

1. Core Spray System

a. Reactor Steam Dome 4, 5 2 (a) C SR 3.3.5.2.2 342 psig PressureLow (Permissive)

b. Core Spray Pump 4, 5 1 per D SR 3.3.5.2.2 577 gpm Discharge FlowLow pump (a) and (Bypass) 830 gpm

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Steam Dome 4, 5 2 (a) C SR 3.3.5.2.2 342 psig PressureLow (Permissive)

b. Low Pressure Coolant 4, 5 1 per D SR 3.3.5.2.2 2526 gpm Injection Pump loop (a)

Discharge FlowLow (Bypass)

3. RHR Shutdown Cooling System (SDC) Isolation

a. Reactor Vessel Water (b) 1 per trip B SR 3.3.5.2.1 3.8 inches LevelLow system SR 3.3.5.2.2

4. Reactor Water Cleanup (RWCU) System Isolation

a. Reactor Vessel Water (b) 1 per trip B SR 3.3.5.2.1 3.8 inches LevelLow system SR 3.3.5.2.2 (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.

Quad Cities 1 and 2 3.3.5.2-3 Amendment No. 237/230

RCIC System Instrumentation 3.3.5.3 3.3 INSTRUMENTATION 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.3 The RCIC System instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.

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

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Enter the Condition Immediately channels inoperable. referenced in Table 3.3.5.3-1 for the channel.

B. As required by B.1 Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.3-1. initiation capability AND B.2 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

(continued)

Quad Cities 1 and 2 3.3.5.3-1 Amendment No. 199/195

RCIC System Instrumentation 3.3.5.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.1 OPERABLE status.

and referenced in Table 3.3.5.3-1.

D. As required by D.1 --------NOTE---------

Required Action A.1 Only applicable if and referenced in RCIC pump suction is Table 3.3.5.3-1. not aligned to the suppression pool.

Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable. discovery of loss of RCIC initiation capability AND D.2.1 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

OR D.2.2 Align RCIC pump 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months suction to the suppression pool.

E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable.

Time of Condition B, C, or D not met.

Quad Cities 1 and 2 3.3.5.3-2 Amendment No. 199/195

RCIC System Instrumentation 3.3.5.3 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.5.3-1 to determine which SRs apply for each RCIC 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 as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 5; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1, 3, and 4 provided the associated Function maintains RCIC initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.3.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.2 CALIBRATE the trip unit. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.3 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.4 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.5.3-3 Amendment No. 248/243

RCIC System Instrumentation 3.3.5.3 Table 3.3.5.3-1 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation CONDITIONS REQUIRED REFERENCED CHANNELS FROM REQUIRED SURVEILLANCE ALLOWABLE FUNCTION PER FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 4 B SR 3.3.5.3.1 -55.2 inches LevelLow Low SR 3.3.5.3.2 SR 3.3.5.3.3 SR 3.3.5.3.4 SR 3.3.5.3.5

2. Reactor Vessel Water 2 C SR 3.3.5.3.1 50.34 inches LevelHigh SR 3.3.5.3.2 SR 3.3.5.3.3 SR 3.3.5.3.4 SR 3.3.5.3.5

3. Contaminated Condensate 2 D SR 3.3.5.3.3 598 ft 1 inch Storage Tank (CCST) SR 3.3.5.3.4 Level-Low SR 3.3.5.3.5

4. Suppression Pool Water 2 D SR 3.3.5.3.3 15 ft LevelHigh SR 3.3.5.3.4 11.25 inches SR 3.3.5.3.5

5. Manual Initiation 1 C SR 3.3.5.3.5 NA Quad Cities 1 and 2 3.3.5.3-4 Amendment No. 204/200

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time for Condition F AND not met.

G.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR As required by Required Action C.1 and referenced in Table 3.3.6.1-1.

H. As required by H.1 Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action C.1 standby liquid and referenced in control subsystem Table 3.3.6.1-1. (SLC) inoperable.

OR H.2 Isolate the Reactor 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Water Cleanup System.

I. As required by I.1 Initiate action to Immediately Required Action C.1 restore channel to and referenced in OPERABLE status.

Table 3.3.6.1-1.

Quad Cities 1 and 2 3.3.6.1-3 Amendment No. 199/195

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 3)

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

5. Reactor Water Cleanup System Isolation

a. SLC System Initiation 1,2,3 1 H SR 3.3.6.1.7 NA

b. Reactor Vessel Water 1,2,3 2 F SR 3.3.6.1.1 3.8 inches LevelLow SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7

6. RHR Shutdown Cooling System Isolation

a. Reactor Vessel 1,2,3 2 F SR 3.3.6.1.2 130 psig PressureHigh SR 3.3.6.1.4 SR 3.3.6.1.7

b. Reactor Vessel Water 3 2 I SR 3.3.6.1.1 3.8 inches LevelLow SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 Quad Cities 1 and 2 3.3.6.1-8 Amendment No. 248/243

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

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

1. Reactor Vessel Water 1,2,3 2 SR 3.3.6.2.1 3.8 inches LevelLow SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Drywell PressureHigh 1,2,3 2 SR 3.3.6.2.2 2.43 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1,2,3,(a) 2 SR 3.3.6.2.1 9 mR/hr RadiationHigh SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1,2,3,(a) 2 SR 3.3.6.2.1 100 mR/hr Radiation-High SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During movement of recently irradiated fuel assemblies in secondary containment.

Quad Cities 1 and 2 3.3.6.2-5 Amendment No. 248/243

CREV System Isolation Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Ventilation (CREV) System Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3 2 C SR 3.3.7.1.1 3.8 inches LevelLow 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 PressureHigh 1,2,3 2 C SR 3.3.7.1.2 2.43 psig SR 3.3.7.1.4 SR 3.3.7.1.6

3. Main Steam Line 1,2,3 2 per MSL B SR 3.3.7.1.1 248.1 psid(b)

Flow-High SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6

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

5. Reactor Building 1,2,3,(a) 2 B SR 3.3.7.1.1 9 mR/hr Ventilation Exhaust SR 3.3.7.1.2 RadiationHigh SR 3.3.7.1.4 SR 3.3.7.1.6 (a) During movement of recently irradiated fuel assemblies in the secondary containment.

(b) Function 3 is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint.

Quad Cities 1 and 2 3.3.7.1-5 Amendment No. 248/243

ECCSOperating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.1 ECCSOperating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.

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 Low Pressure A.1 Restore LPCI pump to 30 days Coolant Injection OPERABLE status.

(LPCI) pump inoperable.

B. One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition subsystem to OPERABLE A. status.

OR One Core Spray subsystem inoperable.

C. One LPCI pump in each C.1 Restore one LPCI pump 7 days subsystem inoperable. to OPERABLE status.

D. Required Action and --------------NOTE------------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.

B, or C not met. ------------------------------

D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)

Quad Cities 1 and 2 3.5.1-1 Amendment No. 245/240

RPV Water Inventory Control 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.2 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 A.1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperable. subsystem to OPERABLE status.

B. Required Action and B.1 Initiate action to Immediately associated Completion establish a method of Time of Condition A water injection not met. capable of operating without offsite electrical power.

(continued)

Quad Cities 1 and 2 3.5.2-1 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.1 Verify secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . containment boundary is capable of being established in less than the DRAIN TIME.

AND C.2 Verify each secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months containment penetration flow path is capable of being isolated in less than the DRAIN TIME.

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

(continued)

Quad Cities 1 and 2 3.5.2-2 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 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 Immediately 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 Immediately establish secondary containment boundary.

AND (continued)

Quad Cities 1 and 2 3.5.2-3 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.3 Initiate action to Immediately isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room.

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

E. Required Action and E.1 Initiate action to Immediately associated Completion restore DRAIN TIME to Time of Condition C or 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D not met.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify DRAIN TIME 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.5.2-4 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for the required ECCS injection/ In accordance spray subsystem, the: with the Surveillance

a. Suppression pool water level is Frequency 8.5 ft; or Control Program

b. Contaminated condensate storage tank(s) water volume is 140,000 available gallons.

SR 3.5.2.3 Verify, for the required ECCS In accordance injection/spray subsystem, locations with the susceptible to gas accumulation are Surveillance sufficiently filled with water. Frequency Control Program SR 3.5.2.4 --------------------NOTE-------------------

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

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

(continued)

Quad Cities 1 and 2 3.5.2-5 Amendment No. 257/252

RPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.5 Operate the required ECCS injection/spray In accordance subsystem through the recirculation line with the for 10 minutes. Surveillance Frequency Control Program SR 3.5.2.6 Verify each valve credited for In accordance automatically isolating a penetration flow with the path actuates to the isolation position on Surveillance an actual or simulated isolation signal. Frequency Control Program SR 3.5.2.7 -------------------NOTE--------------------

Vessel injection/spray may be excluded.

Verify the required ECCS injection/spray In accordance subsystem can be manually operated. with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.5.2-6 Amendment No. 266/261

RPV Water Inventory Control 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM 3.5.2 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 A.1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperable. subsystem to OPERABLE status.

B. Required Action and B.1 Initiate action to Immediately associated Completion establish a method of Time of Condition A water injection not met. capable of operating without offsite electrical power.

(continued)

Quad Cities 1 and 2 3.5.2-1 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.1 Verify secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . containment boundary is capable of being established in less than the DRAIN TIME.

AND C.2 Verify each secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months containment penetration flow path is capable of being isolated in less than the DRAIN TIME.

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

(continued)

Quad Cities 1 and 2 3.5.2-2 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 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 Immediately 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 Immediately establish secondary containment boundary.

AND (continued)

Quad Cities 1 and 2 3.5.2-3 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.3 Initiate action to Immediately isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room.

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

E. Required Action and E.1 Initiate action to Immediately associated Completion restore DRAIN TIME to Time of Condition C or 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D not met.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify DRAIN TIME 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In accordance with the Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.5.2-4 Amendment No. 199/195

RPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for the required ECCS injection/ In accordance spray subsystem, the: with the Surveillance

a. Suppression pool water level is Frequency 8.5 ft; or Control Program

b. Contaminated condensate storage tank(s) water volume is 140,000 available gallons.

SR 3.5.2.3 Verify, for the required ECCS In accordance injection/spray subsystem, locations with the susceptible to gas accumulation are Surveillance sufficiently filled with water. Frequency Control Program SR 3.5.2.4 --------------------NOTE-------------------

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

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

(continued)

Quad Cities 1 and 2 3.5.2-5 Amendment No. 257/252

RPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.5 Operate the required ECCS injection/spray In accordance subsystem through the recirculation line with the for 10 minutes. Surveillance Frequency Control Program SR 3.5.2.6 Verify each valve credited for In accordance automatically isolating a penetration flow with the path actuates to the isolation position on Surveillance an actual or simulated isolation signal. Frequency Control Program SR 3.5.2.7 -------------------NOTE--------------------

Vessel injection/spray may be excluded.

Verify the required ECCS injection/spray In accordance subsystem can be manually operated. with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.5.2-6 Amendment No. 266/261

RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (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 RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System A.1 Verify by Immediately inoperable. administrative means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System 14 days to OPERABLE status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Quad Cities 1 and 2 3.5.3-1 Amendment No. 245/240

PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 --------NOTES--------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is isolated.

D. MSIV leakage rate D.1 Restore leakage rate 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months not within limit. to within limit.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, C, or D not met in MODE 1, 2, or 3. E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Quad Cities 1 and 2 3.6.1.3-4 Amendment No. 199/195

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 ------------------NOTE-------------------

Not required to be met when the 18 inch primary containment vent and purge 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, provided the drywell vent and purge valves and their associated suppression chamber vent and purge valves are not open simultaneously.

Verify each 18 inch primary containment In accordance vent and purge valve, except for the with the torus purge valve, is closed. Surveillance Frequency Control Program (continued)

Quad Cities 1 and 2 3.6.1.3-5 Amendment No. 248/243

Secondary Containment 3.6.4.1 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 A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to 2, or 3. OPERABLE status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met. -----------------------------

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Secondary containment C.1 --------NOTE---------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel ---------------------

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

Quad Cities 1 and 2 3.6.4.1-1 Amendment No. 245/240

SCIVs 3.6.4.2 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 A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable. one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

Quad Cities 1 and 2 3.6.4.2-1 Amendment No. 233/229

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Required Action and D.1 --------NOTE---------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during ---------------------

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

Quad Cities 1 and 2 3.6.4.2-3 Amendment No. 233/229

SGT System 3.6.4.3 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 A.1 Restore SGT 7 days inoperable. subsystem to OPERABLE status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, -----------------------------

or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Required Action and ------------NOTE------------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A ----------------------------

not met during movement of recently C.1 Place OPERABLE SGT Immediately irradiated fuel subsystem in assemblies in the operation.

secondary containment.

OR (continued)

Quad Cities 1 and 2 3.6.4.3-1 Amendment No. 245/240

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

D. Two SGT subsystems D.1 Restore one SGT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable in MODE 1, subsystem to 2, or 3. OPERABLE status.

E. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition D when entering MODE 3.

not met. -----------------------------

E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F. Two SGT subsystems F.1 --------NOTE--------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel --------------------

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

Quad Cities 1 and 2 3.6.4.3-2 Amendment No. 245/240

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

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

The main 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. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3 for to OPERABLE status.

reasons other than Condition C.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, -----------------------------

or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. CREV system C.1 Initiate action to Immediately inoperable due to implement mitigating inoperable CRE actions.

boundary in MODE 1, 2, or 3. AND C.2 Verify mitigating 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits AND (continued)

Quad Cities 1 and 2 3.7.4-1 Amendment No. 245/240

CREV System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Restore CRE boundary 90 days to OPERABLE status D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C AND not met in MODE 1, 2, or 3. D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. CREV System inoperable ------------NOTE-------------

during movement of LCO 3.0.3 is not applicable.

recently irradiated -----------------------------

fuel assemblies in the secondary containment. E.1 Suspend movement of Immediately recently irradiated OR fuel assemblies in the secondary CREV System inoperable containment.

due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment.

Quad Cities 1 and 2 3.7.4-2 Amendment No. 245/240

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System 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. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1, AC System to OPERABLE 2, or 3. status.

B. Required Action and --------------NOTE-----------

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, -----------------------------

or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Control Room Emergency ------------NOTE-------------

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during -----------------------------

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

Quad Cities 1 and 2 3.7.5-1 Amendment No. 245/240

AC SourcesShutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

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

(continued)

Quad Cities 1 and 2 3.8.2-3 Amendment No. 233/229

AC SourcesShutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

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

AND B.3 Initiate action to Immediately restore required DG to OPERABLE status.

Quad Cities 1 and 2 3.8.2-4 Amendment No. 233/229

AC SourcesShutdown 3.8.2 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 3.8.1.10 through SR 3.8.1.12,and SR 3.8.1.14 through SR 3.8.1.19.

2. SR 3.8.1.13 and SR 3.8.1.19 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 In accordance SRs of Specification 3.8.1, except with applicable SR 3.8.1.9, SR 3.8.1.20, and SR 3.8.1.21, SRs are applicable.

Quad Cities 1 and 2 3.8.2-5 Amendment No. 199/195

DC SourcesShutdown 3.8.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 -------------------NOTE--------------------

The following SRs are not required to be performed for the 250 VDC electrical power subsystem: SR 3.8.4.6, SR 3.8.4.7, and SR 3.8.4.8.

For DC electrical power subsystems required In accordance to be OPERABLE the following SRs are with applicable applicable: SRs SR 3.8.4.1, SR 3.8.4.2, SR 3.8.4.3, SR 3.8.4.4, SR 3.8.4.5, SR 3.8.4.6, SR 3.8.4.7, and SR 3.8.4.8.

Quad Cities 1 and 2 3.8.5-2 Amendment No. 199/195

Distribution Systems Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

AND A.2.3 Initiate actions to Immediately restore required AC and DC electrical power distribution subsystems to OPERABLE status.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and In accordance voltage to required AC and DC electrical with the power distribution subsystems. Surveillance Frequency Control Program Quad Cities 1 and 2 3.8.8-2 Amendment No. 248/243

Quad Cities Nuclear Power Station, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 4 -

PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (MARK-UP)

TOC Page i B 3.3.5.3-13 B 3.6.1.3-9 TOC Page ii B 3.3.6.1-19 B 3.6.1.3-10 TOC Page iii B 3.3.6.1-20 B 3.6.2.2-2 B 3.3.5.1-9 B 3.3.6.1-25 B 3.6.4.1-2 B 3.3.5.1-12 B 3.3.6.2-4 B 3.6.4.1-4 B 3.3.5.1-13 B 3.3.6.2-5 B 3.6.4.2-2 B 3.3.5.1-15 B 3.3.6.2-7 B 3.6.4.2-5 B 3.3.5.1-30 B 3.3.7.1-4 B 3.6.4.3-3 B 3.3.5.1-32 B 3.3.7.1-7 B 3.6.4.3-4 B 3.3.5.1-35 B 3.5.1-1 B 3.6.4.3-5 B 3.3.5.2-1 B 3.5.1-6 B 3.6.4.3-6 B 3.3.5.2-2 B 3.5.2-1 B 3.7.4-4 B 3.3.5.2-3 B 3.5.2-2 B 3.7.4-7 B 3.3.5.2-4 B 3.5.2-3 B 3.7.5-3 B 3.3.5.2-5 B 3.5.2-4 B 3.7.5-4 B 3.3.5.2-6 B 3.5.2-5 B 3.7.5-5 B 3.3.5.2-7 B 3.5.2-6 B 3.8.2-1 B 3.3.5.2-8 B 3.5.2-7 B 3.8.2-3 B 3.3.5.2-9 B 3.5.2-8 B 3.8.2-4 B 3.3.5.2-10 B 3.5.2-9 B 3.8.2-5 B 3.3.5.3-1 B 3.5.2-10 B 3.8.2-6 B 3.3.5.3-2 B 3.5.2-11 B 3.8.2-7 B 3.3.5.3-3 B 3.5.2-12 B 3.8.5-1 B 3.3.5.3-4 B 3.5.2-13 B 3.8.5-3 B 3.3.5.3-5 B 3.5.2-14 B 3.8.5-4 B 3.3.5.3-6 B 3.5.2-15 B 3.8.5-5 B 3.3.5.3-7 B 3.5.2-16 B 3.8.8-1 B 3.3.5.3-8 B 3.5.2-17 B 3.8.8-2 B 3.3.5.3-9 B 3.5.3-1 B 3.8.8-3 B 3.3.5.3-10 B 3.5.3-2 B 3.8.8-4 B 3.3.5.3-11 B 3.5.3-8 B 3.10.8-3 B 3.3.5.3-12 B 3.6.1.3-4 B 3.10.8-4

TABLE OF CONTENTS 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) .................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 3.3.1.3 Oscillation Power Range Monitor (OPRM)

Instrumentation ...................................B 3.3.1.3-1 B 3.3.2.1 Control Rod Block Instrumentation ...................B 3.3.2.1-1 B 3.3.2.2 Feedwater System 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 3.3.4.1 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 Pressure Vessel (RPV) Water Inventory Instrumentation ...................................B 3.3.5.2-1 B 3.3.5.32 Reactor Core Isolation Cooling (RCIC) System Instrumentation ...................................B 3.3.5.32-1 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 Relief Valve Instrumentation ........................B 3.3.6.3-1 B 3.3.7.1 Control Room Emergency Ventilation (CREV)

System Instrumentation ............................B 3.3.7.1-1 B 3.3.7.2 Mechanical Vacuum Pump Trip Instrumentation .........B 3.3.7.2-1 (continued)

Quad Cities 1 and 2 i Revision 41

TABLE OF CONTENTS B 3.3 INSTRUMENTATION (continued)

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 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 and Relief Valves ...........................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 SystemHot Shutdown ...............................B 3.4.7-1 B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling SystemCold 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 SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCSOperating ......................................B 3.5.1-1 B 3.5.2 RPV Water Inventory ControlECCSShutdown ............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 Pressure ....................................B 3.6.1.4-1 B 3.6.1.5 Drywell Air Temperature .............................B 3.6.1.5-1 B 3.6.1.6 Low Set Relief Valves ...............................B 3.6.1.6-1 B 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers ..........................................B 3.6.1.7-1 B 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers ......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 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray ..B 3.6.2.4-1 B 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure ..........................................B 3.6.2.5-1 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 (continued)

Quad Cities 1 and 2 ii Revision 00

TABLE OF CONTENTS (continued)

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 Diesel Generator Cooling Water (DGCW) System ........B 3.7.2-1 B 3.7.3 Ultimate Heat Sink (UHS) ............................B 3.7.3-1 B 3.7.4 Control Room Emergency Ventilation (CREV) System ....B 3.7.4-1 B 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) 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 B 3.7.9 Safe Shutdown Makeup Pump (SSMP) System .............B 3.7.9-1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC SourcesOperating ................................B 3.8.1-1 B 3.8.2 AC SourcesShutdown .................................B 3.8.2-1 B 3.8.3 Diesel Fuel Oil Properties and Starting Air .........B 3.8.3-1 B 3.8.4 DC SourcesOperating ................................B 3.8.4-1 B 3.8.5 DC SourcesShutdown .................................B 3.8.5-1 B 3.8.6 Battery Cell Parameters .............................B 3.8.6-1 B 3.8.7 Distribution SystemsOperating ......................B 3.8.7-1 B 3.8.8 Distribution SystemsShutdown .......................B 3.8.8-1 B 3.9 REFUELING OPERATIONS B 3.9.1 Refueling Equipment Interlocks ......................B 3.9.1-1 B 3.9.2 Refuel Position One-Rod-Out Interlock ...............B 3.9.2-1 B 3.9.3 Control Rod Position ................................B 3.9.3-1 B 3.9.4 Control Rod Position Indication .....................B 3.9.4-1 B 3.9.5 Control Rod OPERABILITYRefueling ...................B 3.9.5-1 B 3.9.6 Reactor Pressure Vessel (RPV) Water LevelIrradiated Fuel .............................B 3.9.6-1 B 3.9.7 Reactor Pressure Vessel (RPV) Water LevelNew Fuel or Control Rods ..............................B 3.9.7-1 B 3.9.8 Residual Heat Removal (RHR)High Water Level ........B 3.9.8-1 B 3.9.9 Residual Heat Removal (RHR)Low Water Level .........B 3.9.9-1 B 3.10 SPECIAL OPERATIONS B 3.10.1 Reactor Mode Switch Interlock Testing ...............B 3.10.1-1 B 3.10.2 Single Control Rod WithdrawalHot Shutdown ..........B 3.10.2-1 B 3.10.3 Single Control Rod WithdrawalCold Shutdown .........B 3.10.3-1 B 3.10.4 Single Control Rod Drive (CRD)

RemovalRefueling .................................B 3.10.4-1 B 3.10.5 Multiple Control Rod WithdrawalRefueling ...........B 3.10.5-1 B 3.10.6 Control Rod TestingOperating .......................B 3.10.6-1 B 3.10.7 SHUTDOWN MARGIN (SDM) TestRefueling ................B 3.10.7-1 B 3.10.8 Inservice Leak and Hydrostatic Testing Operation ....B 3.10.8-1 Quad Cities 1 and 2 iii Revision 51

ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND Diesel Generators (continued) monitored by four redundant differential pressure instruments and the Drywell PressureHigh variable is monitored by four redundant pressure switches. The output of each switch/instrument is connected to relays whose contacts are connected to two trip systems. Each trip system is arranged in a one-out-of-two taken twice logic.

One trip system starts the unit DG and the other trip system starts the common DG (DG 1/2). The DGs receive their initiation signals from the CS System initiation logic. The DGs can also be started manually from the control room and locally from the associated DG room. Upon receipt of a loss of coolant accident (LOCA) initiation signal, each DG is automatically started, is ready to load in approximately 13 seconds, and will run in standby conditions (rated voltage and speed, with the DG output breaker open). The DGs will only energize their respective Essential Service System (ESS) buses if a loss of offsite power occurs (Refer to Bases for LCO 3.3.8.1).

APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 1, 2, and 3. The ECCS is initiated LCO, and to preserve the integrity of the fuel cladding by limiting APPLICABILITY the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits.

ECCS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

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, footnote (ab), is added to show that certain ECCS instrumentation Functions are also required to be OPERABLE to perform DG initiation.

(continued)

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ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.a, 2.a. Reactor Vessel Water LevelLow Low (continued)

SAFETY ANALYSES, LCO, and LevelLow Low Function are only required to be OPERABLE APPLICABILITY when the LPCI System is required to be OPERABLE to ensure no single instrument failure can preclude LPCI initiation.

Refer to LCO 3.5.1 and LCO 3.5.2, "ECCSShutdown," for Applicability Bases for the low pressure ECCS subsystems; LCO 3.8.1, "AC SourcesOperating"; and LCO 3.8.2, "AC SourcesShutdown," for Applicability Bases for the DGs.

1.b, 2.b. Drywell PressureHigh High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell PressureHigh Function in order to minimize the possibility of fuel damage. The Drywell PressureHigh Function, along with the Reactor Water LevelLow Low Function, is directly assumed in the LOCA analysis (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.

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 PressureHigh Function is required to be OPERABLE when the ECCS or DG is required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE. Thus, four channels of the CS Drywell PressureHigh Function are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude CS and DG initiation. Also, four channels of the LPCI Drywell PressureHigh Function are required to be OPERABLE in MODES 1, 2, and 3 to ensure no single instrument failure can preclude LPCI initiation.

In MODES 4 and 5, the Drywell PressureHigh Function is 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 DGs.

(continued)

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ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.c, 2.c. Reactor Steam Dome PressureLow (Permissive)

SAFETY ANALYSES, LCO, and Low reactor steam dome pressure signals are used as APPLICABILITY permissives for the low pressure ECCS subsystems. This (continued) 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 channels also delay CS and LPCI pump starts on Reactor Vessel Water LevelLow Low until reactor steam dome pressure is below the setpoint. The Reactor Steam Dome PressureLow (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 PressureLow 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 PressureLow (Permissive) signals are initiated from two pressure switches that sense the reactor steam dome pressure.

The Allowable Value is low enough to prevent overpressuring 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 Reactor Steam Dome PressureLow Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

(continued)

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ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.d, 2.f. Core Spray and Low Pressure Coolant Injection SAFETY ANALYSES, Pump Discharge FlowLow (Bypass) (continued)

LCO, and APPLICABILITY OPERABLE to ensure that no single instrument failure can preclude the ECCS function. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

1.e, 2.e. Core Spray and Low Pressure Coolant Injection Pump StartTime Delay Relay The purpose of this time delay is to stagger the start of CS and LPCI pumps that are in each of Divisions 1 and 2, thus limiting the starting transients on the 4160 V ESS buses.

This Function is only necessary when power is being supplied from the standby power sources (DG). The CS and LPCI Pump StartTime 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 StartTime Delay relays and two LPCI Pump StartTime Delay Relays, one for each CS pump and one for LPCI pump B and D. While each time delay relay is dedicated to a single pump start logic, a single failure of a LPCI Pump StartTime Delay Relay could result in the failure of the three low pressure ECCS pumps, powered from the same ESS bus, to perform their intended function (e.g.,

as in the case where both ECCS pumps on one ESS 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).

The Allowable Values for the CS and LPCI Pump StartTime Delay Relays are chosen to be short enough so that ECCS operation is not degraded.

Each CS and LPCI Pump StartTime Delay Relay Function is required to be OPERABLE only when the associated LPCI subsystem is required to be OPERABLE. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the CS and LPCI subsystems.

(continued)

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ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.1, B.2, and B.3 (continued) both trip systems lose initiation capability, or (f) two Function 2.j 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 DGs 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 DGs being concurrently declared inoperable. For Required Action B.2, redundant automatic initiation capability (i.e., loss of automatic start capability for Functions 3.a and 3.b) is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system.

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. A Notes isare 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 (continued)

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ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C.1 and C.2 (continued) channels, associated with a recirculation pump are inoperable such that both trip systems lose initiation capability, (f) two or more Function 2.h channels are inoperable such that both trip systems lose initiation capability, (g) two Function 2.i channels are inoperable, or (h) two Function 2.k channels are inoperable. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion 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 Functions 1.e, and 2.e, the affected portions are the associated low pressure ECCS pumps. For Functions 1.c and 2.c, the affected portions are the associated ECCS pumps and valves.

For Functions 2.g, 2.h, 2.i, and 2.k, the affected portions are the associated LPCI valves.

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.e, 2.c, 2.e, 2.g, 2.h, 2.i, and 2.k. Required Action C.1 is not applicable to Function 3.g (which also requires entry into this Condition if a channel in this Function is inoperable), since it is the HPCI Manual Initiation Function which is not assumed in any accident or transient analysis. Thus, a total loss of HPCI Manual 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. Required Action C.1 is also not applicable to Function 3.c (which also requires entry into (continued)

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ECCS Instrumentation B 3.3.5.1 BASES ACTIONS E.1 and E.2 (continued)

Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the Core Spray and Low Pressure Coolant Injection Pump Discharge FlowLow (Bypass) Functions result 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 Functions 1.d and 2.f (i.e., low pressure ECCS). Redundant automatic initiation capability is lost if (a) two Function 1.d channels are inoperable or (b) two Function 2.f channels 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 low pressure ECCS pump to be declared inoperable. However, since channels for more than one low pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure 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 low pressure ECCS Functions. 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 4 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 (continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION B 3.3.5.2 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 2.1.1.3 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 2.1.1.3 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 (continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES BACKGROUND (continued) 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.

The purpose of the RPV Water Inventory Control Instrumentation is to support the requirements of LCO 3.5.2, Reactor Pressure Vessel (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 (RHR) Shutdown Cooling (SDC) and Reactor Water Cleanup (RWCU) 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 SAFETY ANALYSES, not required to mitigate any events or accidents evaluated LCO, and in the safety analyses. RPV water inventory control is APPLICABILITY required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 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 operated to maintain adequate reactor vessel water level.

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE As discussed in References 1, 2, 3, 4, and 5, operating SAFETY ANALYSES, experience has shown RPV water inventory to be significant LCO, and to public health and safety. Therefore, RPV Water Inventory APPLICABILITY Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

(continued)

Permissive and interlock setpoints are generally considered as nominal values without regard to measurement accuracy.

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 PressureLow (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. 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 signals are assumed to be OPERABLE and capable of permitting manual operation of the required ECCS subsystem from the Control Room.

The Reactor Steam Dome PressureLow (Permissive) signals are initiated from two pressure switches that sense the reactor steam dome pressure.

The Allowable Value is low enough to prevent overpressurizing the equipment in the low pressure ECCS.

Two channels of Reactor Steam Dome PressureLow Function are only required to be OPERABLE in MODES 4 and 5 when the associated ECCS subsystem is required to be OPERABLE by LCO 3.5.2.

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE 1.b, 2.b. Core Spray and Low Pressure Coolant Injection SAFETY ANALYSES, Pump Discharge FlowLow (Bypass)

LCO, and APPLICABILITY The minimum flow instruments are provided to protect the (continued) associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not sufficiently 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.

One flow transmitter per CS pump and one flow transmitter per LPCI loop are used to detect the associated subsystems' flow rates. The logic is arranged such that each transmitter causes its associated minimum flow valve to open when flow is low with the pump running. The logic will close the minimum flow valve once the closure setpoint is exceeded. The Pump Discharge FlowLow (Bypass) Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump. The Core Spray Discharge FlowLow (Bypass) Allowable Value is also low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core. For LPCI, the closure of the minimum flow valves is not credited.

Each channel of Pump Discharge FlowLow (Bypass) Function is only required to be OPERABLE in MODES 4 and 5 when the associated ECCS subsystem is required to be OPERABLE by LCO 3.5.2 to ensure the pumps are capable of injecting into the RPV when manually operated from the Control Room.

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE Residual Heat Removal (RHR) Shutdown Cooling (SDC) System SAFETY ANALYSES, Isolation LCO, and APPLICABILITY 3.a - Reactor Vessel Water Level-Low (continued)

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 SDC System isolation may be credited for automatic isolation of penetration flow paths associated with the RHR SDC System.

The Reactor Vessel Water LevelLow Function receives input from four reactor vessel water level channels. Each channel inputs into one of four trip strings. Two trip strings make up a trip system and both trip systems must trip to cause an isolation of the RHR SDC suction isolation valves. Any channel will trip the associated trip string. Only one trip string must trip to trip the associated trip system. The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation. Therefore, one trip string in each trip system is required to provide for automatic RHR SDC system isolation.

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.

Residual Heat Removal Shutdown Cooling System Isolation Functions isolate some Group 2 valves (RHR SDC isolation valves).

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE Reactor Water Cleanup (RWCU) System Isolation SAFETY ANALYSES, LCO, and 4.a - Reactor Vessel Water level-Low APPLICABILITY (continued) 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 RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.

The Reactor Vessel Water LevelLow Isolation Function receives input from four reactor vessel water level channels. Each channel inputs into one of four trip stings.

Two trip strings make up a trip system and both trip systems must trip to cause an isolation of the RWCU valves. Any channel will trip the associated trip string. Only one trip string must trip to trip the associated trip system. The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation. Therefore, one trip string in each trip system is required to provide for automatic RWCU system isolation.

The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the ECCS 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.

RWCU Functions isolate some Group 3 valves (RWCU isolation valves).

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES (continued)

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 3.3.5.2-1. The applicable Condition referenced in the table is Function 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 Residual Heat Removal (RHR) Shutdown Cooling (SDC) System Isolation, Reactor Vessel Water Level-Low, and Reactor Water Cleanup System Isolation, Reactor Vessel Water Level-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.

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES ACTIONS C.1 (continued)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. If a required channel of the permissive is inoperable, manual operation of ECCS may be prohibited. Therefore, the affected channel(s) 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 affected channel(s) in the trip condition, manual operation may be performed.

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 the trip condition.

D.1 If a CS or 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 system 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 minimum flow 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.

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES (continued)

SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RPV REQUIREMENTS Water Inventory Control instrumentation Function are found in the SRs column of Table 3.3.5.2-1.

SR 3.3.5.2.1 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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 3.3.5.2.2 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.

(continued)

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RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.2 (continued)

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

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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.

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RCIC System Instrumentation B 3.3.5.32 B 3.3 INSTRUMENTATION B 3.3.5.32 Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate makeup water when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level precluding initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps. A more complete discussion of RCIC System operation is provided in the Bases of LCO 3.5.3, "RCIC System."

The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water LevelLow Low level. The variable is monitored by four level indicating instruments. The outputs are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement. The logic can also be initiated by use of a manual initiation push button.

Once initiated, the RCIC logic seals in and can be reset by the operator only when the reactor vessel water level signals have cleared.

The RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow to the reactor vessel.

The RCIC System also monitors the water levels in the contaminated condensate storage tanks (CCSTs) and the suppression pool since these are the two sources of water for RCIC operation. Reactor grade water in the CCST is the normal source. Upon receipt of a RCIC initiation signal, the CCST suction valve is automatically signaled to open (it is normally in the open position) unless both pump suction valves from the suppression pool are open. If the water level in any CCST falls below a preselected level, first the suppression pool suction valves automatically open, and then when these valves are fully open the CCST suction valve automatically closes. Two level switches are used to detect low water level in each CCST. The outputs for these (continued)

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RCIC System Instrumentation B 3.3.5.32 BASES BACKGROUND switches are common between Units 1 and 2. Any switch can (continued) cause the suppression pool suction valves to open and the CCST suction valve to close. The suppression pool suction valves also automatically open and the CCST suction valve closes if high water level is detected in the suppression pool (one-out-of-two logic). To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes.

The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level trip (two-out-of-two logic), at which time the RCIC turbine steam supply valve, and minimum flow valve to the suppression pool close. The RCIC System automatically restarts if a Reactor Vessel Water LevelLow Low signal is subsequently received.

APPLICABLE The function of the RCIC System to provide makeup coolant to SAFETY ANALYSES, the reactor is used to respond to transient events. The LCO, and RCIC System is not an Engineered Safety Feature System and APPLICABILITY no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the RCIC System, and therefore its instrumentation, meets Criterion 4 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

The OPERABILITY of the RCIC System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.32-1. 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.

Allowable Values are specified for each RCIC System instrumentation 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 (continued)

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RCIC System Instrumentation B 3.3.5.32 BASES APPLICABLE conservative than the nominal trip setpoint, but within its SAFETY ANALYSES, Allowable Value, is acceptable. A channel is inoperable if LCO, and its actual trip setpoint is not within its required APPLICABILITY Allowable Value. Trip setpoints are those predetermined (continued) 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 (or design limits) are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors. These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.

The individual Functions are required to be OPERABLE in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure > 150 psig since this is when RCIC is required to be OPERABLE. Refer to LCO 3.5.3 for Applicability Bases for the RCIC System.

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

1. Reactor Vessel Water LevelLow Low Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel (continued)

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RCIC System Instrumentation B 3.3.5.32 BASES APPLICABLE 1. Reactor Vessel Water LevelLow Low (continued)

SAFETY ANALYSES, LCO, and may be threatened. Should RPV water level decrease too far, APPLICABILITY fuel damage could result. Therefore, the RCIC System is initiated at Reactor Vessel Water LevelLow Low to assist in maintaining water level above the top of the active fuel.

Reactor Vessel Water LevelLow Low signals are initiated from four level indicating instruments 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 LevelLow Low Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be sufficient to avoid injection of low pressure ECCS.

Four channels of Reactor Vessel Water LevelLow Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.

2. Reactor Vessel Water LevelHigh High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Reactor Vessel Water LevelHigh signal is used to close the RCIC turbine steam supply valve, to prevent overflow into the main steam lines (MSLs). The minimum flow valve to the suppression pool also closes.

Reactor Vessel Water LevelHigh signals for RCIC are initiated from two level indicating instruments from the narrow range water level measurement instrumentation, which 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.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES APPLICABLE 2. Reactor Vessel Water LevelHigh (continued)

SAFETY ANALYSES, LCO, and The Reactor Vessel Water LevelHigh Allowable Value is high APPLICABILITY enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.

Two channels of Reactor Vessel Water LevelHigh Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.

3. Contaminated Condensate Storage Tank LevelLow Low level in a CCST indicates the unavailability of an adequate supply of makeup water from this normal source.

Normally, the suction valve between the RCIC pump and the CCST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CCSTs.

However, if the water level in the CCSTs fall below a preselected level, first the suppression pool suction valves automatically open, and then the CCST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CCST suction valve automatically closes.

Two level switches are used to detect low water level in each CCST. The Contaminated Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CCST.

While four channels of Contaminated Condensate Storage Tank LevelLow Function are available, only two channels are required to be OPERABLE when RCIC is required to be OPERABLE and both CCSTs are aligned to the RCIC System. In addition, when one CCST is isolated from the unit RCIC System, the two channels required are those associated with the CCST that is aligned to RCIC. These requirements will ensure that no single instrument failure can preclude RCIC swap to suppression pool source. Refer to LCO 3.5.3 for RCIC Applicability Bases.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES APPLICABLE 4. Suppression Pool Water LevelHigh SAFETY ANALYSES, LCO, and Excessively high suppression pool water level could result APPLICABILITY in the loads on the suppression pool exceeding design values (continued) should there be a blowdown of the reactor vessel pressure through the relief valves. Therefore, signals indicating high suppression pool water level are used to transfer the suction source of RCIC from the CCSTs to the suppression pool to eliminate the possibility of RCIC continuing to provide additional water from a source outside primary containment. This Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CCST suction valve automatically closes.

Suppression pool water level signals are initiated from two level switches. The Allowable Value for the Suppression Pool Water LevelHigh Function is set low enough to ensure that RCIC will be aligned to take suction from the suppression pool before the water level reaches the point at which suppression design loads would be exceeded. The Allowable Value is confirmed by performance of a CHANNEL FUNCTIONAL TEST. This is acceptable since the design layout of the installation ensures the switches will trip at a level lower than the Allowable Value.

Two channels of Suppression Pool Water LevelHigh Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source.

Refer to LCO 3.5.3 for RCIC Applicability Bases.

5. Manual Initiation The Manual Initiation push button switch introduces a signal into the RCIC System initiation logic that is redundant to the automatic protective instrumentation and provides manual initiation capability. There is one push button for the RCIC System.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES APPLICABLE 5. Manual Initiation (continued)

SAFETY ANALYSES, LCO, and The Manual Initiation Function is not assumed in any APPLICABILITY accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the RCIC function as required by the NRC in the plant licensing basis.

There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button. One channel of Manual Initiation is required to be OPERABLE when RCIC is required to be OPERABLE.

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

A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.32-1. The applicable Condition referenced in the Table is Function dependent.

Each time a required channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES ACTIONS B.1 and B.2 (continued)

Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System. In this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. In this situation (loss of 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.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after discovery of loss of RCIC initiation capability.

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 the RCIC System cannot be automatically initiated due to two inoperable, untripped Reactor Vessel Water LevelLow Low channels in the same trip system. 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 or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not credited in any accident or transient analysis, 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. 1) 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, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES ACTIONS C.1 (continued)

A risk based analysis was performed and determined that an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water LevelHigh Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation (high water level trip) capability. As stated above, this loss of automatic RCIC initiation (high water level trip) capability was analyzed and determined to be acceptable. This Condition also applies to the Manual Initiation Function. This is allowed since this Function is not assumed in any accident or transient analysis, thus a total loss of manual initiation capability (Required Action C.1) for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed.

The Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events.

D.1, D.2.1, and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic initiation capability being lost for the RCIC System. In this case if both CCSTs are available RCIC automatic initiation (RCIC source swap over) capability is lost if two required Function 3 channels are inoperable and untripped.

If one CCST is not available, automatic initiation capability is lost if two channels associated with the aligned CCST are inoperable and untripped. In addition, automatic initiation (RCIC source swap over) capability is lost if two Function 4 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Actions D.2.1 and D.2.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued)

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 D.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. 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 or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, 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. 1) 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, the channel must be placed in the tripped condition per Required Action D.2.1, which performs the intended function of the channel (shifting the suction source to the suppression pool). Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case where shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken.

E.1 With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES (continued)

SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3.3.5.32-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:

(a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 5; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1, 3, and 4, provided the associated Function maintains RCIC initiation capability.

Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 1) 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 RCIC will initiate when necessary.

SR 3.3.5.32.1 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 parameter on other similar 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 will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

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

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES SURVEILLANCE SR 3.3.5.32.1 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.32.2 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than that accounted for in the appropriate setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.32.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

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

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

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RCIC System Instrumentation B 3.3.5.32 BASES SURVEILLANCE SR 3.3.5.32.4 REQUIREMENTS (continued) A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.32.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. GENE-770-06-2A, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," December 1992.

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 6.a. Reactor Vessel PressureHigh (continued)

SAFETY ANALYSES, LCO, and on the reactor recirculation loop B suction line. Two APPLICABILITY channels (both providing input into two trip systems) of Reactor Vessel PressureHigh Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1, 2, and 3, since these are the only MODES in which the reactor can be pressurized; thus, equipment 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 residual heat removal shutdown cooling suction valves.

6.b. Reactor Vessel Water LevelLow 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 LevelLow 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 LevelLow 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 LevelLow Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (b) to Table 3.3.6.1-1), only one channel per trip system (with an (continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 6.b. Reactor Vessel Water LevelLow (continued)

SAFETY ANALYSES, LCO, and isolation signal available to one shutdown cooling pump APPLICABILITY suction isolation valve) of the Reactor Vessel Water Level-Low Function is required to be OPERABLE in MODES 4 and 5, provided the 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 LevelLow 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 LevelLow 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, interlocks and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path.

This Function is in the logic for the Group 2 isolation of residual heat removal shutdown cooling suction and injection valves.

ACTIONS A Note has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable primary containment isolation instrumentation channel.

(continued)

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Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS I.1 and I.2 (continued)

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 REQUIREMENTS Primary 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 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 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. 9 and 10) 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 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 (continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE The specific Applicable Safety Analyses, LCO, and SAFETY ANALYSES, Applicability discussions are listed below on a Function by LCO, and Function basis.

APPLICABILITY (continued)

1. Reactor Vessel Water LevelLow 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.

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 Reactor Vessel Water LevelLow Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation of systems on Reactor Vessel Water LevelLow support actions to ensure that any offsite releases are within the limits calculated in the safety analysis (Ref. 2).

Reactor Vessel Water LevelLow signals are initiated from 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 of Reactor Vessel Water LevelLow 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 LevelLow Allowable Value was chosen to be the same as the Reactor Protection System (RPS)

Reactor Vessel Water LevelLow Allowable Value (LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation"), since this could indicate that the capability to cool the fuel is being threatened.

The Reactor Vessel Water LevelLow 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 (continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 1. Reactor Vessel Water LevelLow (continued)

SAFETY ANALYSES, LCO, and Function is also required to be OPERABLE during operations APPLICABILITY with a potential for draining the reactor vessel (OPDRVs) to (continued) ensure that offsite dose limits are not exceeded if core damage occurs.

2. Drywell PressureHigh 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 isolation and initiating of the systems on Drywell PressureHigh supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis (Ref. 2).

High drywell pressure signals are initiated from pressure switches that sense the pressure in the drywell. Four channels of Drywell PressureHigh 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 RPS Drywell PressureHigh Function Allowable Value (LCO 3.3.1.1) since this is indicative of a loss of coolant accident (LOCA).

The Drywell PressureHigh 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.

(continued)

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Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3, 4. Reactor Building Exhaust RadiationHigh and SAFETY ANALYSES, Refueling Floor RadiationHigh (continued)

LCO, and APPLICABILITY 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 uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. Due to radioactive decay, these Functions are only required to isolate secondary containment 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 ).

ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.

A.1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months depending on the Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those Functions that have channel components common to RPS instrumentation and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Functions that do not have channel components common to RPS instrumentation), has been shown to be acceptable (Refs. 3 and 4) to permit restoration of any inoperable channel to OPERABLE status.

This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status (continued)

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CREV System Isolation Instrumentation B 3.3.7.1 BASES APPLICABLE 1. Reactor Vessel Water LevelLow (continued)

SAFETY ANALYSES, LCO, and Reactor Vessel Water LevelLow signals are initiated from APPLICABILITY 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 of Reactor Vessel Water LevelLow Function are available (two channels per trip system) and are required to be OPERABLE to ensure that a single instrument failure can preclude control room emergency zone isolation. The Reactor Vessel Water LevelLow Allowable Value was chosen to be the same as the Reactor Protection System (RPS) Reactor Vessel Water Level-Low Allowable Value (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation").

The Reactor Vessel Water LevelLow Function is required to be OPERABLE in MODES 1, 2, and 3, and during operations with a potential for draining the reactor vessel (OPDRVs) to ensure that the control room personnel are protected during a LOCA. In MODES 4 and 5 at times other than OPDRVs, the probability of a vessel draindown event resulting in a release of radioactive material into the environment is minimal. In addition, adequate protection is performed by the Refueling Floor RadiationHigh and Reactor Building Exhaust RadiationHigh Functions. Therefore, this Function is not required in other MODES and specified conditions.

2. Drywell PressureHigh High pressure in the drywell could indicate a break in the reactor coolant pressure boundary. A high drywell pressure signal could indicate a LOCA and will automatically initiate isolation of the control room emergency zone, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

Drywell PressureHigh signals are initiated from four pressure switches that sense drywell pressure. Four channels of Drywell PressureHigh Function are available (two channels per trip system) and are required to be OPERABLE to ensure that no single instrument failure can (continued)

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CREV System Isolation Instrumentation B 3.3.7.1 BASES APPLICABLE 4, 5. Refueling Floor RadiationHigh and Reactor Building SAFETY ANALYSES, Ventilation Exhaust RadiationHigh (continued)

LCO, and APPLICABILITY The Refueling Floor RadiationHigh Function and Reactor Building Ventilation Exhaust RadiationHigh Function are required to be OPERABLE in MODES 1, 2, and 3 and during movement of recently irradiated fuel assemblies in the secondary containment and operations with a potential for draining the reactor vessel (OPDRVs), to ensure that control room personnel are protected during a LOCA or, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g.,

OPDRVs), the probability of a LOCA or fuel damage is low; thus, the Functions are not required. Also due to radioactive decay, these Functions are only required to initiate isolation of the control room emergency zone 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 ).

ACTIONS A Note has been provided to modify the ACTIONS related to CREV System isolation instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System isolation instrumentation channel.

A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.7.1-1. The applicable Condition specified in the Table is Function dependent.

Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.

(continued)

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ECCSOperating B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM B 3.5.1 ECCSOperating 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 contaminated condensate storage tank (CCST), it is capable of providing a source of water for the HPCI, LPCI and CS systems.

On receipt of an initiation signal, ECCS pumps automatically start; the system aligns and the pumps inject water, taken either from the CCST 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 relief valves and safety/relief valve (S/RV) 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 (continued)

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ECCSOperating B 3.5.1 BASES LCO the limits specified in Reference 9 could be exceeded. All (continued) ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by Reference 9.

LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut-in permissive pressure 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: a) when the system is being realigned to or from the RHR shutdown cooling mode and; b) when the system is in the RHR shutdown cooling mode, whether or not the RHR pump is operating. 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. ECCS rRequirements for MODES 4 and 5 are specified in LCO 3.5.2, "RPV Water Inventory ControlECCSShutdown."

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI System. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI 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 If any 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 ECCS (continued)

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RPV Water Inventory ControlECCSShutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM B 3.5.2 RPV Water Inventory ControlECCSShutdown 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 2.1.1.3 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, "ECCSOperating."

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is SAFETY ANALYSES 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 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material to the environment should an unexpected draining event occur.The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated loss of coolant accident (LOCA). The long term 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.

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). It is reasonable to assumed, based on engineering judgement, that while in MODES 4 and 5, (continued)

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RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES APPLICABLE one low pressure ECCS injection/spray subsystem can be SAFETY ANALYSES manually operated from the control room to maintain adequate (continued) 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.

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).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 2.1.1.3.

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 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 2.1.1.3 and can be managed as part of normal plant operation.

OneTwo low pressure ECCS injection/spray subsystems isare required to be OPERABLE and capable of being manually operated from the control room to provide defense-in-depth should an unexpected draining event occur. AThe low pressure ECCS injection/spray subsystems consists of either one two Core Spray (CS) subsystems, orand onetwo Low Pressure Coolant Injection (LPCI) subsystems. AEach CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or contaminated condensate storage tank(s) (CCST) to the RPV).

AEach LPCI subsystem consists of one motor (continued)

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RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES LCO driven pump, piping, and valves to transfer water from the reactor pressure vessel (

(continued) suppression pool or the CCST(s) to the RPV. In MODES 4 and 5, OPERABLE CCSTs can be credited to support the OPERABILITY of the required ECCS subsystem. 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 valves are not required to be open. Management of gas voids is important to ECCS injection/spray subsystem OPERABILITY.

A LPCI subsystem 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 not otherwise inoperable. Alignment and operation for decay heat removal includes: a) when the system is being realigned to or from the RHR shutdown cooling mode and; b) when the system is in the RHR shutdown cooling mode, whether or not the RHR pump is operating. 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.the restrictions on DRAIN TIME, sufficient time will be available following an unexpected draining event to manually align and operate a LPCI subsystem from the control room to maintain RPV water inventory prior to the RPV water level reaching the TAF.

(continued)

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RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES (continued)

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, RPV Water Inventory Control, and RCIC System. RPV water inventory control is required to protect Safety Limit 2.1.1.3 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 23 ft 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.

(continued)

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RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES (continued)

ACTIONS A.1 and B.1 If any onethe required low pressure ECCS injection/spray subsystem is inoperable, itthe 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 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 TIMEremaining available subsystem and the low probability of an unexpected draining vessel draindown event that would result in loss of RPV water inventory.

IfWith 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 operated 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.suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

(continued)

Quad Cities 1 and 2 B 3.5.2-5 Revision 50

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES ACTIONS C.1, C.2, D.1, D.2, and CD.3 (continued)

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 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.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 required 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 . 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.

(continued)

Quad Cities 1 and 2 B 3.5.2-6 Revision 50

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES ACTIONS C.1, C.2, D.1, D.2, and CD.3 (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. A secondary containment penetration flow path can be considered isolated when one barrier in the flow path is in place. Examples of suitable barriers include, but are not limited to, a closed secondary containment isolation valve (SCIV), a closed manual valve, a blind flange, or another sealing device that sufficiently seals the penetration flow path. 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.

(continued)

Quad Cities 1 and 2 B 3.5.2-7 Revision 50

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES ACTIONS C.1, C.2, D.1, D.2, and CD.3 (continued)

If at least one required 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. The 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.

(continued)

Quad Cities 1 and 2 B 3.5.2-8 Revision 50

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES ACTIONS D.1, D.2, D.3, and D.4 (continued)

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 should an unexpected draining event occur. The Note to Required Action D.1 states that either the ECCS injection/spray subsystem or the additional method of water injection must be capable of operating without offsite electrical power. The additional method of water injection may be manually operated and may consist of one or more systems or subsystems. The additional method of water injection must be able to 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 into 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.

(continued)

Quad Cities 1 and 2 B 3.5.2-9 Revision 00

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES ACTIONS D.1, D.2, D.3, and D.4 (continued)

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. A secondary containment penetration flow path can be considered isolated when one barrier in the flow path is in place. Examples of suitable barriers include, but are not limited to, a closed secondary containment isolation valve (SCIV), a closed manual valve, a blind flange, or another sealing device that sufficiently seals the penetration flow path.

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 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 .

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS 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 (continued)

Quad Cities 1 and 2 B 3.5.2-10 Revision 00

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 (continued)

REQUIREMENTS 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 2.1.1.3 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 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 (SDC)

System is only considered an intact system when misalignment issues (Reference 6) have been precluded by functional valve interlocks or by isolation devices, such that redirection of RPV water out of an RHR SDC subsystem is precluded.

Further, the RHR SDC system is only considered an intact closed system if its controls have not been transferred to Remote Shutdown, which disables the interlocks and isolation signals.

The exclusion of penetration flow paths from the determination of DRAIN TIME must consider the potential (continued)

Quad Cities 1 and 2 B 3.5.2-11 Revision 00

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 (continued)

REQUIREMENTS effects of a single operator error or 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.21 The minimum water level of 8.5 feet above the bottom of the suppression chamber required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for athe required CS System subystem orand LPCI subsystem pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, theall ECCS injection/spray subsystems isare inoperable unless it isthey are aligned to an OPERABLE CCST(s).

When suppression pool level is < 8.5 feet, tThe required CS orand LPCI subsystems isare considered OPERABLE only if itthey can take suction from the CCST(s), and the CCST(s) water volume is sufficient to provide the required NPSH and vortex prevention for the CS pump and LPCI pump. Therefore, a verification that either the suppression pool water level is 8.5 feet or that the required low pressure ECCS injection/spray subsystems isare aligned to take suction from the CCST(s) and the CCST(s) contain 140,000 available gallons of water, equivalent to 12 ft in both CCSTs when they are crosstied (normal configuration) and 13.5 ft in one CCST when they are not crosstied, ensures that the required low pressure ECCS injection/spray subsystems can supply at least 140,000 gallons of makeup water to the RPV.

(continued)

Quad Cities 1 and 2 B 3.5.2-12 Revision 00

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.21 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

However, as noted, only one required low pressure ECCS injection/spray subsystem may take credit for the CCST option during OPDRVs. During OPDRVs, the volume in the CCST(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 CCST(s). This ensures the other required ECCS subsystem has adequate makeup volume.

SR 3.5.2.32, SR 3.5.2.4, and SR 3.5.2.5 The required low pressure ECCS injection/spray flow path piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the ECCS injection/spray subsystem and may also prevent water hammer and pump cavitation.

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.

Maintaining the pump discharge lines of the required ECCS injection/spray subsystem sufficiently full of water ensures that the ECCS subsystem will perform properly. 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 criterion 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 (continued)

Quad Cities 1 and 2 B 3.5.2-13 Revision 43

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.3 (continued)

REQUIREMENTS 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 subset 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 sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

The Bases provided for SR 3.5.1.1, SR 3.5.1.5, and SR 3.5.1.8 are applicable to SR 3.5.2.2, SR 3.5.2.4, and SR 3.5.2.5, respectively.

SR 3.5.2.43 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 available exist 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 initiation signal is allowed to be (continued)

Quad Cities 1 and 2 B 3.5.2-14 Revision 43

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.4 (continued)

REQUIREMENTS 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

In MODES 4 and 5, the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, this SR is modified by Note 1 that allows one LPCI subsystem to 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 not otherwise inoperable.

Alignment and operation for decay heat removal includes:

a) when the system is being realigned to or from the RHR shutdown cooling mode and; b) when the system is in the RHR shutdown cooling mode, whether or not the RHR pump is operating. Because of the 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. This will ensure adequate core cooling if an inadvertent RPV draindown should occur. Note 2 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.

SR 3.5.2.5 Verifying that the required ECCS injection/spray subsystem can be manually started and operated for at least 10 minutes demonstrates that the subsystem is operationally ready to mitigate a draining event. Testing the ECCS injection/spray (continued)

Quad Cities 1 and 2 B 3.5.2-15 Revision 43

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.5 (continued)

REQUIREMENTS subsystem through the full flow test recirculation line is necessary to avoid overfilling the refueling cavity. The minimum operating time of 10 minutes is based on engineering judgement.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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 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 selected Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.7 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)) is capable of being manually operated from the control room, and without delay, to provide additional RPV Water Inventory, if needed.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

Quad Cities 1 and 2 B 3.5.2-16 Revision 43

RPV Water Inventory ControlECCSShutdown B 3.5.2 BASES (continued)

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.

6. General Electric Service Information Letter No. 388, "RHR Valve Misalignment During Shutdown Cooling Operation for BWR 3/4/5/6," February 1983.1. UFSAR, Section 6.3.3.1.2.1.

Quad Cities 1 and 2 B 3.5.2-17 Revision 50

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (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 makeup water and maintain RPV water level above the top of the core. 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 contaminated condensate storage tank (CCST) and the suppression pool. Pump suction is normally aligned to the CCST to minimize injection of suppression pool water into the RPV. However, if the CCST water supply is low, or the suppression pool level is high, 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 makeup water for a wide range of reactor pressures 150 psig to 1120 psig. 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 to the CCST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

(continued)

Quad Cities 1 and 2 B 3.5.3-1 Revision 00

RCIC System B 3.5.3 BASES BACKGROUND The RCIC pump is provided with a minimum flow bypass line, (continued) which discharges to the suppression pool. The valve in this line automatically opens on an initiation signal combined with low flow 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 CCST.

The height of water ( 12 ft) in the CCST is sufficient to maintain the piping full of water up to the first isolation valve in the discharge piping. The relative 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 SAFETY ANALYSES events by providing makeup coolant to the reactor. The RCIC System is not an Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, the system satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The OPERABILITY of the RCIC System provides makeup water 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.

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. and iIn 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 Inventory Control." the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV.

(continued)

Quad Cities 1 and 2 B 3.5.3-2 Revision 50

RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.5 (continued)

REQUIREMENTS low-low water level signal received subsequent to an RPV high water level trip and that the suction is automatically transferred from the CCST to the suppression pool on a CCST low water level signal. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.32 overlaps this Surveillance to provide complete testing of the assumed design function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that excludes vessel injection 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.

REFERENCES 1. UFSAR, Section 5.4.6.

2. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

3. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

Quad Cities 1 and 2 B 3.5.3-8 Revision 43

PCIVs B 3.6.1.3 BASES LCO 10 CFR 50 Appendix R requirements) to be de-activated and (continued) closed, are considered OPERABLE when the valves are closed and de-activated. These passive isolation valves and devices are those listed in Reference 1.

MSIVs must meet additional leakage rate 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, however, are required to be OPERABLE when the to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LCO 3.3.5.26.1, "Reactor Pressure Vessel (RPV) Water Inventory Control Primary Containment Isolation Instrumentation." (This does not include the valves that isolate the associated instrumentation.)

ACTIONS The ACTIONS are modified by a Note allowing penetration flow path(s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.

A second Note has been added to provide 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 appropriate compensatory actions for each inoperable PCIV.

Complying with the Required Actions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions.

(continued)

Quad Cities 1 and 2 B 3.6.1.3-4 Revision 00

PCIVs B 3.6.1.3 BASES ACTIONS D.1 (continued)

With the MSIV leakage rate (SR 3.6.1.3.10) not within limit, the assumptions of the safety analysis may not be met.

Therefore, the leakage must be restored to within limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices.

The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allows a period of time to restore MSIV leakage rate to within limit given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown.

E.1 and E.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.

F.1 and F.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 residual heat removal (RHR) shutdown cooling isolation valves, an (continued)

Quad Cities 1 and 2 B 3.6.1.3-9 Revision 00

PCIVs B 3.6.1.3 BASES (continued)

ACTIONS F.1 and F.2 (continued) 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 vent and purge valves are closed as required or, if open, opened for an allowable reason. If a vent or purge valve is opened in violation of this SR, the valve is considered inoperable.

The torus purge valve, 1601-56, is normally open for pressure control, therefore this valve is excluded from this SR. However, this is acceptable since this valve is designed to automatically close on LOCA conditions. The SR is modified by a Note stating that the SR is not required to be met when the vent or purge 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 provided the drywell vent and purge valves and their associated suppression chamber vent and purge valves are not open simultaneously. The 18 inch vent and purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions, is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.

(continued)

Quad Cities 1 and 2 B 3.6.1.3-10 Revision 43

Suppression Pool Water Level B 3.6.2.2 BASES (continued)

APPLICABLE Initial suppression pool water level affects suppression SAFETY ANALYSES pool temperature response calculations, calculated drywell pressure for a DBA, calculated pool swell loads for a DBA LOCA, and calculated loads due to relief valve discharges.

Suppression pool water level must be maintained within the limits specified so that the safety analysis of Reference 1 remains valid.

Suppression pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).

LCO A limit that suppression pool water level be 14 ft 1 inch and 14 ft 5 inches above the bottom of the suppression chamber 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 the downcomers are covered, HPCI and RCIC turbine exhausts are covered, and relief valve 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 RHR Suppression Pool 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.

(continued)

Quad Cities 1 and 2 B 3.6.2.2-2 Revision 00

Secondary Containment B 3.6.4.1 BASES APPLICABLE The secondary containment performs no active function in SAFETY ANALYSES response to this limiting event; however, its leak tightness (continued) is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission products entrapped within the secondary containment structure will be treated by the SGT System prior to discharge to the environment.

Secondary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

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 ).

(continued)

Quad Cities 1 and 2 B 3.6.4.1-2 Revision 54

Secondary Containment B 3.6.4.1 BASES ACTIONS C.1 and C.2 (continued)

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 this activity 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

Quad Cities 1 and 2 B 3.6.4.1-4 Revision 43

SCIVs B 3.6.4.2 BASES APPLICABLE established. The accident for which the secondary SAFETY ANALYSES containment boundary is required is a loss of coolant (continued) accident (Ref. 1). The secondary containment performs no active function in response to this limiting event, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

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 the Technical Requirements Manual (Ref. 2).

The normally closed manual SCIVs are considered OPERABLE when the valves are closed and blind flanges are in place, or open under administrative controls. These passive isolation valves or devices are listed in Reference 2.

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. 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 ).

(continued)

Quad Cities 1 and 2 B 3.6.4.2-2 Revision 31

SCIVs B 3.6.4.2 BASES ACTIONS B.1 (continued)

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 this activity 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.

Required Action D.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 fuel 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.

(continued)

Quad Cities 1 and 2 B 3.6.4.2-5 Revision 31

SGT System B 3.6.4.3 BASES LCO releases. Meeting the LCO requirements for two OPERABLE (continued) subsystems ensures operation of at least one SGT subsystem in the event of a single active failure. OPERABILITY of a subsystem also requires the associated cooling air damper remain OPERABLE.

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 ).

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 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 overall plant risk is minimized. 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 .

(continued)

Quad Cities 1 and 2 B 3.6.4.3-3 Revision 40

SGT System B 3.6.4.3 BASES ACTIONS B.1 (continued)

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 8) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action B.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.

However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met. The allowed Completion Time is 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, 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 this activity must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend (continued)

Quad Cities 1 and 2 B 3.6.4.3-4 Revision 40

SGT System B 3.6.4.3 BASES ACTIONS C.1, C.2.1, and C.2.2 (continued)

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.

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 correct the problem that is commensurate with the importance of supporting the required radioactivity release control radioactivity release control function. Therefore, one SGT subsystem must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time provides a period of time to function in MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring the SGT System) occurring during periods where the required radioactivity release control function may not be maintained is minimal.

E.1 If one 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 overall plant risk is minimized. 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 . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk of MODE 4 (Ref. 8) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

(continued)

Quad Cities 1 and 2 B 3.6.4.3-5 Revision 40

SGT System B 3.6.4.3 BASES ACTIONS E.1. (continued)

Required Action E.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.

However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

F.1 and F.2 When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of this activity 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 F.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.

(continued)

Quad Cities 1 and 2 B 3.6.4.3-6 Revision 40

CREV System B 3.7.4 BASES LCO The LCO is modified by a Note allowing the CRE boundary to (continued) 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 the CRE isolation is indicated.

APPLICABILITY In MODES 1, 2, and 3, the CREV System must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA, since the DBA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREV System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. Dduring movement of recently irradiated fuel assemblies in the secondary containment; and

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

Due to radioactive decay, the CREV 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 ).

ACTIONS A.1 With the CREV System inoperable for reasons other than an inoperable CRE boundary in MODE 1, 2, or 3, the inoperable CREV System must be restored to OPERABLE status within 7 days. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period.

(continued)

Quad Cities 1 and 2 B 3.7.4-4 Revision 34

CREV System B 3.7.4 BASES ACTIONS (continued) D.1 and D.2 In MODE 1, 2, or 3, if the inoperable CREV System or the CRE boundary cannot be restored to OPERABLE status within the required 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.

E.1 and E.2 LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel movement can occur in MODE 1, 2, or 3, 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 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, but would not require immediate suspension of movement of recently irradiated fuel assemblies. The NOTE to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

With the CREV System inoperable, during movement of recently irradiated fuel assemblies in the secondary containment, or during OPDRVs or with the CREV System 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. This places the unit in a condition that minimizes accident risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of this activity shall not preclude completion of movement of a component to a safe position.

Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

(continued)

Quad Cities 1 and 2 B 3.7.4-7 Revision 40

Control Room Emergency Ventilation AC System B 3.7.5 BASES APPLICABILITY emergency zone temperature will not exceed equipment (continued) OPERABILITY limits following control room emergency zone isolation.

In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room Emergency Ventilation AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. Dduring movement of recently irradiated fuel assemblies in the secondary containment; and

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

Due to radioactive decay, the Control Room Emergency Ventilation AC 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 ).

ACTIONS A.1 With the Control Room Emergency Ventilation AC System inoperable in MODE 1, 2, or 3, the system must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on the low probability of an event occurring requiring control room emergency zone isolation and the availability of alternate nonsafety cooling methods.

B.1 In MODE 1, 2, or 3, if the inoperable Control Room Emergency Ventilation AC System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes overall plant 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 . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 4) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be (continued)

Quad Cities 1 and 2 B 3.7.5-3 Revision 40

Control Room Emergency Ventilation AC System B 3.7.5 BASES ACTIONS B.1 (continued) short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action B.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.

However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met. The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.1 and C.2 LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel movement can occur in MODE 1, 2, or 3, the Required Actions of Condition C 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. Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of recently irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

With the Control Room Emergency Ventilation AC System inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, action must be taken immediately to suspend activities that (continued)

Quad Cities 1 and 2 B 3.7.5-4 Revision 40

Control Room Emergency Ventilation AC System B 3.7.5 BASES ACTIONS C.1 and C.2 (continued) 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 this activity shall not preclude completion of movement of a component to a safe position.

Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action 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 emergency zone heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. UFSAR, Section 9.4.

2. UFSAR, Section 9.1.4.3.2.

3. NRC Safety Evaluation Report for the Holtec International HI-STORM 100 Storage System (Docket Number 72-1014, Certificate Number 1014, Amendment 2).

4. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

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AC SourcesShutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC SourcesShutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC SourcesOperating."

Movement of a Spent Fuel Cask containing Spent Nuclear Fuel in a sealed Multi-Purpose Canister (MPC) and using a single failure-proof crane is not considered to be "movement of irradiated fuel assemblies in secondary containment" (Refs.

1 and 2).

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES 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 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 mitigate 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 ).

In general, when the unit is shutdown 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 (continued)

Quad Cities 1 and 2 B 3.8.2-1 Revision 31

AC SourcesShutdown B 3.8.2 BASES (continued)

LCO One offsite circuit supplying the onsite Class 1E power distribution subsystem(s) of LCO 3.8.8, "Distribution SystemsShutdown," ensures that all required loads are powered from offsite power. An OPERABLE DG, associated with a Distribution System Essential Service System (ESS) 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 DG 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 reactor vessel draindown).

The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective ESS bus(es), and of accepting required loads during an accident. Qualified offsite circuits are those that are described in the UFSAR and are part of the licensing basis for the unit. The offsite circuit from the 345 kV switchyard consists of the incoming breakers and disconnects to the 12 or 22 reserve auxiliary transformer (RAT), associated 12 or 22 RAT, and the respective circuit path including feeder breakers to 4160 kV ESS buses required by LCO 3.8.8. Another qualified circuit is provided by the bus tie between the corresponding ESS buses of the two units.

The required DG must be capable of starting, accelerating to rated speed and voltage, connecting to its respective 4160 V ESS bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 13 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the 4160 V ESS buses. These capabilities are required to be met from a variety of initial conditions such as DG in standby with engine hot and DG in standby with engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillances. Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. The necessary portions of the DG Cooling Water System capable of providing cooling to the required DG is also required.

(continued)

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AC SourcesShutdown B 3.8.2 BASES LCO It is acceptable for divisions to be cross tied during (continued) shutdown conditions, permitting a single offsite power circuit to supply all required divisions.

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 providing adequate coolant inventory makeupthat provide core cooling 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 (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 ) 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 can occur in MODE 1, 2, or 3, 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, 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, but would not require immediate suspension of movement of recently irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.1 An offsite circuit is considered inoperable if it is not (continued)

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AC SourcesShutdown B 3.8.2 BASES ACTIONS A.1 (continued) available to one required ESS 4160 V ESS bus. If two or more 4160 V ESS 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 that are not powered from offsite power, appropriate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS. Required features remaining powered from a qualified offsite circuit, even if that circuit is considered inoperable because it is not powering other required features, are not declared inoperable by this Required Action. For example, if both Division 1 and 2 ESS buses are required OPERABLE by LCO 3.8.8 and only the Division 1 ESS buses are not capable of being powered from offsite power, then only the required features powered from Division 1 ESS buses are required to be declared inoperable.

A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, and B.3, and B.4 With the required offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable per Required Action A.1. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made.

With the required DG 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 sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.

(continued)

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AC SourcesShutdown B 3.8.2 BASES ACTIONS A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, and B.3, and B.4 (continued)

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources 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 ESS 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 to be applicable. SR 3.8.1.9 is not required to be met since only one offsite circuit is required to be OPERABLE. SR 3.8.1.20 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE. SR 3.8.1.21 is not required to be met because the opposite unit's DG is not required to be OPERABLE in MODES 4 and 5, and during movement of recently irradiated fuel assemblies in secondary containment. 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 DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and to preclude de-energizing a required 4160 V ESS bus or disconnecting a required offsite circuit during performance (continued)

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AC SourcesShutdown B 3.8.2 BASES SURVEILLANCE SR 3.8.2.1 (continued)

REQUIREMENTS of SRs. With limited AC sources available, a single event could compromise both the required circuit and the DG. It is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the DG and offsite circuit are required to be OPERABLE.

Note 2 states that SRs 3.8.1.13 and 3.8.1.19 are not required to be met when its associated ECCS subsystem(s) are not required to be OPERABLE. These SRs demonstrate the DG response to an ECCS initiation signal (either alone or in conjunction with a loss of offsite power signal). This is consistent with the ECCS instrumentation requirements that do not require the ECCS initiation signals when the associated ECCS subsystem is not required to be OPERABLE per LCO 3.5.2, "RPV Water Inventory ControlECCSShutdown."

REFERENCES 1. UFSAR, Section 9.1.4.3.2.

2. NRC Safety Evaluation Report for the Holtec International HI-STORM 100 Storage System (Docket Number 72-1014, Certificate Number 1014, Amendment 2).

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DC SourcesShutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC SourcesShutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC SourcesOperating."

Movement of a Spent Fuel Cask containing Spent Nuclear Fuel in a sealed Multi-Purpose Canister (MPC) and using a single failure-proof crane is not considered to be "movement of irradiated fuel assemblies in secondary containment" (Refs.

3 and 4).

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation and during movement of recently irradiated fuel assemblies in the secondary containment.

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 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 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 power is only (continued)

Quad Cities 1 and 2 B 3.8.5-1 Revision 31

DC SourcesShutdown B 3.8.5 BASES LCO The DC electrical power subsystems with: a) the required 250 VDC subsystem consisting of one 250 VDC battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus; and b) the required 125 VDC subsystem consisting of one battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus are required to be OPERABLE to support some of the required DC distribution subsystems required OPERABLE by LCO 3.8.8, "Distribution SystemsShutdown." 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). The associated alternate 125 VDC electrical power subsystem may be used to satisfy the requirements of the 125 VDC subsystem.

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 adequate core cooling 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;

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.

Due to radioactive decay, DC electrical power is only required to mitigate 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 ).

(continued)

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DC SourcesShutdown B 3.8.5 BASES APPLICABILITY The DC electrical power requirements for MODES 1, 2, and 3 (continued) 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 can occur in MODE 1, 2, or 3, 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, 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, but would not require immediate suspension of movement of recently irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.1, A.2.1, A.2.2, and A.2.3, and A.2.4 By allowance of the option to declare required features inoperable with associated DC electrical power subsystem(s) inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. However, 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 DC electrical power subsystems 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 (continued)

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DC SourcesShutdown B 3.8.5 BASES ACTIONS A.1, A.2.1, A.2.2, and A.2.3, and A.2.4 (continued) restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8 to be applicable. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of each SR.

This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE 250 VDC source from being discharged below their capability to provide the required power supply or otherwise rendered inoperable during the performance of SRs. It is the intent that these SRs must still be capable of being met, but actual performance is not required.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15.

3. UFSAR, Section 9.1.4.3.2.

4. NRC Safety Evaluation Report for the Holtec International HI-STORM 100 Storage System (Docket Number 72-1014, Certificate Number 1014, Amendment 2).

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Distribution Systems-Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution SystemsShutdown BASES BACKGROUND A description of the AC and DC electrical power distribution systems is provided in the Bases for LCO 3.8.7, "Distribution SystemsOperating."

Movement of a Spent Fuel Cask containing Spent Nuclear Fuel in a sealed Multi-Purpose Canister (MPC) and using a single failure-proof crane is not considered to be "movement of irradiated fuel assemblies in secondary containment" (Refs.

3 and 4).

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature (ESF) 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, Reactor Coolant 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.

(continued)

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Distribution Systems-Shutdown B 3.8.8 BASES APPLICABLE Due to radioactive decay, AC and DC electrical power is only SAFETY ANALYSES required to mitigate fuel handling accidents involving (continued) 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 ).

The AC and DC electrical power distribution systems satisfy 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 features. This LCO explicitly requires energization of the portions of the electrical distribution system, including the opposite unit electrical distribution systems, necessary to support OPERABILITY of Technical Specifications required systems, 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 thatto provide adequate core coolingcoolant 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 (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 ) are available; (continued)

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Distribution Systems-Shutdown B 3.8.8 BASES APPLICABILITY c. Systems necessary to mitigate the effects of events (continued) 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 can occur in MODE 1, 2, or 3, 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, 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, but would not require immediate suspension of movement of recently irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

A.1, A.2.1, A.2.2, A.2.3, and A.2.4, and A.2.5 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 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, (continued)

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Distribution SystemsShutdown B 3.8.8 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, and A.2.4, and A.2.5 (continued) 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 required 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

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Inservice Leak and Hydrostatic Testing Operation B 3.10.8 BASES BACKGROUND Removal of heat addition from recirculation pump operation (continued) and reactor core decay heat is coarsely controlled by Control Rod Drive Hydraulic System flow and Reactor Water Cleanup System non-regenerative heat exchanger operation.

Test conditions are focused on maintaining a steady state pressure, and tightly limited temperature control poses an unnecessary burden on the operator and may not be achievable in certain instances.

A hydrostatic and/or system leakage test is performed at operating pressure on the primary system. Scram time testing, controlled by TS 3.1.4 and TS 3.10.3, is typically scheduled in parallel with these tests.

Other testing may be performed in conjunction with the allowances for inservice leak or hydrostatic tests and control rod scram time test.

APPLICABLE Allowing the reactor to be considered in MODE 4 when the SAFETY ANALYSES reactor coolant temperature is > 212°F, during, or as a consequence of, 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 (ECCS). 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. 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 any large primary system leak that (continued)

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Inservice Leak and Hydrostatic Testing Operation B 3.10.8 BASES APPLICABLE could result in draining of the RPV, the reactor SAFETY ANALYSES vessel would rapidly depressurize, allowing the low pressure (continued) 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 these tests, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and during postulated accident conditions.

As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation at reactor coolant temperatures > 212°F, can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures

> 212°F, performance of inservice leak and hydrostatic testing would also necessitate the inoperability of some subsystems normally required to be OPERABLE when > 212°F.

Additionally, even with required minimum reactor coolant temperatures 212°F, RCS temperatures may drift above 212°F during the performance of inservice leak and hydrostatic testing or during subsequent control rod scram time testing, which is typically performed in conjunction with inservice leak and hydrostatic testing. While this Special Operations LCO is provided for inservice leak and hydrostatic testing, and for scram time testing initiated in conjunction with an inservice leak or hydrostatic test, parallel performance of other tests and inspections is not precluded.

(continued)

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v t e Letter Sequence Request
EPID:L-2018-LLA-0049, TSTF-542 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other: ML18254A430
MONTHYEAR2018-03-23 [Table View]

RS-18-006, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.

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RS-18-006, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.

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3.0 REGULATORY ANALYSIS

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