Text

Submittal of Application to Revise Technical Specifications to Adopt TSTF-542, Revision 2, Reactor Pressure Vessel Water Inventory Control
	ML17228A042
Person / Time
Site:	Cooper
Issue date:	08/07/2017
From:	Dent J; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NLS2017048
	Download: ML17228A042 (183)
	v • d • e

H Nebraska Public Power District Always there when you need us 50.90 NLS2017048 August 7, 2017 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control" Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Nebraska Public Power District (NPPD) is submitting a request for an amendment to the Technical Specifications (TS) for Cooper Nuclear Station (CNS). The proposed change replaces existing TS requirements related to "operations with a potential for draining the reactor vessel" with new requirements on Reactor Pressure Vessel Water Inventory Control 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.

NPPD requests Nuclear Regulatory Commission (NRC) approval of the proposed TS change and issuance of the requested license amendment by August 7, 2018. Once approved, the amendment shall be implemented within 60 days. provides a description and assessment of the proposed TS 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.

This proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Renewed Facility Operating License through Amendment 259 dated June 20, 2017, have been incorporated into this request.

This request is submitted under oath pursuant to 10 CFR 50.30(b).

By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.9l(b)(l). Copies to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance with 10 CFR 50.4(b)(l).

This letter contains no regulatory commitments.

COOPER NUCLEAR STATION PO Box 98 j Brownville, NE 68321*0098 Telephone: {402) 825*3811 / Fax: (402) 825*5211 www.nppd.com

NLS2017048 Page 2 of2 Should you have any questions concerning this matter, please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

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

Executed On: --~~/_7~/_2.._0_1~7__

Date Sincerely, () ~

~*'~T)

Vice President - Nuclear and Chief Nuclear Officer

/dv Attachments: 1. Description and Assessment of Technical Specifications Changes

2. Proposed Technical Specifications Changes (Mark-up)

3. Revised Technical Specifications Pages

4. Proposed Technical Specifications Bases Changes (Mark-up) -

Information Only cc: Regional Administrator w/ attachments USNRC - Region IV Cooper Project Manager w/ attachments USNRC - NRR Plant Licensing Branch IV Senior Resident Inspector w/ attachments USNRC-CNS Nebraska Health and Human Services w/ attachments Department of Regulation and Licensure NPG Distribution w/o attachments CNS Records w/ attachments

NLS2017048 Page 1 of 9 Attachment 1 Description and Assessment of Technical Specifications Changes Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 1.0 Description 2.0 Assessment 2.1 Applicability of Published Safety Evaluation 2.2 Variations 2.3 Editorial Corrections 3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Analysis 4.0 Environmental Evaluation

NLS2017048 Page 2of9

1.0 DESCRIPTION

The proposed change replaces existing Technical Specifications (TS) requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control 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 Nebraska Public Power District (NPPD) has reviewed the safety evaluation provided to the Technical Specifications Task Force (TSTF) on December 20, 2016, as well as the information provided in TSTF-542. NPPD has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the Nuclear Regulatory Commission (NRC) staff are applicable to Cooper Nuclear Station (CNS) and justify this amendment for the incorporation of the changes to the CNS TS.

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

3.3 .5.1 Emergency Core Cooling System (ECCS) Instrumentation 3.3.6.1 Primary Containment Isolation Instrumentation 3.3.6.2 Secondary Containment Isolation Instrumentation 3.3.7.1 Control Room Emergency Filter (CREF) System Instrumentation 3.5.2 ECCS - Shutdown 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 3.6.4.1 Secondary Containment 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) 3.6.4.3 Standby Gas Treatment (SGT) System 3.7.4 Control Room Emergency Filter (CREF) System 3.8.2 AC Sources - Shutdown 3.8.5 DC Sources - Shutdown 3.8.8 Distribution Systems - Shutdown 2.2 Variations NPPD is proposing the following variations from the TS changes described in the TSTF-542 or the applicable parts of the NRC staffs safety evaluation. These variations do not affect the applicability of TSTF-542 or the NRC staffs safety evaluation to the proposed license amendment.

NLS2017048 Page 3 of9 2.2.1 The CNS 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 CNS TS .

STS Table 3.3.5.1-1 , Function 1.c, 2.c, and 2.d, in part, are titled Reactor Steam Dome Pressure - Low. In CNS TS, Function 1.c, 2.c, and 2.d, in part, are titled Reactor Pressure - Low.

STS section 3.3.7.1 is titled Main Control Room Environmental Control System Instrumentation. The equivalent section in CNS TS is titled Control Room Emergency Filter System Instrumentation.

STS section 3.7.4 is titled Main Control Room Environmental Control System. The equivalent section in CNS TS is titled Control Room Emergency Filter System.

STS Limiting Condition for Operation (LCO) 3.8.10 is titled Distribution Systems -

Shutdown. The equivalent CNS TS section is numbered 3.8.8.

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). CNS was not licensed to the l 0 CFR 50, Appendix A, GDC. CNS was designed and constructed to meet the principle design criteria described in the Atomic Energy Commission's (AEC) proposed rule, "General Design Criteria for Nuclear Power Plant Construction Permits," published in the Federal Register on July 11, 1967 (32 FR 10213). The degree of conformance to the 1967 proposed GDC is described in Appendix F, "Conformance to AEC Proposed General Design Criteria" to the Updated Safety Analysis Report for CNS.

CNS' current licensing basis incorporates the proposed GDC that are equivalent to the 10 CFR Part 50, Appendix A, GDC 13, 14, 30 and 33 . The proposed license change is consistent with the AEC proposed GDC in that the design requirements for instrumentation, reactor coolant leak detection, the reactor coolant pressure boundary, and reactor coolant makeup are unaffected.

2.2.3 The CNS TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement (SR) Frequencies for Specifications 3.3.5.3 and 3.5.2 are "In accordance with the Surveillance Frequency Control Program." This variation is editorial in nature.

2.2.4 NPPD has chosen to implement the Reactor Pressure Vessel Water Inventory Control Instrumentation specification as TS 3.3.5.3 and to not renumber the existing TS 3.3.5.2.

This variation is editorial in nature.

2.2.5 In Tables 3.3.5.1-1 , 3.3.6.2-1and3.3.7.1-1 , NPPD has chosen not to remove Note (a) and re-letter the subsequent notes as shown in TSTF-542, but instead to replace the Note (a) text with "Deleted" and leave the subsequent note designations, as is. This variation is editorial in nature.

NLS2017048 Page 4 of9 2.2.6 The CNS TS contain a NOTE in SR 3.5.2.3 regarding realignment to the Low Pressure Coolant Injection mode that is the same as the NOTE in the STS LCO 3.5.2. NPPD will relocate the NOTE from the SR to the LCO section. This has no effect on the adoption of the TSTF-542 and increases consistency between the CNS TS and STS. This variation is editorial in nature.

2.2.7 Current CNS TS SR 3.5.2.1 (new SR 3.5.2.2) verifies sufficient suppression pool water level for required ECCS injection/spray subsystems. The CNS SR is a combination of the TSTF-542 SR 3.5.2.2 and SR 3.5.2.3, which are for Low Pressure Coolant Injection (LPCI) and Core Spray (CS), respectfully. In addition, the option for using a condensate storage tank as a make-up source in TSTF-542 SR 3.5.2.3 does not exist in the CNS SR.

This option was removed in CNS TS Amendment 252. These differences do not alter the conclusion that the proposed change is applicable to CNS. This variation is editorial in nature.

2.2.8 TSTF-542, Table 3.3.5.2-1 , Function 2b, Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass), uses "l per pump" for the Required Channels per Function. NPPD proposes the use of "subsystem" in place of "pump" as more appropriate since the instrument is associated with a LPCI subsystem that has two pumps, rather than an individual pump. This variation is editorial in nature.

2.2 .9 NPPD proposes to delete CNS TS 3.6.1.3 , Condition F, and both of its associated Required Actions. The Applicability for TS 3 .6.1.3, "Primary Containment Isolation Valves (PCIVs)," is "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 3.3.5.3. Thus, there are no longer any PCIVs required to be operable by TS 3.6.1.3 during Modes 4 or 5.

These requirements are addressed by the proposed TS 3.3 .5.3 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. This variation is technical in nature and similar to variations described in the license amendment requests submitted for Dresden Nuclear Power Station and Clinton Power Station.

2.2.10 The STS contain requirements that differ from the CNS TS.

There are STS requirements on which TSTF-542 is based, related to "manual initiation," that do not appear in the CNS TS . STS Table 3 .3 .5.1-1 contains Functions 1.e and 2.h, Manual Initiation, for CS and LPCI, respectfully. The "manual initiation" logic does not exist in the CNS design. These functions, as well as the related TSTF-542 surveillance requirements, SR 3.3.5.2.3 and SR 3.5.2.8, do not apply to CNS.

NLS2017048 Page 5 of9 As an alternative, NPPD proposes that TS 3.5.2, "Reactor Pressure Vessel (RPV)

Water Inventory Control," include an SR 3.5.2.7 to verify that the CNS required ECCS injection/spray subsystem can be manually operated through the manipulation of subsystem components from the Main Control Room.

The manual operation of the required ECCS injection/spray subsystem for the control ofreactor cavity or RPV inventory is a relatively simple evolution and involves 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, which is assumed to occur at full power, and thus, there is adequate time to take manual actions (i.e., hours versus minutes).

Adequate time to take action is assured since the proposed TS 3.5.2, Condition E, prohibits plant conditions that result in drain times that are less than one hour.

Therefore, there is sufficient time for the licensed operators to take manual action to stop an unanticipated draining event, and to manually start an ECCS injection/spray subsystem or the additional method of water injection.

Since the ECCS injection/spray subsystem can be placed in service using manual means in a short period of time (i.e., within the time frames 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.

Current SR 3.5.1.6 and SR 3.5.2.4 manually operate the ECCS injection/spray pumps to verify each required ECCS injection/spray pump develops the specified flow rate against a system head corresponding to the specified reactor pressure at a frequency specified by the lnservice Testing (IST) Program. The IST Program requires the ECCS injection/spray subsystems motor operated injection valves, minimum flow valves and test flow path valves (with the exception of the CS test flow path valves) be cycled to demonstrate operability and compliance with IST stroke time requirements at a frequency specified by the IST Program. The CS test flow path valves are part of the IST Program but do not have stroke time requirements. The CS valves are cycled for position indication verification only. The manual operation of the ECCS injection/spray subsystem to demonstrate operability required by the proposed SR 3.5.2.7 is equivalent to the testing that is presently required to be performed on the ECCS injection/spray subsystems.

This variation is technical in nature due to plant design differences and similar to variations described in the license amendment requests submitted for Dresden Nuclear Power Station and Edwin I. Hatch Nuclear Plant.

CNS does not have the capability to perform Channel Checks for proposed Table 3.3.5.3 -1, Functions 1.a, "Reactor Pressure - Low (Injection Permissive)," 1.b, "Core Spray Pump Discharge Flow - Low (Bypass)," 2.a, "Reactor Pressure - Low

NLS2017048 Attachment 1 Page 6of9 (Injection Permissive)," and 2.b, "Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass)." The current CNS TS do not include Channel Checks for these functions; therefore, no Channel Check Surveillance Requirement was added for these functions. This variation is technical in nature and similar to a variation described in the license amendment request submitted for Dresden Nuclear Power Station.

The CNS TS do not include LCO 3.7.5, Control Room Air Conditioning System, or LCO 3.8.8, Inverters - Shutdown, as shown on the STS pages ofTSTF-542. There will be no corresponding change to the CNS TS. This variation is editorial in nature.

2.2.11 The CNS TS contain requirements that differ from the STS on which TSTF-542 was based, but are encompassed in the TSTF-542 justification:

CNS TS Table 3.3.5.1-1 contains Function 1.e, "Core Spray Pump Start -Time Delay Relay," that does not appear in the STS table. The Function is required to be operable in Modes 1, 2, 3, 4 and 5. Modes 4 and 5 are being deleted from this Function as this is related to automatic ECCS initiation. This is the same justification as provided in the TSTF for STS Table 3.3.5.1-1, Function 2.f, "Low Pressure Coolant Injection Pump Start - Time Delay Relay." This variation is technical in nature and justified by the discussion in Section 3.4.1 of the TSTF-542 justification.

CNS TS Table 3.3.6.2-1 , "Secondary Containment Isolation Instrumentation," and Table 3.3.7.1-1, "Control Room Emergency Filter System Instrumentation," contain different instrumentation functions than contained in the equivalent STS tables. The TSTF change to the tables is the deletion of the applicability of OPDRVs. An equivalent change is made to the functions in the CNS tables. This variation is editorial in nature.

2.3 Editorial Corrections An editorial error was introduced with the approval and implementation of Amendment 258. In the Table of Contents, Sections 3.10 and 3.10.4 are shown to start on pages 3.10.1 and 3.10.9, respectively. These should be identified as pages 3.10-1 and 3.10-9. Marked up and corrected pages are included in this license amendment request.

It was noted that a period was needed at the end of LCO 3.3. 7.1, Applicability.

In the Table of Contents, the word "continued" was added in the appropriate locations at the top and bottom of the pages.

NLS2017048 Page 7 of9

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Nebraska Public Power District (NPPD) requests adoption ofTSTF-542 "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications, into the Cooper Nuclear Station 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. The editorial changes are purely administrative, so do not impact the analysis below.

NPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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 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 (T AF). 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

NLS2017048 Page 8 of9 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 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 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.

NLS2017048 Page 9 of9 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 T AF 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, NPPD concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92( c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22( c)(9).

Therefore, pursuant to 10 CFR 5 l .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

NLS2017048 Page 1of42 Attachment 2 Proposed Technical Specifications Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 TOC page i 3.3-52 3.6-36 TOC page ii 3.3-56 3.6-38 TOC page iii 3.3-60 3.6-40 1.1-3 3.3-64 3.6-41 1.1-3 insert 3.3-66 3.6-42 3.3-31 3.5-1 3.7-8 3.3-32 3.5-7 3.7-9 3.3-33 3.5-8 3.8-11 3.3-37 3.5-9 3.8-12 3.3-38 3.5-10 3.8-20 3.3-39 3.5-12 3.8-21 3.3-47 3.6-11 3.8-29 3.3-48 3.6-34 3.8-30 3.3-49 3.6-35

TABLE OF CONTENTS 1.0 USE AND APPLICATION ............................................ ..... ................................. 1.1-1 1.1 Definitions ......... .. .......... .. ....................................................................... 1.1-1 1.2 Logical Connectors ............ ... .................................. ............... .. .. .. .. ........ 1.2-1 1.3 Completion Times .............................. .............. ... .............................. ..... 1.3-1 1.4 Frequency .. ........ .... .. .............................................................................. 1.4-1 2.0 SAFETY LIMITS (SLs) ........................... .. ......................................................... 2.0-1 2.1 SLs .... ........... .......... .... ....................... ... .. ..................... ......... ....... ......... . 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 3.1.1 SHUTDOWN MARGIN (SOM) .. ... ......... ......................... ........................ 3.1-1 3.1.2 Reactivity Anomalies ... ....... ................. .. ..... ... .... ....... ..... ................... ...... 3 .1-5 3.1.3 Control Rod OPERABILITY .. ... .. .. ....... ..... .... ..... ........... .. ........................ 3.1-7 3.1.4 Control Rod Scram Times .. ...... ......... .... .... .. ................... ....................... . 3.1- 12 3.1.5 Control Rod Scram Accumulators .... .... .... ............ ... ....... ..................... ... 3.1-15 3.1.6 Rod Pattern Control .. ... ..... .... .. .... .... .. ...... ...... ...... ..... .. ... ........................ 3.1-18 3.1.7 Standby Liquid Control (SLC) System ............... .... ....... .. ........................ 3. 1-20 3.1 .8 Scram Discharge Volume (SOV) Vent and Drain Valves ........................ 3.1-25 3.2 POWER DISTRIBUTION LIMITS ... ...... ........ ......... ... ...................... .......... .... 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ..... .. ....... ... ....................... ...... ......... .. ....... .. .......... .... ....... 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ....... ..... ................. ......... 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ................. .. .................... 3.2-4 3.3 INSTRUMENTATION ........... ..... ...... ........ ...... .... .... ....... ................... ... .... .... . 3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation .. ........................ .. ..... 3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation ... ... ................................. 3.3-9 3.3.2.1 Control Rod Block lnstrumentation ................... .. .... ................................ 3.3-14 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation ........... ............. ............ ... ..... ....... ..... ................ .. .. 3.3-20 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation .. . ... .... .. .. ... ................ 3.3-22 3.3.3.2 Alternate Shutdown System ........ .... ..... ...... .... ....................................... . 3.3-26 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ......... ........................ .... 3.3-28 3.3.5.1 Emergency Core Cooling System (ECCS) 3.3.5.2 React~~sg~r~~~~~~~f~n* c~~,i~g* ('Rc"1'c). s'y~t~~ ...................................... 33 1

-~ 50 Instrumentation .. ............ ... ... ... .. ....... .... ....... ......... ........ ..... ........... 3 . 3-~ 3 57 3.3.6.1 Primary Containment Isolation Instrumentation .... .................................. 3 . 3- ~ 1 3.3.6.2 Secondary Containment Isolation Instrumentation .. .............................. . 3.3-3.3.6.3 Low-Low Set (LLS) Instrumentation .. ....... ... ....... .. .. .............. .................. 3.3-# ~

Contro~~~Z~ 1~~~~~een~~t~~I~~~ ~~~~:.~ 3 _ 3 -~

3.3.7.1 3.3.8.1 Loss of Power (LOP) Instrumentation ......... .. .. ..... ................................ .. 3.3 r4 70 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ..... ..... ... ........ .... ..... ...... ... ............. ...... ................................ 3.3-Coo er i Amendment No. 258---

3.3 .5.3 Reactor Pressure Vessel (RPV) Water Inventory Control lnstrumentation ......... ..... ..... ..... ... .. ....... 3.3-47

TABLE OF CONTENT~

3.4 REACTOR COOLANT SYSTEM (RCS) ..... ... .... ........ ....... .... .. .. .. .. ......... .... ... 3.4-1 3.4.1 Recirculation Loops Operating ..... ..... ....... .. ......... .. ...... ... .... ......... ....... .... 3.4-1 3.4 .2 Jet Pumps ... .. ...... .... .... ..... ... ... ........ ..... ..... .......... ........ .... .... ........ ..... .... ... 3.4-4 3.4.3 Safety/Relief Valves (SRVs) and Safety Valves (SVs) ............. .. ............ 3.4-6 3.4.4 RCS Operational LEAKAGE .. ........ ... ........ ....... ...... .... ..... ...... .... ......... .... 3.4-8 3.4.5 RCS Leakage Detection Instrumentation ........ ... ...... .. .. .. ...... ..... .. .. .. ..... .. 3.4-10 3.4.6 RCS Specific Activity ......... .................. ..... .. ... ............... .... ..... ... .... .... ...... 3.4-12 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown .... .... .... ... ....... ... ....... ... ..... ...... ...... .. ...... .. .. 3.4-14 3.4 .8 Residual Heat Removal (RHR) Shutdown Cooling

System - Cold Shutdown .. .. ........... ....... .. .... .... .. .................... .. ..... . 3.4-17 3.4.9 RCS Pressure and Temperature (PIT) Limits ..... .. .... ...... .. .......... .. ... ....... 3.4-19 3.4 .10 R ..... ... .. .. .. ..... ................. ... ... 3.4-23

, RPV WATER INVENTORY CONTROL, 3.5 EMERGENCY CORE COOLING SYSTEMS (ECC AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ........ .... .. ..... ......... ....... .. ... ........ .. . 3.5-1 3.5.1 ECCS - Operating ....... . .. ...... .. 3.5-1 3.5.2 EGGS Sh1:1telewn .(. ... JRPV Water Inventory Control 1.. ..... ... 3.5-7 3.5.3 RCIC System .......... .. ... .... ..... ... .. .......... .... ... ... ......... ............. ....... .. .... .. ... 3 . 5- ~

3.6 CONTAINMENT SYSTEMS .... .. ..... ....... ...... ... .... ... ..... ........................ ......... . 3.6-~ ~

3.6.1.1 Primary Containment ..... ..... ... ..... ...... .... ...... .. .... ... ... ............ .. ..... .. ..... ..... 3.6-1 3.6.1.2 Primary Containment Air Lock .... .... .......... .. ....... ....... .. .. .. .............. .......... 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ..... ..... ........... .. ............... 3.6-8 3.6.1.4 Drywell Pressure .... .............. ............ .. .................. ... ...... ... .... ... ... .. .. .. .. .... 3.6-16 3.6.1.5 Drywell Air Temperature .... ..... ... ... .. .... ... ....... ... ... ................................... 3.6-17 3.6.1.6 Low-Low Set (LLS) Valves ............. ... ............ .. ...... .... .......... .. ... ... .......... . 3.6-18 3.6.1 .7 Reactor Building-to-Suppression Chamber Vacuum Breakers .. .. ... ... ...... .... ... ............... ..... .... ..... .. ...... .. ....... ..... ............. 3.6-20 3.6.1.8 Suppression-Chamber-to-Drywell Vacuum Breakers ....... ... ...... ........... .. 3.6-23 3.6.1.9 Residual Heat Removal (RHR) Containment Spray ... ... .... .... ... .............. 3.6-25 3.6.2 .1 Suppression Pool Average Temperature .. ........ .. ... .. ........ .... .. ..... ....... ..... 3.6-27 3.6.2.2 Suppression Pool Water Level .. .. .... ........ ... ...... ..... ... ............. ..... ............ 3.6-30 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ....... .. ... ... .... .. ..... ....... ... ... ........... ......... ... ..... ... .. .... ........ ..... 3.6-31 3.6 .3.1 Primary Containment Oxygen Concentration ..... ...... ... ... ......... ...... .. ....... 3.6-33 3.6.4 .1 Secondary Containment. ... ... .............. .... ... ..... .............. .. .. ........ .... .. .... .... 3.6-34 3.6.4 .2 Secondary Containment Isolation Valves (SCIVs) .... .. ............. ... ..... ....... 3.6-36 3.6.4.3 Standby Gas Treatment (SGT) System ........... ........ ... ....... ........ .......... ... 3.6-40 3.7 PLANT SYSTEMS .. ..................... .... ........ ................ ... ......... ...... ... ... ... .... ..... 3.7-1 3.7 .1 Residual Heat Removal Service Water Booster (RHRSWB)

System ............. ....... ...... .... ........... ............. ........... ....... .... .. .. .... .. ... 3.7-1 3.7.2 Service Water (SW) System and Ultimate Heat Sink (UHS) .. ........ .................................................. ............... 3.7-3 3.7.3 Reactor Equipment Cooling (REC) System .. ......... .. ..... .. ...... .... .. ......... .. . 3.7-6 3.7.4 Control Room Emergency Filter (CREF) System ............. .... .. .... ... .... .... . 3.7-8 3.7.5 Air Ejector Offgas .. ...... .. ..... ..... .. ........... .... ... .. .... .... ......... .. .. .. ............. ..... 3.7-11 3.7.6 Spent Fuel Storage Pool Water Level .. ... ..... .. ...... .. ........ .... ... .. ... ............ 3.7-13 3.7.7 The Main Turbine Bypass System ...... ... ..... .. .......... ............. ..... ............ .. 3.7-14 Cooper ii Amendment No. ~

~I

. ~

TABLE OF CONTENTS~

3.8 ELECTRICAL POWER SYSTEMS .. .......... ... .. .... ........ .... .. ......... ..... .... ......... 3.8-1 3.8 .1 AC Sources - Operating ..... ....... .... ..... .. ... ... .... ....... .. .... .......... .......... ..... .. 3.8-1 3.8 .2 AC Sources - Shutdown ... ... ... .. ...... ........ .. .. ......... ..... .... .... .. .......... .. .... .... 3.8-10 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air .... .... ...... ...... ... ..... .................. 3.8-13 3.8.4 DC Sources - Operating .... .... ...... ... ..... .. ......... ......... .... .... ..... ...... ............ 3.8-16 3.8 .5 DC Sources - Shutdown ...................... ... ..... ......... ..... ... .. ................. ....... 3.8-20 3.8.6 Battery Cell Parameters ............ ... .......... .............. ........ ....... ......... ...... .... 3.8-22 3.8 .7 Distribution Systems - Operating ... .. .. .. .. ... .... ........ ..... .... .. .......... ...... ... .... 3.8-26 3.8.8 Distribution Systems - Shutdown ......... .. ...... ... .................. ..... .... .. ........ .. 3.8-29 3.9 REFUELING OPERATIONS .. ............................ ... ..... .... ........ .. .... ... ..... ..... ... 3.9-1 3.9.1 Refueling Equipment Interlocks ..... ................... ... ................. ..... ... ...... .... 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock ... ...... .......... .... .... ..... .... ..... ......... 3.9-3 3.9.3 Control Rod Position ... .. ..... ..... ....... .... ...... ................... .. ............. ..... ... .... 3.9-4 3.9.4 Control Rod Position Indication ... ......... ... .... .......... ..... ..... .... ........... ...... .. 3.9-5 3.9.5 Control Rod OPERABILITY - Refueling .. ...... .... ... .................... ............ .. 3.9-7 3.9.6 Reactor Pressure Vessel (RPV) Water Level .... ... ............... .... ..... ........ .. 3.9-8 3.9.7 Residual Heat Removal (RHR) - High Water Level ... .............. ............... 3.9-9 3.9.8 Residual Heat Removal (RHR) - Low Water Level. .... ... .... .. ..... ...... .. ...... 3.9-12 3.10 SPECIAL OPERATIONS ............... ...... ..... .... ...... ... .. .................... ..... ~1 3.10.1 lnservice Leak and Hydrostatic Testing Operation ... ...... ..... ..... ........ .... 3.1 0-1 3.1 0.2 Reactor Mode Switch Interlock Testing ........... ............ ........ ................. 3.1 0-4 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ... ........... .... ......... .... .. .. 3.10-6 3.10.4 Single Control Rod Withdrawal - Cold Shutdown .... ..... ......... ... ;..:..:..:.; ** ***... 3~9 3.10.5 Single Control Rod Drive (CRD) ~,

Removal - Refueling ........ .. .. .. .......... .... .. .. : .... .. ..... ...... ................. 3.10-13 3.10.6 Multiple Control Rod Withdrawal - Refueling .......... ....... ........ ............. .. 3.1 0-16 3.10.7 Control Rod Testing - Operating ....... .............................. .. ...... ............. 3.1 0-18 3.10.8 SHUTDOWN MARGIN (SOM) Test - Refueling .... ... ............... ..... ........ 3.10-20 4.0 DESIGN FEATURES .. ............ .......... ...... ..... .... ............. .. ... .... ............... .... ... 4.0-1 4.1 Site Location ... ......... .. .. ... ...... ... .... ........ ........ .... ................. ..... .. ... ... ....... . 4.0-1 4.2 Reactor Core ..... .... .. ........................ ......... ... ........ ..... .. ....... ..... ....... ..... ... 4.0-1 4.3 Fuel Storage .. ....... ........... .... .. ........ .. ... ........ ... ... ..... .. ..... .... ...... ... ..... ....... 4.0-2 5.0 ADMINISTRATION CONTROLS ..... ... ... ....... ...... ......... .: .. ......... .. .. ........... .... 5.0-1 5.1 Responsibility .... ....... ... ....... ... ........ ................... .. ..... .......... .......... ........... 5.0-1 5.2 Organization .... ..... .... ... ........ ........ .... .... ......... .................... .... ................. 5.0-2 5.3 Unit Staff Qualifications ........ ... .. ....... .. .. .. .... ... ..... .... ... ..... ..... ...... .... ......... 5.0-4 5.4 Procedures ..... .... .... .... ..... .. ..... .... ................... .. .... .... ............. ... ...... ... .. ... 5.0-5 5.5 Programs and Manuals ... .... .................... .... .... ..... .. ........ ................ ........ 5.0-6 5.6 Reporting Requirements .... ........ ......... ... ... ......... .. .... ...... ......... .. .... ... .... . 5.0-20 5.7 High Radiation Area .. ..... .. .. .. .. .... ....... .... ........ ........... ............. .. ....... ........ 5.0-24 Cooper iii Amendment No. -258-

Definitions 1.1

1. 1 Definitions DOSE EQUIVALENT 1-131 1-133, 1-134, and 1-135 actually present. The DOSE (continued) EQUIVALENT 1-131 concentration is calculated as follows :

DOSE EQUIVALENT 1-131=(l-131)+0.0060 (l-132) + 0.17 (1-133) + 0.0010 (l-134) + 0.029 (1-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11 , "Limiting Values of Radionuclide Insert 1 Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion. and Ingestion," 1989.

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 drywall 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 drywall that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE 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 unit length of RATE (LHGR) fuel rod . It is the integral of the heat flux over the heat transfer area associated with the unit length.

(continued)

Cooper 1.1-3 Amendment No . ~

Insert 1 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 thro ugh 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 r eactor ooolant through the penetration flow paths;

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

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

c) The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory; d) No additional draining events oocur; and e) Realistic cross-sectional areas and drain rates are used.

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

ECCS Instrumentation 3.3.5.1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System {ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.1-1 .

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.1-1 for the channel.

8. As required by Required B. 1 -----------NOTES-- - - - -

Action A. 1 and referenced in 1. ORiy appliGable iR Table 3.3.5.1-1 . MOD!i:S 1, 2, anEt a.

-r.- Only applicable for Functions 1.a. 1.b, 2.a, 2.b, and 2.h.

Declare supported 1 hourfrom feature(s) inoperable when discovery of loss of its redundant feature initiation capability ECCS initiation capability for feature{s) in both is inoperable. divisions AND (continued)

Cooper 3.3-31 Amendment No. ~

ECCS Instrumentation r 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 8.2 --------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. i nit i at ion capability AND 8.3 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

C. As required by c.1 --------NOTES----- ---

Required Action A.I 1. Only ap13licable and referenced in i n MODES 1, 2, Table 3.3.5.1-1. a11d 3 .

-27 Only applicable for Functions 1.c, l.e, 2.c, 2.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. 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}

Cooper 3.3-32 Amendment No. -tTB--

ECCS Instrumentation 3.3.5.1 ACTIONS (continued}

CONDITION REQUIRED ACTION COMPLETION TIME

0. As required by 0.1 --------NOTE---------

Required Action A.l Only applicable if and referenced in HPCI pump suction is Table 3.3.5.1-1. not aligned to the suppression pool.

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

OR 0.2.2 Align the HPCI pump 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months suction to the suppression pool.

E. As required by E. l --------NOTE£--------

Required Action A.I }. enly appl; eable and referenced in ~A M99~S l , 2, Tab 1e 3. 3. 5 .1-1. and 3.

~ Only applicable for Functions l.d and 2.g.

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 capabi 1i ty for is inoperable. subsystems in both divisions ANO (continued)

Cooper 3.3-33 Amendment No. -+T&-

ECCS Instrumentation 3.3.5.1 Table 3 .3 .5 .1-1(page1 of6)

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

1. Core Spray System a Reactor Vessel 1,2.3.. 4(b) B SR 3.3.5 1.1 .'!. -113 inches Water Level - Low SR 3.3.5. 1.2( (d)

Low Low (level 1) ~ SR 3.3.5.1.4 c)

SR 3.3.5.1.5

b. Drywall Pressure- 1.2.3 4(b) B ~ 1.84 psig SR 3 3 5. 1 2( )(d}

High SR 3 3.5.1 4 c SR 3.3.5.1.5

c. Reactor Pressure- 1,2,3 4 c SR 3.3.5.1 .2 .'!. 291 ps ig and Low (Injection SR 3.3.51 .4 Pem1issive) SR 3.3 5 1 5 5 436 psig

~ SR 9.9.5.1 i! ::_ 29~ fil& i9 aR~

SR 9.9.5 1."4 SR a 3.5 1.S C 4J5 psi9 d Core Spray Pump 1,2.l. 1 per pump E SR 3.35 .1 2(c)(d) .'!. 1370 gpm Discharge Flow - SR 3.3.5.1 4 Low ~ SR 3.3.5.1 5 (Bypass)

e. Core Sp ray Pump 1.2,3.. 1 per pump c SR 3.3.5.1.2 > 9 seconds Start-Time Delay SR 3.3.5.1 4 and Relay ~ SR 3 3.5. 1 5 ~ 11 seconds

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Vessel 1.2.:t. 4 B SR 3.3.5. 1 1 _! -113 inches Water Level - Low SR 3.3.5. 1.2(c)(d)

Low Low (Level 1) ~ SR 3.3.5. 1.4 SR 3 3.5.1.5 (continued)

(a)

~ [De l eted] I (b) Also required to initiate the associated diesel generator (DG).

(c} If the as-found channel setpoint is outside its predefined as-found tolerance . then the channel shall be evaluated to verify that ij is function i09 as required before returning the channel to service.

(d) The instrument channel setpoln! shall be reset to a value that is within the as-left tolerance arovnd the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise , the channel shall be declared inoperable. Setpolnts more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint} to confirm channel performance. The Limiting Trip Setpoin\ and the methodologies used lo determine the as -found and the as-left tolerances are specified in the Technical Requirements Manua l.

Cooper 3.3-37 Amendment No. ~

ECCS Instrumentation 3.3.5.1 Table 3.3.5 1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTIONA.1 REQUIREMENTS VALUE 2 LPCI Sys1em (continued) 1,2,3 4 B SR 3.3.5.1.2 :; 1.84 ps ig b Drywell Pressure - SR 3.3.5. 1.4(c)(d)

High SR 3.3.5. 1.5

c. Reactor Pressure

1,2,3 4 c SR 3.3.51 .2 ~ 291 psig and Low (Injection SR 3.3.5 .1.4 Permissive) SR 3.3.5.1 5 ~ 436 psig SR 3 3.6.1.2 :! <1111 psig aPd SR a .a.6 u 6R a 3.5.U ! 4li p&i!J

d. Reactor Pressure - 4 c SR 3.3.5.1.2 ~ 199 psig and Low (Recirculation SR 3.3.5 .1.4 ~ 246 psig Discharge Valve SR 3 .3 5 1 5 Permissive)

e. Reactor Vessel 1,2,3 2 B SR 3.3.5. 1. 1 > .1g3 , 1g Shroud Level

SR 3.3.5 . I .2 loch es Leveto SR 3.3.5 .1.4 SR 3.3.5.1.5 I. Low Pressure 1,2,3.. 1 per pump c SR 3.3.5.1.2 Coolant Injection SR 3.3.5.1.4 Pump Start -Time SR 3.3.5 .1.5 Delay Relay Pumps B.C > 4.5 seconds and

5.5 seconds PumpsA.D ~ 0.5 second (conOnued)

(a)

(c) If the as-found channel setpoint Is outs ide its p redefined as-found tolerance, then the channel shaU be evaluated to verify that it is functioning as required before returning the channel to service.

(d) The instrument channel setpoint shall be reset to a value that is wrthin the as-left tolerance around the Lim iting Trip Setpoiflt (LTSP) at the completion of the surveillance. olheJWise. the channel sha1I be declared Inoperable. Setpolnts more conservative than the l TSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the SuNeittance procedures (Nominal Tnp Setpomt) lo confirm cllannel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified 1n the Technical Requirements Manual.

(e) With associated recirculation pump discharge valve open.

Cooper 3.3-38 Amendment No. ~

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1*1(page3 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQU IRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCT ION CONDITIONS FUNCTION ACTIONA.1 REQUIREMENTS VALUE

2. LPCI System (continued)

g. Low Pressure Coolant 1,2,3,. 1 per subsystem E SR 3.3.5.1.2 :t: 2107gpm Injection Pump Discharge SR 3 .3.5.1.41<X*I Flow - Low (Bypass) ~ SR 3.3.5.1 .5

h. Containment Pressure -

High 1,2,3 4 B SR 3.3.5.1.2 SR 3.3.5.1 .4 SR 3.3.5.1.5

t: 2 psig I

3. High Pressure Coolant Injection (HPCI) System

a. Reactor Vessel Water Level 1, 4 B SR 3.3.5.1.1 :t: -42 inches

- Low Low (Level 2) SR 3.3.5.1.2 2m,3cll SR 3.3.5.1.4C*X*I SR 3.3.5.1.5

b. Drywetl Pressure - High 1. 4 B SR 3.3.5. 1.2 s 1.84 pslg SR 3.3.5.1.41*x*1 t'>, j '> SR 3.3.5.1.5

c. Reactor Vessel Water Level 1, 2 c SR 3.3.5.1 .1 s 54inches

- High (level 8) SR 3.3.5.1.2 SR 3.3.5.1.4 t '>, 3ll) SR 3.3.5. 1.5

d. Emergency Condensale 1, 2 D SR 3.3.5.1.2 :t: 23 inches Storage Tank (ECST) Level - SR 3.3.5.1.3 Low 2<ll, 3<1) SR 3 .3.5.1.5

e. Suppression Pool Water 1, 2 D SR 3.3.5.1.2 s 4 inches Level - High SR 3.3.5.1.4 ill, 3<ll SR 3.3.5.1.5 continued (a ) WbeA Ille associe1ec1 ECCS s uhs)<StMl(s) ere *equi*ed to be OPEliABLE per LCO 3 5 2, ECCS Sbuld- ~[Deleted] l (c) If the as-found channel setpoint is outside Its predefined as-found tolerance. then the channel shall be evaluated l<> verify that it Is functioning as required before returning the channel to service.

(d) The instrument channel setpoint shall be reset to a value that is within the es-left tolerance around ttie Limiting Trip Setpoint (LTSP) at tl1e completion of the surveillance; othefwise, the channel shall be declared inoperable. Selpolnts more conservative than the L TSP are acceptable provided lhat the as-found and as-left tolerances apply to the actual setpolnt Implemented in lhe Surveillance procedures (Nominal Trip Setpoinl) to confirm channel performance . The Limiting Trip SelpOlnt and lhe methodologies used to determine the as-found and the as-left toteranoes are specified in the Technical Requirements Manual.

(f) W ith reactor steam dome pressure >150 psfg .

Cooper 3.3-39 Amendment No. -rs-a-

RPV Water Inventory Control Instrumentation 3.3.5.3 3.3 INSTRUMENTATION 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.3 The RPV Water Inventory Control Instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.3-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

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

B. As required by Required B.1 Declare supported Immediately Action A.1 and referenced in penetration flow path( s)

Table 3.3.5.3-1 . incapable of automatic isolation .

AND B.2 Calculate DRAIN TIME. Immediately C. As required by Required C.1 Place channel in trip . 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action A.1 and referenced in Table 3.3.5.3-1 .

(continued) I Cooper 3.3-47 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Action A.1 and referenced in OPERABLE status.

Table 3.3.5.3-1 .

E. Required Action and E.1 Declare associated low Immediately associated Completion Time pressure ECCS of Condition C or D not met. injection/spray subsystem inoperable.

SURVEILLANCE REQUIREMENTS

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

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

SURVEILLANCE FREQUENCY SR 3.3.5.3.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-48 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5.3 Table 3.3.5.3-1 (page 1 of 1)

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

1. Core Spray System

a. Reactor Pressure - Low 4,5 4 c SR 3.3.5.3.2 s 436 psig (Injection Permissive)

b. Core Spray Pump Discharge 4,5 1 per pump<*> D SR 3.3.5.3.2 2 1370 gpm Flow - Low (Bypass)

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Pressure - Low 4,5 4 c SR 3.3.5.3.2 s 436 psig (Injection Permissive)

b. Low Pressure Coolant 4,5 1 per D SR 3.3.5.3.2 2 2107 gpm Injection Pump Discharge subsystem<*>

Flow - Low (Bypass)

3. RHR System Isolation (b)

a. Reactor Vessel Water 2 in one trip B SR 3.3.5.3.1 2 3 inches Level - Low, Level 3 system SR 3.3.5.3.2

4. Reactor Water Cleanup (RWCU)

System Isolation

( b)

a. Reactor Vessel Water Level 2 in one trip B SR 3.3.5.3.1 2 -42 inches

- Low Low, Level 2 system SR 3.3.5.3.2 (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "Reactor Pressure Vessel Water Inventory Control."

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

Cooper 3.3-49 Amendment No.

Primary Containment Isolatioo Instrumentation

- 3.3 . 6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 Be in MOOE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action C.l and referenced in ANO Table 3.3.6.1 -1.

G.2 Be in MOOE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Required Action and associated Completion Time for Condition F not met.

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.l standby liquid and referenced in control (SLC}

Table 3.3.6.1-1. subsystem(s) 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 1.1 Initiate action to Immediately Required Action C.l restore channel to and referenced in OPERABLE status.

Table 3.3 .6.1-1.

~

1. 2 1n;t;ate aet;en te lmmed~ate~:Y

~se~ate the Res~aija~

Heat RemeYa~ (RMR)

Sh1:1teown Cooling System .

~

Cooper 3.3-~ Amendment No. ~

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

6. RHR Shu tdown Cooling System Isolation

a. Reactor 1,2,3 F SR 3.3.6.1 .2 ~ 72 psig Pressure - High SR 3.3.6.1.4 SR 3.3.6.1 .6

b. Reactor Vessel Water SR 3.3.6.1.1 ~ 3 inches Level - Low (Level 3) SR 3.3.6 .1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 (e) 0 Rlt one tril' :i:y:ite1'1 1 is reeit:ii1 ed iii MODES 4 Bl'ld 5 nheli RI IR Sht:ildoom Coolil'lg Splel'l1 ililegrilt 111eil'lleil'led .

~

Amendment No .

~enti11i2

~ Cooper 3.3-J

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1{page1of1 )

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,J,. 2 SR 3.3.6.2.1 ?:'.. - 42 inches Level - Low Low (Level 2) ~ SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4

2. Drywell Pressure - High 1,2,3 2 SR 3.3.6.2.2 ~ 1.84pslg SR 3.3.6.2.3 SR 3.3.6.2.4

3. Reactor Building Ventilation 1,2,3, 2 SR 3.3.6.2.1 ~ 49 mR/hr Exhaust Plenum -W,(b) SR 3.3.6.2.2 Radiation - High SR 3.3.6.2.3 SR 3.3.6.2.4 (a) Dtiring eperelten!! wwlttl 11 polentiel fe1 d1eining lne reeeter wessel.

j[Deleted] I

~..:::.-------i!..-.:.....:._.:_.:...!....J (b) During movement of recently irradiated fuel assemblies In secondary containment r-@J Cooper 3.3-# Amendment No. -244

CREF System Instrumentation 3.3.7.1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Filter (CREF) System Instrumentation LCO 3.3.7.1 The CREF System instrumentation for each Function in Table 3.3.7.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.7. 1-1 ~

ACTIONS

N 0TE-------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for inoperable. Functions 1 and 2 AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Function 3 B. One or more Functions B.1 Restore CREF System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with CREF System initiation capability.

initiation capability not maintained.

C. Required Action and C.1 Initiate CREF System. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion Time not met. OR C.2 Declare CREF System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable.

Amend me~ 3.3-6~

Cooper I Amendment No.

CREF System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Filter System Instrumentation AP PLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURV EILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQU IREMENTS VALUE

1. Reactor Vessel Water 1,2. 3.- 2 SR 3.3.7.1.1 > - 42 inches Level - Low Low (Level 2) -w- SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4

2. Drywell Pressure - High 1.2.3 2 SR 3.3.7.1.2 ~ 1.84 psig SR 3.3.7.1.3 SR 3.3.7.1.4

3. Reactor Building Ventilation 1,2,3,. 2 SR 3.3.7.1.1 ~ 49 mR/hr Exhaust Plenum ~ (b) SR 3.3.7.1.2 Radiation . High SR 3.3.7.1.3 SR 3.3.7.1.4 (a) Ouri119 opeialio113 oitl1apote11tial101 d1ai11i11g Ilic 1eactrn *essel. -<i<~---il~[D~e~le~t~e~dlJ]I (b) Du ring movement o f lalely irrad ialed fuel assemblies in the secondary containment.

~ Coope< r@J IAmendment No . ~

/\~Fll222 3.34 - ~/06

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating , RPV WATER INVENTORY CONTROL, LCO 3.5 .1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/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 low pressure 7 days ECCS injection/spray ECCS injection/spray subsystem inoperable. subsystem(s) to operable status.

OR -

One LPCI pump in both LPCI subsystems inoperable.

B. Required Action and B.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 not AND met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

Amendment No.

3.5-1 09/18/09

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 .

RPV Wate r One Inventory 3.5 EMERGENCY CO E COOLING SYSTEMS (ECCS) ND REACTOR CORE ISOLATION Control COOLING (RCI ) SYSTEM

, RPV WATER INVENTORY CONTROL,

3. 5.2 Reactor Pressure Vessel (RPV) Water Inventory Control LCO 3.5 . 2 low pressure ECCS injection/spray subsyste shall be 0 PE RAB LE . ------------------------------------------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.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Required A. One 1 eqtJ i 1 ed ECCS A. l Re store required ECCS 4 hours injection/spray injection/spray subsystem inoperable. subsystem to OPERABLE status.

B. Required Action and B.1 Initiate action to Immediately associated Completion 5W5P9RQ gperati9R& establish a method of Time of Condition A with a J:leteRtial fer water injection capable not me t. draiRiAg tke reactor of operating without

'lessel (OPDRVs) . offsite electrical power.

Verify secondary containment boundary is capable of 1--- - - - - -- - - - - - - - - - - - -

being established in less than the DRAIN TIME.

, ,,--14 hours I C. Twe re~uired ECCS !Aitiate actioA to Immedi at ~ *'------'

iAjeetieA/sr:iray s~sJ:ieRd OPORl/s .

subsysteffls iAer:ierable .

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

Verify each secondary containment penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months flow path is capable of Verify one standby gas treatment subsystem is ( c ontinued) being isolated in less than the DRAIN TIME. capable of being placed in operation in less than the DRAIN TIME .

Coop er 3. 5-7 Amendment No. +f.&

IRPV Water Inventory Control

~ ECCS Shutdown

3. 5. 2 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME

0. Requited Action C. z D. 1 IAitiate aetioA to Immediately aAe assoeiatee l"estol"e secoAeary Cem~letioA Time Aet ceAtaiAf!leAt ts

-f!le-h- OPERABLE status .

DRAIN TIME < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . AND D.2 IAitiate actisA ts Immediately

N 0 TE-------------- l"estore eAe staAeby Re qui red ECCS injection/ ~as tl"eatfReAt spray subsystem or subsystem te OPERABLE add itional method of water stat1:.1s .

injection shall be capable of operating without offsite electrical power.

IAitiate actioA ta Immediately restsre isslatisA Initiate action to establish an ea~ability iA eacA additional method of water req~ired secoAdary injection with water sources COAtaiAl!leAt capable of maintaining RPV 13eAetrat i OR fl OH 13atl'I water level > T AF for '.'.'.. 36 Rot isolated .

hours.

Initiate action to establish secondary containment boundary. Initiate action to verify one standby Immediately gas treatment subsystem is capable of being placed in Initiate action to isolate each operation.

secondary containment penetration flow path or verify it can be manually isolated from the control room .

E. Required Action and associated E.1 Initiate action to restore Immediately Completion Time of Condition C or D not met. DRAIN TIME to '.'.'.. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

Cooper 3.5 - 8 Ame ndment No. -H RPV Water Inventory Control SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.5.2.+ Verify, for ea6R required ECCS injection/spray In accordance with v subsystem, the suppression pool water level is~ 12 ft 7 inches.

the Surveillance Frequency Control Program SR 3 . 5. 2.~ Verify, for required ECCS injection/spray In accordance with

[J-fi subsystem, the piping is filled with water from the pump discharge valve to the injection valve.

the Surveillance Frequency Control Program SR 3.5.2.S NOTE

~

One LPCI subsystet11111ay be co11slde1ed Of'Eftbi8LE euFiAg eligAmeAt em1 epeffitiOA for eoeey heat i:emeval if sapable ef l:JeiR§ mamially realigAeEI aAEI Aet etherwise iReperable.

~ ~--

Verify 8asR required ECCS injection/spray subs~ In accordance with EJ----7 .

manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise the Surveillance Frequency Control secured in position, is in the correct position. Program SR 3.5.2.1 Verify DRAIN TIME~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(continued)

In accordance with the Surveillance Frequency Control Program Cooper 3.5-9 Amendment No. '

jRPV Water Inventory Control p ECCS - ShutdoWl'I 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.5.2.4 NO. CORRESPONDIN6 OF TO A REACTOR i SYSTEM FLOW RATE PUMPS PRESSURE OF 1

SR 3.5.2.5 NOTE Vessel i"jeetiefb'spray may be excluded.

Verify each required ECCS injectio1o'spra7 subsyste111 In accordance with actuates eA BA actual er simulated automatic iAitietioA the Surveillance signal. Frequency Control Program Operate the required ECCS injection/spray subsystem through the recirculation line for ~

10 minutes .

.L l \ _ In accordan ce with Verify each valve credited for automatically isolating a the Surveill ance penetration flow path actuates to the isolation position on an Frequency Control actual or simulated isolation signal. Program L.

l \ _ In accordance with

----------------------------N0 TE---------------------------------- the Surveillan ce Vessel injection/spray may be excluded . Frequency Co ntrol

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

Cooper 3.5-10 Amendment No.-259-

RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)\ : REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM

, RPV Water Inventory Control , I 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.

COMPLETION CONDITION REQUIRED ACTION TIME A. RCIC System inoperable. A.1 Verify by administrative means 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System to OPERABLE status . 14 days B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Reduce reactor steam dome 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months pressure to ~ 150 psig .

IAmendment No. ~

~

Amendment 233 3. 5-~

r-@

' ~ 8/00

PC I Vs 3.6.1.3 ACTIONS (c onti nu ed)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more D.1 Restore leakage rate 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration fl owpaths to within l i mi t.

with one or more MS I Vs

~

not within leakage rate limit.

E. Required Action and E. l 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 l , 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

~. ReEl1:1~r::ea AEt~eA aAS F. 1 f11itiate ac:tio11 to Immeeliately asseEiateel bSRlfll eti SA St:!SfleAel Sf1eFatiSA5 fime ef EerH:IH:i el"I A, ~.,~ Ui a 13eteRt~al f:e r:

8 , E, 01 B 11ot niet d1* 8iM;l"I~ the l"e8etel" f:sr:: Pb PJ ~ s) r>eEll:I i r::eEI vessel .

to be 8PERABl:E dt11 i ng MODE 4 or 5 . ~

F. 2 IAHiate aEtisA ts Imme el i atel y l"estel"e 't'ahce~s~ te 8PERA81::E :o;t8t:t1s .

Co oper 3. 6 - 11 Am endme nt No . +f8-

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 recent~ iated fuel assemblies in the secondary containmen~

D1:1FiAg epeFetieAs with e peteAtiel feF EfreiAiAO the reeeteF 'lessel (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, 2, or containment to

3. OPERABLE status.

B. Required Action and B.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 not met. ANO B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Secondary containment C.1 - - - -NOTE-------

inoperable during movement LCO 3.0.3 is not of recently Irradiated fuel applicable.

assemblies in the secondary - -

containment or d~ring OPDR'Js. Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

-AN&-

.... ~\

Cooper 3.6-34 Amendment No. ..z.5a-I

Secondary Containment 3.6.4.1 7\CTIOr~s C. (continued) C.2 Initiate action to suspend lmmedietely OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 .1 Verify secondary containment vacuum is 2: 0.25 inch In accordance with of vacuum water gauge. the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify all secondary containment equipment hatches In accordance with are closed and sealed. the Surveillance Frequency Control Program

SR 3.6.4.1 .3 Verify one secondary containment access door in In accordance with each access opening is closed. the Surveillance Frequency Control Program SR 3.6.4.1 .4 Verify each SGT subsystem can maintain c:: 0.25 inch In accordance with of vacuum water gauge in the secondary containment the Surveillance for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate :s; 1780 cfm. Frequency Control Program Cooper 3.6-35 Amendment No.--256-

SCI Vs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Va lves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of r~cep!,ly irradiated fuel assemblies in the secondary contain men~

OtJring operations with a potential for draining the reactor vessel (OPOR\!s).

ACTIONS

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

1. Penetration flow paths may be unisolated intermittendy 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 SC IVs.

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

(continued)

Cooper 3.6-36 Amendment No. ~

SC IVs 3.6.4.2 ACTIONS (continued )

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 -------NOTE-------

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

during movement of recently irradiated fuel assemblies in the secondary containment Suspend movement of Immediately or during OPDRV-S. recently irradiated fuel assemblies in the secondary containment.

-AN9-D.2 Immediately OPDRVs.

Cooper 3.6-38 Amendment No. ~

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 re~y irradiated fuel assemblies in the secondary containmen.t.k--lj gy1=iAg epeFatieR6 wi&h a peleAtial fer draiAiAg U:ie reaeter 'Je66el (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7days inoperable. to OPERABLE status.

B. Required Action and 8.1 Be in MODE3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met in AND MODE 1, 2, or 3.

B.2 Be In MODE4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

c. Required Action and --- -------NOTE- - - - - - - -

associated Completion Time LCO 3.0.3 is not applicable.

of Condition A not met --------

during movement of recently irradiated fuel assemblies in C.1 Place OPERABLE SGT Immediately the secondary containment subsystem in operation.

er dtJring OPDRVs.

OR (continued)

Cooper 3.6-40 Amendment No. ~I

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

~

C.2.2 Initiate action to suspend Immediately 9P9R¥s.

0. Two SGT subsystems 0 .1 Enter LCO 3.0.3 Immediately inoperable in MOOE 1, 2, or 3.

E. Two SGT subsystems E. 1 -------NOTE------

inoperable during movement LCO 3.0.3 is not of recently irradiated fuel applicable.

assemblies in the secondary -- ---

containment gr duriRg OPDRVs. Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

-=-'-""'

I

, ..'-* ~'

Cooper 3.6-41 Amendment No.~

SGT System 3.6.4.3 ACTIONS eONOl'flO~ REel:HREB ACTl9N 69MPl::Efl0N TIME E. (eeAtiAued) E.2 Initiate aetion te susf'end- lmfflediately ePBRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.-6.4.3.1 Operate each SGT subsystem for<!:: 10 continuous In accordance with hours With heaters operating. the Surveillance Frequency Control Program SR 3.6.4.3.2 Perform required SGT filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or In accordance with simulated initiation signal. . the Surveillance Frequency Control Program SR 3.6.4.3.4 . Verify the SGT units cro~ tie damper is in the correct In accordance with position, and each SGT room air supply check valve the Surveillance and SGT dilution air shutoff valve can be opened . Frequ~ncy Control Program Cooper 3.6-42 Amendment No. ~

CREF System 3.7.4 3.7 PLANT SYSTEMS

3. 7.4 Control Room Emergency Filter ( CREF) System LCO 3. 7.4 The CREF System shall be OPERABLE.

N 0 TE----------------------------------------------------

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

APPLI CABILITY : MODES 1, 2, and 3, During movement of lately irradiated fuel assemblies in the secondary containmen ~

DuFiR§ epe FstieRs ~ peteRtisl fur drsiRiR§ the reseter vessel (OPDRVs).

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. CREF System inoperable A.1 Restore CREF System to 7 days for reasons other than OPERABLE status .

Condition B.

B. CREF System inoperable 8 .1 Initiate action to implement Immediately due to inoperable CRE mitigating actions.

boundary in MODE 1. 2.

or 3 . AND B.2 Verify mitigating actions ensure 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months CRE occupant exposures to radiological and chemical hazards will not exceed limits ,

and CRE occupants are protected from smoke hazards .

AND 90 days B.3 Restore CRE boundary to OPERABLE status.

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 or B AND 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 (continued)

Cooper 3.7-8 Amendment No. -i8-G-

CREF System 3.7.4 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and --------------NOTE---------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of lately irradiated fuel D.1 Suspend movement of lately Immediately assemblies in the irradiated fuel assemblies in secondary containment the secondary containment.

or during OPDR'n.

D.2 Initiate aetiefl te lmmediHtely CREF System suspeAEt OPDRVs.

inoperable due to an inoperable CRE boundary during movement of lately irradiated fuel assemblies in the secondary containment er 61::1ring OPDR\ls.

Cooper 3.7-9 Amendment No. -244-

AC Sources - Shutdown 3.8.2 ACTIONS

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

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite ------------NOTE-------------

circuit inoperable. Enter applicable Condition and Required Actions of LCO 3.8.8, when any required division is . de-energized as a result of Condition A.

A.I Declare affected Immediately required feature(s),

with no offsite power available, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

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

A.2.3 IAitiate actieA te Immediately suspeRd operations witb a poteRtial for draiRiRg the reactor vessel (OPDRVs) .

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

(continued)

Cooper 3 .8-11 Amendment No. -t1&

AC Sources - Shutdown 3.8.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG 8 .1 Suspend CORE Immediately inoperable. AL TERATIONS.

8.2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.

f B.3 Initiate action to suspend OPDRVs.

lmn 1ediately

~ Initiate action to restore required DG to OPERABLE status. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 -~------------~--NC>TES-~~-~---~~~

1. The following SRs are not required to be performed: SR 3.8.1.3, and SR 3.8.1.9 through SR 3.8.1.11 .

2. SR 3.8.1.11 is considered to be met without the ECCS initiation signals OPERABLE when the ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1 .

For AC sources required to be OPERABLE the In accordance SRs of Specification 3.8.1, except with applicable SR 3 .8.1.8, are applicable. SRs Cooper 3.8-12 Amendment No. :218

DC Sources - Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS

3. 8. 5 DC Sources - Shutdown LCO 3.8.5 DC electrical power subsystems shall be OPERABLE to support the DC electrical power distribution ~ubsystem{s) required by LCO 3.8.8, "Distribution Systems -Shutdown."

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

ACTIONS

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

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC electrical power required feature(s) subsystems inoperable. inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary

containment.

Aun

-=

(continued)

Cooper 3.8-20 Amendment No. 178

DC Sources - Shutdown 3.8.5

'-..../'. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. {continued) A. 2.3 Jn;t;ate action to Immediately s~speAe eperatieAs with a peteAtial fer draiRiRg the reactor

1essel .

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: SR 3.8.4.7 and SR 3.8.4.8.

For DC sources required to be OPERABLE, the In accordance following SRs are applicable: with applicable SRs SR 3.8.4.1 SR 3.8.4.4 SR 3.8.4.7 SR 3.8.4.2 SR 3.8.4.5 SR 3.8.4.8 SR 3.8.4.3 SR 3.8.4.6 Cooper 3.8-21 Amendment No. -l Distribution Syste s -Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems - Shutdown LCO 3.8.8 The necessary portions of the AC and DC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

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

ACTIONS

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

LCO 3.0.3 is no t applicable CONDITION REQUIRED ACT ION COMPLETION TIME A. One or more required A. I Declare associated Immediately AC or DC electrical supported required power distribution feature(s) subsystems inoperable . inoperabl e .

OR A.2 . 1 Suspend CORE Immediately ALTERATIONS .

A. 2. 2 Su spend movement of Immedi"ately i rradia ted fuel assembl ies i n t he secondary co nt ai nment.

Afill

-A . 2. 3 lRitiat~ aetieR te Immediately s~speAcl operatioAs

~11 Hi a poteAti al for draiAiAg the reactor vessel .

(c onti nu ed)

Cooper 3.8 - 29 Amendment No . 178

Distribution Systems - Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to restore Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.

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 voltage to In accordance with required AC and DC electrical power distribution the Surveillance subsystems. Frequency Control Program Cooper 3.8-30 Amendment No . .zs.Er

NLS2017048 Page 1of64 Attachment 3 Revised Technical Specifications Pages Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 TOC page i 3.3-54* 3.5-9 TOC page ii 3.3-55* 3.5-10 TOC page iii 3.3-56 3.5-11 1.1-3 3.3-57* 3.5-12 1.1-4 3.3-58* 3.5-13*

1.1-5* 3.3-59* 3.5-14*

1.1-6* 3.3-60 3.6-11 1.1-7* 3.3-61

3.6-34 3.3-31 3.3-62* 3.6-35 3.3-32 3.3-63* 3.6-36 3.3-33 3.3-64 3.6-38 3.3-37 3.3-65* 3.6-40 3.3-38 3.3-66 3.6-41 3.3-39 3.3-67* 3.7-8 3.3-47 3.3-68* 3.7-9 3.3-48 3.3-69* 3.8-11 3.3-49 3.3-70* 3.8-12 3.3-50* 3.3-71

3.8-20 3.3-51

3.5-1 3.8-21 3.3-52 3.5-7 3.8-29 3.3-53* 3.5-8 3.8-30

included due to repagination only Note: TS page 3.6-42 was deleted due to information on page moving to page 3.6-41.

TABLE OF CONTENTS 1.0 USE AND APPLICATION ........ .. ................................................................... ..... 1.1-1 1.1 Definitions ... ........ ....... .... ......... ................ ..... .. ....... ....... ........ ... ...... ......... .... .. 1.1-1 1.2 Logical Connectors ... ............. .... .... .. ... ..... ... ... ..... ..... ... ..... .. ...... ....... ..... ..... ... 1.2-1 1.3 Completion Times .. ....... ..... ..... .......... ... .............. .... .............. .. ......... .. .. ......... 1.3-1 1.4 Frequency ..... ......... ... ... ..................................... .......................................... 1.4-1 2.0 SAFETY LIMITS (SLs) ............................. .... ...... ....... ...... .................. ............. .. . 2.0-1 2.1 SLs .... ... .. ......... .. .... ......... .... ... ...... ........................ .......... ............................. 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 3.1.1 SHUTDOWN MARGIN (SOM) .......... .......... .. .. .... ............ ....... .... ............ 3.1-1 3.1.2 Reactivity Anomalies ........... ...... ......... ....... ... .................. .. ..... ................. 3.1-5 3.1.3 Control Rod OPERABILITY ............................................ ......... .... ... ....... 3.1-7 3.1 .4 Control Rod Scram Times .... ............ ...... .. .................. .. .. .... .. ............ ...... 3.1-12 3.1.5 Control Rod Scram Accumulators .. .... ..... ...... ................ .... ........... .. ...... .. 3.1-15 3.1.6 Rod Pattern Control ............................................................................... 3.1-18 3.1.7 Standby Liquid Control (SLC) System .. ..... .. ...... ................ .... ..... ............ 3.1-20 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves ... ...... ...... .. ....... 3.1-25 3.2 POWER DISTRIBUTION LIMITS .. ........ ........ .... .... ....................................... 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) .. 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ...................................... 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ... ... ............ .............. ....... 3.2-4 3.3 INSTRUMENTATION ...... ..... ............... ... ........... ..... ..... ...... ........... ..... .......... 3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation .... .. .. .. .. ..... ........ ...... .. 3.3-1 3.3.1 .2 Source Range Monitor (SRM) Instrumentation .. .. ............ ........ ............ ... 3.3-9 3.3.2.1 Control Rod Block Instrumentation .. .... ... .... .. .. ...... ....... .. ...... ... .. ....... ... .... 3.3-14 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation .... . 3.3-20 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ...... .................. .. ......... 3.3-22 3.3.3.2 Alternate Shutdown System ........................................ ... ... ................ .. ... 3.3-26 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) lnstrumentation ... ..... .. .... ... .. ..... .... ... .... ... ..... .... ..... ...... . 3.3-28 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ....... ......... .... 3.3-31 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation ....... ... .. 3.3-43 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ......... ... ............ ... ......................... ..... ... ......... ... .. ... ... 3.3-47 3.3.6.1 Primary Containment Isolation Instrumentation ............. ..... .. ....... .... .. ..... 3.3-50 3.3.6.2 Secondary Containment Isolation Instrumentation .... ........... ... .......... ..... 3.3-57 3.3.6.3 Low-Low Set (LLS) Instrumentation .. ... ...... ...... ..................... ... .. .... .. ...... 3.3-61 3.3.7.1 Control Room Emergency Filter (CREF) System Instrumentation .. ........ 3.3-64 3.3.8.1 Loss of Power (LOP) Instrumentation ........ ... ....... ... ........................ ...... . 3.3-67 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .. .... ........... 3.3-70 (continued)

Cooper Amendment No.

TABLE OF CONTENTS (continued) 3.4 REACTOR COOLANT SYSTEM (RCS) .. .... ..... ... ....... .......... .. ..... ... .... .......... 3.4-1 3.4.1 Recirculation Loops Operating ... .. ... ...... .. ... ............... ...... .. ..... .... ..... .... ... 3.4-1 3.4.2 Jet Pumps ...... ... ... ....... .... ... ... .. .... ... ....... ... ....... ... .. ... .... ...... ............ ......... 3.4-4 3.4.3 Safety/Relief Valves (SRVs) and Safety Valves (SVs) ... ...... .. .. ... ..... ...... 3.4-6 3.4.4 RCS Operational LEAKAGE .. ... .. .... .. ....... .. ..... ..... .. ...... .. ... ........ ... .... ..... . 3.4-8 3.4.5 RCS Leakage Detection Instrumentation .... ... ..... .. .... ..... .. ... .... ... ............ 3.4-10 3.4.6 RCS Specific Activity .. .. ...... ..... ..... ..... .. ........ .. ...... .. ....... .. ...... ....... ....... ... 3.4-12 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ... ... ..... ....... ... .. ...... .... ... ..... .. ... ........ ... ...... ... .. 3.4-14 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown ..... .... ... .... ... ... .... ..... .... ....... .. .... ..... ..... ...... .. 3.4-17 3.4.9 RCS Pressure and Temperature (Pff) Limits .. .... .... ..... ... ..... .. .... .. .. ... .. ... 3.4-19 3.4.10 Reactor Steam Dome Pressure .......... ...... .. ....... .. ........ .... ...... ... .. ..... ...... 3.4-23 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM .. .. ... ..... .. ........ .... .. ... ... .... ... ... ...... .... ... .. ... ... ... .... 3.5-1 3.5.1 ECCS - Operating ...... .. .... ........ .... .. .. ............ ...... .. ... ... ... ....... ... .. ..... .. ... .. . 3.5-1 3.5.2 RPV Water Inventory Control .... ... ... .... ......... ..... ....... ... ... ... ... .... ... ....... ... . 3.5-7 3.5.3 RCIC System ......... .. .. .. ... ............ .. .......... ... .... ... ...... .......... .. ... ... ... ....... ... 3.5-12 3.6 CONTAINMENT SYSTEMS .... ...... ..... ... .. .... .. ... ... .... .... ........ ... ..... ..... ... ........ 3.6-1 3.6.1.1 Primary Containment .. ..... ... .. ... .... .. ... ........... .... ...... .... .. .. .......... ... .... ... .... 3.6-1 3.6.1.2 Primary Containment Air Lock .. ... ........ .. ... .......... ....... .... ... ... ... .. .. .......... .. 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .. ....... ..... .. ........... ....... .... 3.6-8 3.6 .1.4 Drywell Pressure .... ...... .. ... ...... ... ...... .... ... ...... .... ....... ............ ... .............. . 3.6-16 3.6.1.5 Drywell Air Temperature ............... ... .... .... ............... ..... .......... .... .... .... ... . 3.6-17 3.6.1.6 Low-Low Set (LLS) Valves ... ... ..... .. ... .... .. .. .... ... ... ... ..... ... ........... ....... ...... 3.6-18 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers ... .. ... ..... .. 3.6-20 3.6.1.8 Suppression-Chamber-to-Drywell Vacuum Breakers .. ....... ..... ... ....... .... . 3.6-23 3.6.1.9 Residual Heat Removal (RHR) Containment Spray .... .... .. .... .. .. ... .......... 3.6-25 3.6.2.1 Suppression Pool Average Temperature ... .. ..... ........ .. .. .. .. ... ........ .... .... .. 3.6-27 3.6.2.2 Suppression Pool Water Level .. ... ..... ............ .... ... .... .... ... .. ..... .. ..... ... .. ... . 3.6-30 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ..... ........ ..... .. . 3.6-31 3.6.3.1 Primary Containment Oxygen Concentration ............. ... .. .... .. ... ..... ......... 3.6-33 3.6.4.1 Secondary Containment. .... ..... .. ... .... ... ......... ... .......... .. .. .... ... ... .. ....... .... .. 3.6-34 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) .. ... ... .. ..... .. .. .. .. .. ........ . 3.6-36 3.6.4.3 Standby Gas Treatment (SGT) System ....... .. .............. ... .. ...... .... ....... .... . 3.6-40 3.7 PLANT SYSTEMS .... ... .... .... ...... ... ..... .. ... .. .. .... ... ... .. ...... .. ... ....... ... .. ... ... ... ... .. 3.7-1 3.7.1 Residual Heat Removal Service Water Booster (RHRSWB) System ... .. 3.7-1 3.7.2 Service Water (SW) System and Ultimate Heat Sink (UHS) ........ .... ..... .. 3.7-3 3.7.3 Reactor Equipment Cooling (REC) System .................. .... ... ... .. ...... ... ..... 3. 7-6 3.7.4 Control Room Emergency Filter (CREF) System .. .... .. .. ...... .... .... ....... .... 3.7-8 3.7.5 Air Ejector Offgas .. ........... .... ...... ... .. .... ..... .. .. ......... ......... ....... ...... .... ....... 3.7-11 3.7.6 Spent Fuel Storage Pool Water Level ..... ..... ......... ... ... .. ....... ..... .. ..... ..... . 3.7-13 3.7.7 The Main Turbine Bypass System .... .... .. ..... .. ...... ..... ... .... ... ... .......... ...... 3.7-14 (continued)

Cooper ii Amendment No.

TABLE OF CONTENTS (continued) 3.8 ELECTRICAL POWER SYSTEMS .... ..... .......................... .... .. .......... ... .. ...... 3.8-1 3.8.1 AC Sources - Operating ........... ..... ............ .... .... ..... ... .... ..... ............. .. ... .. 3.8-1 3.8.2 AC Sources - Shutdown .... ....... ... ... ........... ......... ...... .... ........... ....... ...... .. 3.8-10 3.8.3 Diesel Fuel Oil, Lube Oil , and Starting Air ... .... ... .... ... ... ....... .. ........... ..... . 3.8-13 3.8.4 DC Sources - Operating ... ............ .... ............ ...... .... .......... ... .... .............. . 3.8-16 3.8.5 DC Sources - Shutdown ..... ................... ..... ...... ...................................... 3.8-20 3.8.6 Battery Cell Parameters ............... .............. ...... ..... ... .... ... .... .. ...... ........... 3.8-22 3.8.7 Distribution Systems - Operating .............. ...... ... ............. .. ..... .. ... .... ........ 3.8-26 3.8.8 Distribution Systems - Shutdown ..... .. .............. .. .... .. .. .... ... .... .. ..... ..... .... . 3.8-29 3.9 REFUELING OPERATIONS .. ... ...... ........ ... .. ..... .......... ......... ..... .. .. ... .. .......... 3.9-1 3.9.1 Refueling Equipment Interlocks .. .. .... .... ... .. ...... .. ... ... .. ....... ...... ... .... ......... 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock .... ... ........... .... ....... .......... ... ... .. ... 3.9-3 3.9.3 Control Rod Position ..... ....... .... ......... ......... ..... ...... ....................... .......... 3.9-4 3.9.4 Control Rod Position Indication ............ .... .... .. ...... ........ .. .. .... ... ... .... .. .. ... . 3.9-5 3.9.5 Control Rod OPERABILITY - Refueling ..... ...... ...... ......... .. ... ...... ........ .... 3.9-7 3.9.6 Reactor Pressure Vessel (RPV) Water Level ........... ..... .. ............. ..... .. ... 3.9-8 3.9.7 Residual Heat Removal (RHR) - High Water Level .. ........ ...... ... ... ....... .. . 3.9-9 3.9.8 Residual Heat Removal (RHR) - Low Water Level .... .. .... .. .. ... .... ....... ..... 3.9-12 3.10 SPECIAL OPERATIONS .. .... .... ...... .. .... ........ ............ .. ........... ..... ..... .......... 3.10-1 3.10 .1 lnservice Leak and Hydrostatic Testing Operation ... ... ... ..... .. .... ... ... ..... 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing .... .. .. .................... .. ...... ............ 3.10-4 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ... .... ..... ... .. ..... .... ... ..... . 3.10-6 3.10.4 Single Control Rod Withdrawal - Cold Shutdown ..... ........ .. .... .... .. .. ... ... 3.10-9 3.10.5 Single Control Rod Drive (CRD) Removal - Refueling ..... .. .... ..... ... ....... 3.10-13 3.10.6 Multiple Control Rod Withdrawal - Refueling ... .. ....... .. ...... .... ..... ... ... ..... 3.10-16 3.10.7 Control Rod Testing - Operating .. ........ ... ...... .. ... ...... ................ .. .......... 3.10-18 3.10.8 SHUTDOWN MARGIN (SOM ) Test - Refueling ... .. ..... ........ .......... ....... 3.10-20 4.0 DESIGN FEATURES ................ .......... ..... .. ..... .. ....... .. ...... .. ........ .. .. .... ... ... .. .. ... ... 4.0-1 4.1 Site Location ..... .. .. .. ....... .. .. ....................... ........ .... .. ..... .. .... ........ ........ .. ... ... .. 4.0-1 4.2 Reactor Core .... ..... ................................... .. .... .. ............... .. .. ............ ... ... ..... . 4.0-1 4.3 Fuel Storage .. ..... ................................................ .. ..... .. ..... .... ... ..... .. ... ... ....... 4.0-2 5.0 ADMINISTRATION CONTROLS .................. ........ ...... ....................................... 5.0-1 5.1 Responsibility ........ .......... ..... ...... ........ ..... .. .... ... ... ... ... .. ..... ..... ... ... .. .. ....... ..... 5.0-1 5.2 Organization .... ..... .... ......... ....... .... .. .. ................... ............. ... .. ..... ......... ... .. ... 5.0-2 5.3 Unit Staff Qualifications ............ ......... .. ......... ... ...... ........... .... ........ ........ ....... 5.0-4 5.4 Procedures .. .. ........ ..... .... .......... .... .... ....... .. .. ..... .... ...... ..... ..... ...... ..... ..... ....... 5.0-5 5.5 Programs and Manuals ..... .. ............... ... .... .. .... .. ......... ... ... .. .... .. ... ... .. ... .... ..... 5.0-6 5.6 Reporting Requirements ........... .... ... ....... ..... ..... .. .. .... ..... ...... ...... ...... .. ... ....... 5.0-20 5.7 High Radiation Area ....... ....... .......... ............ .. ...... ..... .. .. .......... .. ..... ..... ... ....... 5.0-24 Cooper iii Amendment No.

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 1-133, 1-134, and 1-135 actually present. The DOSE (continued) EQUIVALENT 1-131 concentration is calculated as follows:

DOSE EQUIVALENT 1-131=(l-131)+0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (1-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

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 (T AF) 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 in continuous communication with the control room , is stationed at the controls, and is capable of closing the penetration flow path isolation devices without offsite power.

(continued)

Cooper 1.1-3 Amendment No.

Definitions 1.1 1.1 Definitions DRAIN TIME (continued) c. The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory;

d. No additional draining events occur; and

e. Realistic cross-sectional areas and drain rates are used .

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

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

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;

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

(continued)

Cooper 1.1-4 Amendment No.

Definitions 1.1 1.1 Definitions LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length of RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length .

LOGIC SYTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST 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 (CPR)

RATIO (MCPR) 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).

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2419 MWt.

(continued)

Cooper 1.1-5 Amendment No.

Definitions 1.1 1.1 Definitions REACTOR PROTECTION The RPS RESPONSE TIME shall be that time segment from SYSTEM(RPS)RESPONSE the time the sensor contacts actuate to the time the scram TIME solenoid valves deenergize.

SHUTDOWN MARGIN (SOM) SOM 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 2: 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 SOM .

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 RESPONSE TIME consists of two components:

a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established ; and

b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.

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.

Cooper 1.1-6 Amendment No.

Definitions 1.1 Table 1.1-1(page1of1)

MODES MODE TITLE REACTOR MODE AVERAGE REACTOR SWITCH POSITION COOLANT TEMPERATURE (oF) 1 Power Operation Run NA 2 Startup Refuel(a) or Startup/Hot NA Standby 3 Hot Shutdown(a) Shutdown > 212 4 Cold Shutdown(a) Shutdown $ 212 5 Refueling{bl Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

Cooper 1.1-7 Amendment No.

ECCS Instrumentation 3.3.5. 1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.1-1 .

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.1-1 for the channel.

B. As required by Required B.1 --------------NOTE-------------

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

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

Cooper 3.3-31 Amendment No.

ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 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 (HPCI) discovery of loss of System inoperable. HPCI initiation capability AND B.3 Place channel in trip . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. As required by Required C.1 --------------NO"TE-------------

Action A.1 and referenced in Only applicable for Table 3.3.5.1-1 . Functions 1.c, 1.e, 2.c, 2.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 when discovery of loss of its redundant feature initiation capability ECCS initiation capability for feature(s) in both is inoperable. 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)

Cooper 3.3-32 Amendment No.

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 --------------NOTE-------------

Action A. 1 and referenced in Only applicable if HPCI Table 3.3.5.1-1 . pump suction is not aligned to the suppression pool.

Declare HPCI System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable. discovery of loss of HPCI initiation capability AND D.2.1 Place channel in trip . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR D.2.2 Align the HPCI pump 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months suction to the suppression pool.

E. As required by Required E.1 --------------NOTE-------------

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

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

Cooper 3.3-33 Amendment No.

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 6)

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 Level 1,2,3 B SR 3.3.5.1.1 ~ -113 inches

- Low Low Low (Level 1) SR 3.3.5.1.2 SR 3.3.5.1.4(cXdl SR 3.3.5.1.5

b. Drywell Pressure - High 1,2,3 B SR 3.3.5.1.2 s 1.84 psig SR 3.3.5.1.4(cXdl SR 3.3.5.1.5

c. Reactor Pressure - Low 1,2,3 4 c SR 3.3.5.1.2 ~ 291 psig and (Injection Permissive) SR 3.3.5.1.4 SR 3.3.5.1.5 s 436 psig

d. Core Spray Pump 1,2,3 1 per pump E SR 3.3.5.1.2 ~ 1370 gpm Discharge Flow - Low SR 3.3.5.1.4(cXdl (Bypass) SR 3.3.5.1.5

e. Core Spray Pump Start - 1,2,3 1 per pump c SR 3.3.5.1.2 ~ 9 seconds Time Delay Relay SR 3.3.5.1.4 and SR 3.3.5.1 .5 s 11 seconds

2. Low Pressure Coolant Injection (LPCI ) System

a. Reactor Vessel Water Level 1,2,3 4 B SR 3.3.5.1 .1 ~ -113 inches

- Low Low Low (Level 1) SR 3.3.5.1.2 SR 3.3.5.1.4(cXdl SR 3.3.5.1.5 continued (a) [Deleted]

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

(c) If the as-found channel setpoint is outside its predefined as-found tolerance , then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(d) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance, otherwise, the channel shall be declared inoperable. Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-tound and the as-left tolerances are specified in the Technical Requirements Manual.

Cooper 3.3-37 Amendment No.

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)

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)

b. Drywell Pressure - High 1,2,3 4 B SR 3.3.5.1.2 :S 1.84 psig SR 3.3.5.1.4(cXdl SR 3.3.5.1.5

c. Reactor Pressure - Low 1,2,3 4 c SR 3.3.5.1.2 ~ 291 psig and (Injection Permissive) SR 3.3.5.1.4 SR 3.3.5.1.5 :S 436 psig

d. Reactor Pressure - Low 4 c SR 3.3.5.1.2 ~ 199 psig and (Recirculation Discharge SR 3.3.5.1.4 :S 246 psig Valve Permissive ) SR 3.3.5.1.5

e. Reactor Vessel 1,2,3 2 B SR 3.3.5.1.1 ~ -193.19 inches Shroud Level - Level 0 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

f. Low Pressure 1,2,3 1 per pump c SR 3.3.5.1 .2 Coolant Injection Pump SR 3.3.5.1.4 Start - Time Delay Relay SR 3.3.5.1.5 Pumps B,C ~ 4.5 seconds and

S 5.5 seconds PumpsA,D :S 0.5 second continued (a) [Deleted]

(c) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

{d) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance, otherwise, the channel shall be declared inoperable. Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.

(e) With associated recirculation pump discharge valve open .

Cooper 3.3-38 Amendment No.

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FU NCTION CONDITIONS FUNCTION ACTION A.1 RE QU IREMENTS VALUE 2 . LPCI System (continued )

g. Low Pressure Coolant 1,2,3 1 per subsystem E SR 3.3.5.1.2 :2: 2107 gpm Injection Pump Discharge SR 3.3.5.1.41cXdl Flow - Low (Bypass) SR 3.3 .5.1.5

h. Containment Pressure - 1,2 ,3 4 B SR 3 .3.5.1.2 :2: 2 psig High SR 3.3.5.1.4 SR 3.3.5.1.5

3. High Pressure Coolant Injection (H PCI ) System

a. Reactor Vessel Water Level 1, 4 B SR 3.3.5.1.1 :2: -42 inches

- Low Low (Level 2) SR 3.3.5.1 .2 2m, 31fJ SR 3.3.5.1.41cXdl SR 3.3.5.1.5

b. Drywell Pressure - High 1, 4 B SR 3.3.5.1.2 s 1.84 psig SR 3.3.5.1.41cXdl 21fJ. 31JJ SR 3.3.5.1 .5

c. Reactor Vessel Water Level 1, 2 c SR 3.3.5.1.1 s 54 inches

- High (Level 8) SR 3.3.5.1.2 SR 3.3.5.1.4 21JJ. 31JJ SR 3.3.5.1.5

d. Emergency Condensate 1, 2 D SR 3.3.5.1 .2 :2: 23 inches Storage Tank (ECST) Level - SR 3.3.5.1 .3 Low 21JJ, 31fJ SR 3.3.5.1.5

e. Suppression Pool Water 1, 2 D SR 3.3.5.1.2 s 4 inches Level - High SR 3.3.5.1.4 21JJ, 311) SR 3.3.5.1.5 contin ued (a) [Deleted]

(c) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as req uired before returning the channel to service.

(d) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Lim iting Trip Setpoint (L TSP ) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance . The Limiting Trip Setpoint and the methodologies used to determine the as-tou nd and the as-left tolerances are specified in the Technical Requirements Manual.

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

Cooper 3.3-39 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5.3 3.3 INSTRUMENTATION 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.3 The RPV Water Inventory Control Instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.3-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

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

B. As required by Required B.1 Declare supported Immediately Action A.1 and referenced in penetration flow path(s)

Table 3.3.5.3-1. incapable of automatic isolation.

AND B.2 Calculate DRAIN TIME. Immediately C. As required by Required C.1 Place channel in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action A.1 and referenced in Table 3.3.5.3-1.

(continued) I Cooper 3.3-47 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Action A.1 and referenced in OPERABLE status.

Table 3.3.5.3-1 .

E. Required Action and E.1 Declare associated low Immediately associated Completion Time pressure ECCS of Condition C or D not met. injection/spray subsystem inoperable.

SURVEILLANCE REQUIREMENTS

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

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

SURVEILLANCE FREQUENCY SR 3.3.5.3.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-48 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5.3 Table 3.3.5.3-1(page1of1)

RPV Water Inventory Control 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 Pressure - Low 4,5 4 c SR 3.3.5.3 .2 :s 436 psig (Injection Permissive)

b. Core Spray Pump Discharge 4,5 1 per pump(*) D SR 3.3.5.3.2 ~ 1370 gpm Flow - Low (Bypass)

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Pressure - Low 4,5 4 c SR 3.3.5.3.2 :s 436 psig (Injection Permissive)

b. Low Pressure Coolant 4,5 1 per D SR 3.3.5.3.2 ~ 2107 gpm Injection Pump Discharge subsystem(*>

Flow - Low (Bypass)

3. RHR System Isolation (b)

a. Reactor Vessel Water 2 in one trip B SR 3.3.5.3.1 ~ 3 inches Level - Low, Level 3 system SR 3.3.5.3.2

4. Reactor Water Cleanup (RWCU)

System Isolation (b)

a. Reactor Vessel Water Level 2 in one trip B SR 3.3.5.3.1 ~ -42 inches

- Low Low, Level 2 system SR 3.3.5.3.2 (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "Reactor Pressure Vessel Water Inventory Control."

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

Cooper 3.3-49 Amendment No.

Primary Containment Isolation Instrumentation 3.3.6.1 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation LCO 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6.1-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for channels inoperable. Functions 2.a, 2.b, 5.d, and 6.b AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Functions other than Functions 2.a, 2.b, 5.d, and 6.b B. One or more Functions with B.1 Restore isolation 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolation capability not capability.

maintained.

(continued)

Cooper 3.3-50 Amendment No.

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Enter the Condition Immediately associated Completion Time referenced in Table of Condition A or B not met. 3.3.6.1-1 for the channel.

D. As required by Required D.1 Isolate associated main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action C.1 and referenced steam line (MSL).

in Table 3.3.6.1-1 .

OR D.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND D.2.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. As required by Required E.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.1 and referenced in Table 3.3.6.1-1.

F. As required by Required F.1 Isolate the affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action C.1 and referenced in penetration flow path(s).

Table 3.3.6.1-1 .

(continued)

Cooper 3.3-51 Amendm ent No.

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action C.1 and referenced in Table 3.3.6.1-1 . AND OR G.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and associated Completion Time for Condition F not met.

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

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

I. As required by Required I. 1 Initiate action to restore Immediately Action C.1 and referenced in channel to OPERABLE Table 3.3.6.1- 1. status.

Cooper 3.3-52 Amendment No.

Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS

N()TES----------------------------------------------------------

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

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.2 Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.3 Perform CHANNEL CALIBRATl()N . In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.4 -------------------------------N()TE--------------------------------

For Function 2.d, rad iation detectors are excluded.

Perform CHANNEL CALIBRATl()N. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.5 Calibrate each radiation detector. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.6 Perform L()GIC SYSTEM FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-53 Amendment No.

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1(page1of3)

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

1. Main Steam Line Isolation

a. Reactor Vessel Water Level

- Low Low Low (Level 1) 1,2,3 2 D SR 3.3.6.1.1 ~ -113 inch es SR 3.3.6.1 .2 SR 3.3.6.1.4 SR 3.3.6.1.6

b. Main Steam Line Pressure - 2 E SR 3.3.6.1.2 ~ 835 psig Low SR 3.3.6.1.3 SR 3.3.6.1.6

c. Main Steam Line Flow - High 1,2,3 2 per MSL D SR 3.3.6.1.2 s 142.7% rated SR 3.3.6.1.4 steam flow SR 3.3.6.1 .6

d. Condenser Vacuum - Low 2 D SR 3.3.6.1.2 ~ 8 inches Hg SR 3.3.6.1.3 vacuum SR 3.3.6.1.6

e. Main Steam Tunnel 1,2,3 2 per location D SR 3.3.6.1.2 s 195°F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

2. Primary Containment Isolation

a. Reactor Vessel Water Level

- Low (Level 3) 1,2,3 2 G SR 3.3.6.1.1 ~ 3 inches SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3 .6.1 .6

b. Drywell Pressure - High 1,2,3 2 G SR 3.3.6.1.2 s 1.84 psig SR 3.3.6.1.4 SR 3.3.6.1.6

c. Reactor Building 1,2,3 2 F SR 3.3.6.1.1 s 49 mR/hr Ventilation Exhaust SR 3.3.6.1 .2 Plenum Radiation - High SR 3.3.6.1.4 SR 3.3.6.1.6

d. Main Steam Line 1,2,3 2 F SR 3.3.6.1.1 s 3 times Radiation - High SR 3.3.6.1.2 full power SR 3.3.6.1.4 background SR 3.3.6.1.5 SR 3.3.6.1.6

e. Reactor Vessel Water Level 1,2,3 2 F SR 3.3.6.1 .1 ~ -113 inches

- Low Low Low (Level 1) SR 3.3.6.1 .2 SR 3.3.6.1.4 SR 3.3.6.1.6 (continued)

(a) With any turbine stop valve not closed .

Cooper 3.3-54 Amendment No.

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1(page2 of 3)

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

3. High Pressure Coolant Injection (H PCI) System Isolation

a. HPCI Steam Line Flow - 1,2,3 F SR 3.3.6.1.2 :S 250% rated High SR 3.3.6.1.4 steam flow SR 3.3.6.1 .6

b. HPCI Steam Line Flow - 1,2,3 F SR 3.3.6.1.2 :S 6 seconds Time Delay Relays SR 3.3.6.1.4 SR 3.3.6.1.6

c. HPCI Steam Supply Line 1,2,3 2 F SR 3.3.6.1.2 ~ 107 psig Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6

d. HPCI Steam Line Space 1,2,3 2 per location F SR 3.3.6.1.2 :S 195°F Temperature - High SR 3.3.6.1 .4 SR 3.3.6.1.6

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line Flow - 1,2,3 F SR 3.3.6.1.2 :S 288% rated High SR 3.3.6.1.4 steam flow SR 3.3.6.1.6

b. RCIC Steam Line Flow - 1,2,3 F SR 3.3.6.1.2 :S 6 seconds Time Delay Relays SR 3.3.6.1.4 SR 3.3.6.1.6

c. RCIC Steam Supply Line 1,2,3 2 F SR 3.3.6.1.2 ~ 61 psig Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1 .6

d. RCIC Steam Line Space 1,2,3 2 per location F SR 3.3.6.1.2 :S 195°F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

5. Reactor Water Cleanup (RWCU)

System Isolation

a. RWCU Flow - High 1,2,3 F SR 3.3.6.1.2 :S 191% of Rated SR 3.3.6.1.4 SR 3.3.6.1.6

b. RWCU System Space 1,2,3 2 per location F SR 3.3.6.1.2 :S 195°F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

c. SLC System Initiation 1,2 H SR 3.3.6.1.6 NA

d. Reactor Vessel Water Level 1,2,3 2 F SR 3.3.6.1.1 ~ - 42 inches

- Low Low (Level 2) SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 Cooper 3.3-55 Amendment No.

Primary Containment Isolation Instrumentation 3.3.6.1 Table3.3.6.1-1 (page3of3)

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

6. RHR Shutdown Cooling System Isolation

a. Reactor Pressure - High 1,2,3 F SR 3.3.6.1.2 s 72 psig SR 3.3.6.1.4 SR 3.3.6.1.6

b. Reactor Vessel Water 3 2 SR 3.3.6.1.1 ~ 3 inches Level - Low (Level 3) SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 Cooper 3.3-56 Amendment No.

Secondary Containment Isolation Instrumentation 3.3.6.2 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation LCO 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6.2-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for inoperable. Functions 1 and 2 AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Function 3 B. One or more Functions with B.1 Restore secondary 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months secondary containment containment isolation isolation capability not capability.

maintained .

C. Required Action and C.1.1 Isolate the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion Time secondary containment of Condition A or B not met. penetration flow path( s ).

OR (continued)

Cooper 3.3-57 Amendment No.

Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months secondary containment isolation valves inoperable .

AND C.2 .1 Place the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months standby gas treatment (SGT} subsystem(s) in operation.

OR C.2.2 Declare associated SGT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months subsystem(s) inoperable.

SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function .

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains secondary containment isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.3-58 Amendment No.

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-59 Amendment No.

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 TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water Level - Low Low 1,2,3 2 SR 3.3.6.2.1 2: - 42 inches (Level 2) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4

2. Drywell Pressure - High 1,2,3 2 SR 3.3.6.2 .2 s 1.84 psig SR 3.3.6.2 .3 SR 3.3.6.2.4

3. Reactor Building Ventilation Exhaust 1,2,3, 2 SR 3.3.6.2.1 s 49 mR/hr Plenum Radiation - High {b) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 (a) (Deleted)

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

Cooper 3.3-60 Amendment No.

LLS Instrumentation 3.3.6.3 3.3 INSTRUMENTATION 3.3.6.3 Low-Low Set (LLS) Instrumentation LCO 3.3.6.3 The LLS valve instrumentation for each Function in Table 3.3.6.3-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One LLS valve inoperable A.1 Restore channel( s) to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months due to inoperable OPERABLE status.

channel( s ).

B. Required Action and B.1 Declare the associated Immediately associated Completion Time LLS valve( s) inoperable.

of Condition A not met.

OR Two LLS valves inoperable due to inoperable channels.

Cooper 3.3-61 Amendment No.

LLS Instrumentation 3.3.6.3 SURVEILLANCE REQUIREMENTS

N()TES----------------------------------------------------------

1. Refer to Table 3.3.6.3-1 to determine which SRs apply for each Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains LLS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.3.1 Perform CHANNEL FUNCTl()NAL TEST for portion In accordance with of the channel outside primary containment. the Surveillance Frequency Control Program SR 3.3.6.3.2 -------------------------------N()TE--------------------------------

Only required to be performed prior to entering MODE 2 during each scheduled outage > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when entry is made into primary containment.

Perform CHANNEL FUNCTl()NAL TEST for portions In accordance with of the channel inside primary containment. the Surveillance Frequency Control Program SR 3.3.6.3.3 Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.3.4 Perform CHANNEL CALIBRATl()N. In accordance with the Surveillance Frequency Control Program SR 3.3.6.3.5 Perform L()GIC SYSTEM FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-62 Amendment No.

LLS Instrumentation 3.3.6.3 Table 3.3.6.3-1 (page1of 1)

Low - Low Set Instrumentation REQUIRED CHANNELS PER SURVEILLANCE ALLOWABLE FUNCTION FUNCTION REQUIREMENTS VALUE

1. Reactor Pressure - High 1 per LLS valve SR 3.3.6.3.3 s 1050 psig SR 3.3.6.3.4 SR 3.3.6.3.5

2. Low - Low Set Pressure Setpoints 2 per LLS valve SR 3.3.6.3.3 Low:

SR 3.3.6.3.4 Open~ 966.5 SR 3.3.6.3.5 psig ands 1010 psig Close ~ 835 psig and s 875 .5 psig High:

Open~ 996.5 psig and s 1040 psig Close ~ 835 psig and s 875.5 psig

3. Discharge Line Pressure Switch 1 per SRV SR 3.3.6.3.1 ~ 25 psig and SR 3.3.6.3.2 s 55 psig SR 3.3.6.3.4 SR 3.3.6.3.5 Cooper 3.3-63 Amendment No.

CREF System Instrumentation 3.3.7.1 3.3 INSTRUMENTATION 3.3. 7.1 Control Room Emergency Filter (CREF) System Instrumentation LCO 3.3.7.1 The CREF System instrumentation for each Function in Table 3.3.7.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.7.1-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for inoperable. Functions 1 and 2 AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Function 3 B. One or more Functions B.1 Restore CREF System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with CREF System initiation capability.

initiation capability not maintained .

C. Required Action and C.1 Initiate CREF System. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion Time not met. OR C.2 Declare CREF System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable.

Cooper 3.3-64 Amendment No.

CREF System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS

N()TES----------------------------------------------------------

1. Refer to Table 3.3.7.1-1 to determine which SRs apply for each CREF Function .

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains CREF initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.2 Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.7.1 .3 Perform CHANNEL CALIBRATl()N . In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.4 Perform L()GIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-65 Amendment No.

CREF System Instrumentation 3.3.7.1 Table 3.3.7.1-1(page1of1)

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

1. Reactor Vessel Water Level - Low Low 1,2,3 2 SR 3.3.7.1.1 ~ - 42 i nches (Level 2) SR 3.3.7.1.2 SR 3.3.7.1 .3 SR 3.3.7.1.4

2. Drywell Pressure - High 1,2,3 2 SR 3.3.7.1.2 s 1.84 psig SR 3.3.7.1.3 SR 3.3.7.1.4

3. Reactor Building Ventilation Exhaust 1,2,3, 2 SR 3.3.7.1 .1 s 49 mR/hr Plenum Radiation - High (b) SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 (a) [Deleted]

(b) During movement of lately irradiated fuel assemblies in the secondary containment.

Cooper 3.3-66 Amendment No.

LOP Instrumentation 3.3.8.1 3.3 INSTRUMENTATION 3.3.8.1 Loss of Power (LOP) Instrumentation LCO 3.3.8.1 The LOP instrumentation for each Function in Table 3.3.8.1-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generator is required to be OPERABLE by LCO 3.8.2, "AC Sources - Shutdown."

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. OPERABLE status.

B. Required Action and B.1 Declare associated diesel Immediately associated Completion Time generator (DG) not met. inoperable.

Cooper 3.3-67 Amendment No.

LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function .

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided the associated Function maintains DG initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.8.1 .1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.8.1 .2 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-68 Amendment No.

LOP Instrumentation 3.3.8.1 Table 3.3.8.1-1(page1of1)

Loss of Power Instrumentation REQU IRED CHANNELS PER SURVEILLANCE ALLOWABLE FUNCTION BUS REQUIREMENTS VALUE

1. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)

a. Bus Undervoltage SR 3.3.8.1.1 ~ 2185 V and SR 3.3.8.1.2 :S 2415 v SR 3.3.8.1.3

b. Time Delay SR 3.3.8.1.1 ~ 0 seconds and SR 3.3.8.1 .2 :S 5 seconds SR 3.3.8.1. 3

2. 4.16 kV Emergency Bus Normal Supply Undervoltage (Loss of Voltage)

a. Bus - Tie Undervoltage SR 3.3.8.1 .1 ~ 2185 Vand SR 3.3.8.1.2 :S 2415 v SR 3.3.8.1 .3

b. Time Delay SR 3.3.8.1 .1 ~ 0 seconds and SR 3.3.8.1.2 :S 5 seconds SR 3.3.8.1.3

3. 4.16 kV Emergency Bus ESST Supply Undervoltage (Loss of Voltage)

a. Bus - Tie Undervoltage SR 3.3.8.1.1 ~ 2185 V and SR 3.3.8.1 .2 :S 2415 v SR 3.3.8.1.3

b. Time Delay SR 3.3.8.1.1 ~ 0 seconds and SR 3.3.8.1.2 :S 5 seconds SR 3.3.8.1.3

4. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage)

a. Bus Undervoltage 2 SR 3.3.8.1.1 ~ 3828 V and SR 3.3.8.1.2 :S 3932 v SR 3.3.8.1.3

b. Time Delay (LOCA) 2 SR 3.3.8.1.1 ~ 6.7 seconds and SR 3.3.8.1.2 :S 8.3 seconds SR 3.3.8.1.3

c. Time Delay (Non-LOCA) SR 3.3.8.1 .1 ~ 11 .2 seconds and SR 3.3.8.1 .2 :S 13.8 seconds SR 3.3.8.1.3

5. 4.16 kV Emergency Bus ESST Supply Undervoltage (Degraded Voltage)

a. Bus Undervoltage SR 3.3.8.1 .1 ~ 3828 V and SR 3.3.8.1 .2 :S 3932 v SR 3.3.8.1.3

b. Time Delay SR 3.3.8.1.1 ~ 13.4 seconds and SR 3.3.8.1.2 :S 16.6 seconds SR 3.3.8.1.3 Cooper 3.3-69 Amendment No.

RPS Electric Power Monitoring 3.3.8.2 3.3 INSTRUMENTATION 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring LCO 3.3.8.2 Two RPS electric power monitoring assemblies shall be OPERABLE for each inservice RPS motor generator set or alternate power supply.

APPLICABILITY: MODES 1 and 2, MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both inservice power A.1 Remove associated 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months supplies with one electric inservice power supply(s) power monitoring assembly from service.

inoperable.

B. One or both inservice power B.1 Remove associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months supplies with both electric inservice power supply(s) power monitoring from service.

assemblies inoperable .

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 or B not met in MODE 1 or 2.

(continued)

Cooper 3.3-70 Amendment No.

RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to fully insert Immediately associated Completion Time all insertable control rods of Condition A or B not met in core cells containing in MODE 5 with any control one or more fuel rod withdrawn from a core assemblies.

cell containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.2.1 Perform CHANNEL CALIBRATION. The Allowable In accordance with Values shall be: the Surveillance Frequency Control

a. Overvoltage::;; 131 V with time delay set to Program

3.8 seconds.

b. Undervoltage ~ 109 V, with time delay set to

3.8 seconds.

c. Underfrequency ~ 57.2 Hz, with time delay set to s 3.8 seconds.

SR 3.3.8.2.2 Perform a system functional test. In accordance with the Surveillance Frequency Control Program Cooper 3.3-71 Amendment No.

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/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 ECCS A.1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable. subsystem( s) to operable status.

OR One LPCI pump in both LPCI subsystems inoperable.

B. Required Action and B.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 not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

Cooper 3.5-1 Amendment No.

RPV Water Inventory Control 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control LCO 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be

~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

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

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

APPLICABILITY: MODES 4 and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required ECCS A.1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray subsystem injection/spray subsystem inoperable. to OPERABLE status.

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

(continued)

Cooper 3.5-7 Amendment No.

RPV Water Inventory Control 3.5.2 ACTIONS (continued)

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 gas 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months treatment subsystem is capable of being placed in operation in less than the DRAIN 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 establish Immediately 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 (continued)

Cooper 3.5-8 Amendment No.

RPV Water Inventory Control 3.5.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.2 Initiate action to establish Immediately secondary containment boundary.

AND D.3 Initiate action to isolate Immediately each secondary containment penetration flow path or verify it can be manually isolated from the control room .

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

E. Required Action and E.1 Initiate action to restore Immediately associated Completion Time DRAIN TIME to :::: 36 of Condition C or D not met. hours.

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

Cooper 3.5-9 Amendment No.

RPV Water Inventory Control 3.5.2 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 SR 3.5.2.2 Verify, for a required ECCS injection/spray In accordance with subsystem , the suppression pool water level is ~ 12 ft the Surveillance 7 inches. Frequency Control Program SR 3.5.2.3 Verify, for the required ECCS injection/spray In accordance with subsystem , the piping is filled with water from the the Surveillance pump discharge valve to the injection valve. Frequency Control Program SR 3.5.2.4 Verify, for the required ECCS injection/spray In accordance with subsystem , each manual, power operated, and the Surveillance automatic valve in the flow path , that is not locked , Frequency Control sealed, or otherwise secured in position, is in the Program correct position.

SR 3.5.2.5 Operate the required ECCS injection/spray In accordance with subsystem through the recirculation line for ~ 10 the Surveillance minutes. Frequency Control Program SR 3.5.2.6 Verify each valve credited for automatically isolating a In accordance with penetration flow path actuates to the isolation position the Surveillance on an actual or simulate isolation signal. Frequency Control Program (continued)

Cooper 3.5-10 Amendment No.

RPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.2.7 --------------------------------N()TE-------------------------------

Vessel injection/spray may be excluded.

Verify the required ECCS injection/spray subsystem In accordance with can be manually operated. the Surveillance Frequency Control Program Cooper 3.5-11 Amendment No.

RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

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

ACTIONS

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

LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable. A.1 Verify by administrative 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months means High Pressure Coolant Injection System is OPERABLE.

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

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months dome pressure to ~ 150 psig .

Cooper 3.5-12 Amendment No.

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water In accordance with from the pump discharge valve to the injection valve. the Surveillance Frequency Control Program SR 3.5.3.2 Verify each RCIC System manual, power operated, In accordance with and automatic valve in the flow path , that is not the Surveillance locked, sealed, or otherwise secured in position, is in Frequency Control the correct position. Program SR 3.5.3.3 --------------------------------N()TE-------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1020 psig and ~ 920 In accordance with psig, the RCIC pump can develop a flow rate ~ 400 the Surveillance gpm against a system head corresponding to reactor Frequency Control pressure. Program SR 3.5.3.4 --------------------------------N()TE-------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 165 psig , the RCIC In accordance with pump can develop a flow rate ~ 400 gpm the Surveillance against a system head corresponding to reactor Frequency Control pressure. Program (continued)

Cooper 3.5-13 Amendment No.

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.3.5 -------------------------------N()TES------------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

2. Vessel injection may be excluded.

Verify the RCIC System actuates on an actual or In accordance with simulated automatic initiation signal. the Surveillance Frequency Control Program Cooper 3.5-14 Amendment No.

PC IVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more penetration D. 1 Restore leakage rate to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flowpaths with one or more within limit.

MSIVs not within leakage rate 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, B, C, or D AND 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 Cooper 3.6-11 Amendment No.

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, 2, or containment to

3. OPERABLE status.

B. Required Action and B.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 not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Secondary containment C.1 -------------NOTE--------------

inoperable during movement LCO 3.0.3 is not of recently irradiated fuel applicable.

assemblies in the secondary -----------------------------------

containment.

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

Cooper 3.6-34 Amendment No.

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is ~ 0.25 inch In accordance with of vacuum water gauge. the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify all secondary containment equipment hatches In accordance with are closed and sealed. the Surveillance Frequency Control Program SR 3.6.4.1.3 Verify one secondary containment access door in In accordance with each access opening is closed. the Surveillance Frequency Control Program SR 3.6.4.1.4 Verify each SGT subsystem can maintain ~ 0.25 inch In accordance with of vacuum water gauge in the secondary containment the Surveillance for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate s 1780 cfm . Frequency Control Program Cooper 3.6-35 Amendment No.

SC IVs 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 SCI Vs.

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

AND (continued)

Cooper 3.6-36 Amendment No.

SC IVs 3.6.4.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 --------------NOTE-------------

associated Completion Time LCO 3.0 .3 is not of Condition A or B not met applicable.

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

Cooper 3.6-38 Amendment No.

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 subsystem 7 days inoperable. to OPERABLE status.

B. Required Action and B.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 not met in AND MODE 1, 2, or 3.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Required Action and ---------------------NOTE-------------------

associated Completion Time LCO 3.0.3 is not applicable.

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

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

(continued)

Cooper 3.6-40 Amendment No.

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Two SGT subsystems D.1 Enter LCO 3.0.3 Immediately inoperable in MODE 1, 2, or 3.

E. Two SGT subsystems E.1 -------------NOTE--------------

inoperable during movement LCO 3.0.3 is not of recently irradiated fuel applicable.

assemblies in the secondary -----------------------------------

containment.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for ~ 10 continuous In accordance with hours with heaters operating. the Surveillance Frequency Control Program SR 3.6.4.3.2 Perform required SGT filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or In accordance with simulated initiation signal. the Surveillance Frequency Control Program SR 3.6.4.3.4 Verify the SGT units cross tie damper is in the correct In accordance with position, and each SGT room air supply check valve the Surveillance and SGT dilution air shutoff valve can be opened . Frequency Control Program Cooper 3.6-41 Amendment No.

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

N 0 TE---------------------------------------------------

Th e main control room envelope (CRE) boundary may be opened intermittently under administrative control.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CREF System inoperable for A.1 Restore CREF System to 7 days reasons other than OPERABLE status.

Condition B.

B. CREF System inoperable B.1 Initiate action to implement Immediately due to inoperable CRE mitigating actions.

boundary in MODE 1, 2, or

3. AND B.2 Verify mitigating actions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ensure CRE occupant exposures to radiological and chemical hazards will not exceed limits, and CRE occupants are protected from smoke hazards.

AND B.3 Restore CRE boundary to OPERABLE status. 90 days (continued)

Cooper 3.7-8 Amendment No.

CREF System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COM PLETION 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 or B not met AND 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 ---------------------NOTE--------------------

associated Completion Time LCO 3.0.3 is not applicable.

of Condition A not met -----------*------*--------------------------------

during movement of lately irradiated fuel assemblies in D. 1 Suspend movement of Immediately the secondary containment. lately irradiated fuel assemblies in the OR secondary contai nment.

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

Cooper 3.7-9 Amend ment No.

AC Sources - Shutdown 3.8.2 ACTIONS

N 0 TE--------------------------------------------------------

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit --------------------N 0 TE---------------------

inoperable. Enter applicable Condition and Required Actions of LCO 3.8.8, when any required division is de-energized as a result of Condition A.

A.1 Declare affected required Immediately feature(s), with no offsite power available, inoperable.

OR A.2.1 Suspend CORE Immediately AL TERATIONS .

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

AND A.2 .3 Initiate action to restore Immediately required offsite power circuit to OPERABLE status.

(continued)

Cooper 3.8-11 Amendment No.

AC Sources - Shutdown 3.8.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. AL TERATIONS.

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

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 -------------------------------N 0 TE---------------------------------

1. The following SRs are not required to be performed : SR 3.8.1 .3, and SR 3.8.1 .9 through SR 3.8.1.11 .

2. SR 3.8.1 .11 is considered to be met without the ECCS initiation signals OPERABLE when the ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1 .

For AC sources required to be OPERABLE the SRs In accordance with of Specification 3.8.1 , except SR 3.8.1 .8, are applicable SRs applicable.

Cooper 3.8-12 Amendment No.

DC Sources - Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 DC electrical power subsystems shall be OPERABLE to support the DC electrical power distribution subsystem(s) required by LCO 3.8.8, "Distribution Systems - Shutdown ."

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

ACTIONS

N 0 TE---------------------------------------------------------

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required DC A.1 Declare affected required Immediately electrical power feature( s) inoperable.

subsystems inoperable.

OR A.2.1 Suspend CORE Immediately AL TERATIONS.

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

A.2 .3 Initiate action to restore Immediately required DC electrical power subsystems to OPERABLE status.

Cooper 3.8-20 Amendment No.

DC Sources - Shutdown 3.8.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 ----------------------------------N 0 TE------------------------------

T he following SRs are not required to be performed :

SR 3.8.4.7 and SR 3.8.4.8.

For DC sources required to be OPERABLE, the In accordance with following SRs are applicable: applicable SRs SR 3.8.4.1 SR 3.8.4.4 SR 3.8.4.7 SR 3.8.4.2 SR 3.8.4.5 SR 3.8.4.8 SR 3.8.4.3 SR 3.8.4.6 Cooper 3.8-21 Amendment No.

Distribution Systems - Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems - Shutdown LCO 3.8.8 The necessary portions of the AC and DC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

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

ACTIONS

N 0 TE-----------------------------------------------------------

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required AC A.1 Declare associated Immediately or DC electrical power supported required distribution subsystems feature( s) inoperable.

inoperable.

OR A.2 .1 Suspend CORE Immediately ALTERATIONS.

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

(continued)

Cooper 3.8-29 Amendment No.

Distribution Systems - Shutdown 3.8.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2.3 Initiate actions to restore Immediately 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 voltage to In accordance with required AC and DC electrical power distribution the Surveillance subsystems. Frequency Control Program Cooper 3.8-30 Amendment No.

NLS2017048 Page 1of66 Attachment 4 Proposed Technical Specifications Bases Changes (Mark-up) -

Information Only Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 B 3.3-94 B 3.3-161 B 3.6-76 B 3.3.96 B 3.3-162 B 3.6-79 B 3.3-97 B3 .3-176 B 3.6-84 B 3.3-98 B 3.3-177 B 3.6-85 B 3.3-99 B 3.5-1 B 3.6-86 B3.3-101 B 3.5-6 B 3.7-20 B 3.3-109 B 3.5-18 B 3.7-21 B3 .3-lll B 3.5-19 B 3.7-22 B 3.3-113 B 3.5-20 B 3.8-23 B 3.3-132 B 3.5-21 B 3.8-24 B3.3-133 B 3.5-22 B 3.8-25 B 3.3-134 B 3.5-23 B 3.8-26 B 3.3-135 B 3.5-24 B 3.8-27 B 3.3-136 B 3.5-25 B 3.8-28 B 3.3-137 B 3.5-26 B 3.8-47 B 3.3-138 B 3.5-27 B 3.8-48 B 3.3-139 B 3.5-28 B 3.8-49 B 3.3-149 B3.6-18 B 3.8-63 B 3.3-150 B 3.6-23 B 3.8-64 B 3.3-153 B 3.6-61 B 3.8-65 B 3.3-154 B 3.6-71 B 3.10-2 B 3.3-160 B 3.6-72

ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND (continued) of Power (LOP) Instrumentation," for a discussion of these signals.) 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. The DG initiation signal is a sealed in signal and must be manually reset. The DG initiation logic is reset by resetting the associated ECCS initiation logic. Upon receipt of a loss of coolant accident (LOCA) initiation signal, each DG is automatically started, is ready to load in approximately 14 seconds, and will run in standby conditions (rated voltage and speed, with the DG output breaker open). The DGs will only energize their respective Engineered Safety Feature buses if a loss of offsite power occurs. (Refer to Bases for LCO 3.3.8.1.)

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The actions of the ECCS are explicitly assumed in the safety analyses of References 6, 7, and 8. The ECCS is initiated to preserve the integrity of the fuel cladding by limiting 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) (Ref.

5). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

Permissive and interlock setpoints allow the blocking of trips during plant startups, and restoration of trips when the permissive conditions are not satisfied, but they are not explicitly modeled in the Safety Analysis.

These permissives and interlocks ensure that the starting conditions are consistent with the safety analysis, before preventive or mitigating actions occur. Because these permissives or interlocks are only one of multiple conservative starting assumptions for the accident analysis, they are generally considered as nominal values without regard to measurement accuracy.

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 set within the setting tolerance of the specified LTSPs, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint metho assumptions. Table 3.3.5.1-1 contains ootnotes. Feotnote (8) olarifies that the assooiated funotions are required to be OPERABL ~ h. h MODES 4 Bnd 5 en ly when tl'leir supperted EGGS Bre re~uir~ w IC OPERABLE per LOO 3.5.2 , EGGS-SI rutdowr r. Footr rote (b), added to show that certain ECCS instrumentation Functions also perform DG initiation .

Cooper B 3.3-94 02/22/16

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSIS , LCO, and APPLICABILITY (continued)

(Level 1) is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in References 6 and

8. In addition, the Reactor Vessel Water Level-Low Low Low (Level 1)

Function is directly assumed in the analysis of the recirculation line break (Ref. 7). The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level-Low Low Low (Level 1) signals are initiated from four level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

The Reactor Vessel Water Level-Low Low Low (Level 1) Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling.

Four channels of Reactor Vessel Water Level-Low Low Low (Level 1)

Function are only required to be OPERABLE when the ECCS are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Per FootRote (a) to Tobie 3.3.5.1 1, tl'~is EGGS function is on l~ required to be OPERABLE in MODES 4 8nd 5 whenever the assgciated ECCS is required tg be OPERABbE per bCO 3.5.2. Refer to LGO a.5.1 and LGO 3.5.2, "EGGS St=lutdown," for Applicability Bases for the low pressure E::GGS subsysterns; LGO J.8.1, "AC Smm;es Operating"; and bCO J .8.2, "AC Sources Shutdown,"

for Applicability Bases for the DGs.

1.b, 2.b. Drvwell Pressure-High High pressure in the drywall 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 Pressure-High Function in order to minimize the possibility of fuel damage. The DGs are initiated from Function 1.b signals. The Drywell Pressure-High Function, along with the Reactor Water Level-Low Low Low (Level 1) Function, is directly assumed in the analysis of the recirculation line break (Ref. 8). 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 drywall 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.

Cooper B 3.3-96 0212211e I

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSIS, LCO, and APPLICABILITY (continued)

The Drywell Pressure-High 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 requ ired to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure-High Function are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude ECCS and DG initiation. In MODES 4 and 5, the Drywell Pressure-High 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.

1.c, 2.c. Reactor Pressure-Low (Injection Permissive)

Low reactor 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 and a break in the RCPB has occurred, respectively. The Reactor Pressure-Low is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 6 and 8. In addition , the Reactor Pressure-Low Function is directly assumed in the analysis of the recirculation line break (Ref. 7). 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 Pressure-Low signals are initiated from four pressure switches that sense the reactor 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.

Four channels of Reactor Pressure-Low 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 . Per Footnote (a) to Table a.a.e.1 1, this EGGS funotion is only required to be OPERABLE in MODES 4 Bfld S wl=teRe*vreF tl=te asseeiated EGGS is requiFed te be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 snd LCO a.6.2 for Applieability Bases for the low pressure EGGS subsystems.

Cooper B 3.3-97 02122116 I

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSIS, LCO, and APPLICABILITY (continued) 1.d, 2.g. Core Spray and Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass)

The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The LPCI and CS Pump Discharge Flow-Low Functions are assumed to be OPERABLE. The minimum flow valves for CS and LPCI are not required to close to ensure that the low pressure ECCS flows assumed during the transients and accidents analyzed in References 6, 7, and 8 are met.

The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

One flow transmitter per CS pump and one differential pressure switch per LPCI subsystem are used to detect the associated subsystems' flow rates. The logic is arranged such that each switch or transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for approximately 3.5 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the RHR shutdown cooling mode. The Pump Discharge Flow-Low Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump.

Each channel of Pump Discharge Flow-Low Function (two CS channels and four LPCI channels) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude the ECCS function. Per FootRote (a) to Table 3.3.5.1 1, this EGGS fuRetioR is oRly re~uiFed to be OPERABLE iR MODES 4 and e whenever the assooiated EGGS is required to be OPERABLE ~ef LGO 3.5.2. Ref-ef to LCO 3.5.1 aRd LGO 3.5.2 for Applicability Bases fer the low pressufe EGGS subsystems.

1.e. Core Spray Pump Start-Time Delay Relay The purpose of this time delay is to delay the start of the CS pumps to enable sequential loading of the appropriate AC source. This Function is necessary when power is being supplied from the offsite sources or the standby power sources (DG). The CS Pump Start-Time Delay Relays are assumed to be OPERABLE in the accident analyses requiring ECCS Cooper B 3.3-98 11/25/12

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSIS, LCO, and APPLICABILITY (continued) 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 Core Spray Pump Start-Time Delay Relays, one for each CS pump. Each time delay relay is dedicated to a single pump start logic, such that a single failure of a Core Spray Pump Start-Time Delay Relay will not result in the failure of more than one CS pump. In this condition, one of the two CS pumps will remain OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Value for the Core Spray Pump Start-Time Delay Relays is chosen to be long enough so that the power source will not be overloaded and short enough so that ECCS operation is not degraded.

Each channel of Core Spray Pump Start-Time Delay Relay Function is required to be OPERABLE only when the associated CS subsystem is required to be OPERABLE. Refer to LOO 3.5.1 and LOO 3.5.2 for Applicability Bases for tt:le GS subsystems.

2.d. Reactor Pressure-Low (Recirculation Discharge Valve Permissive)

Low reactor pressure signals are used as permissives for recirculation discharge valve closure. This ensures that the LPCI subsystems inject into the proper RPV location assumed in the safety analysis. The Reactor Pressure-Low is one of the Functions assumed to be OPERABLE and capable of closing the valve during the transients analyzed in References 6 and 8. 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 Pressure-Low Function is directly assumed in the analysis of the recirculation line break (Ref. 7).

The Reactor Pressure-Low signals are initiated from four pressure switches that sense the reactor dome pressure.

The Allowable Value is chosen high enough that the valves close prior to when LPCI injection flow into the core is required (as assumed in the safety analysis) and low enough to avoid excessive differential pressures.

Four channels of the Reactor Pressure-Low Function are only required to be OPERABLE in MODES 1, 2, and 3 with the associated recirculation pump discharge valve open. With the valve(s) closed, the function of the instrumentation has been performed; thus, the Function is not required .

In MODES 4 and 5, the loop injection location is not critical since LPCI Cooper B 3.3-99 11/25/12

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSIS , LCO, and APPLICABILITY (continued) are assumed to be OPERABLE in the accident 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 four LPCI Pump Start-Time Delay Relays, one in each of the RHR pump start logic circuits. While each time delay relay is dedicated to a single pump start logic, a single failure of a LPCI Pump Start-Time Delay Relay could result in the failure of the two low pressure ECCS pumps, powered for the same ESF bus, to perform their intended function (e.g., as in the case where both ECCS pumps on one ESF bus start simultaneously due to an inoperable time delay relay). This still leaves four 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 Value for the LPCI Pump Start-Time Delay Relays is chosen to be long enough so that most of the starting transient of the first pump is complete before starting the second pump on the same 4.16 kV emergency bus and short enough so that ECCS operation is not degraded .

Each LPCI Pump Start-Time Delay Relay Function is required to be OPERABLE only when the associated LPCI subsystem is required to be OPERABLE. Per Footnote (e) to Table 3.3.5.1* 1, tnis ECCS function is only requiree to be OPERABLE in MODES 4 and 6 *.vhenever the assooiated EGGS is required to be OPERABLE per LGO a.6.2. Refer to LGO a .6.1 and LGO a.6.2 for Applioability Bases for the LPGI subsystems.

2.h Containment Pressure - High The Containment Pressure - High Function serves as an interlock perm issive to allow the RHR System to be manually aligned from the LPCI mode to the containment spray mode after containment pressure has exceeded the trip setting . The permissive ensures that containment pressure is elevated before the manual transfer is allowed. This ensures that LPCI is available to prevent or minimize fuel damage until such time that the operator determines that containment pressure control is needed .

The Containment Pressure - High Function is implicitly assumed in the analysis of LOCAs inside containment (Ref. 10) since the analysis assumes that containment spray is manually initiated when containment pressure is high. Containment Pressure - High signals are initiated from four pressure switches that sense drywell pressure. The four instruments also provide an isolation of the containment spray mode of RHR on decreasing containment pressure following manual actuation of the system . This isolation function is not credited in accident analysis for mitigating excessive depressurization of the containment, therefore is not a TS function .

Cooper B 3.3-101 02/22/16

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS (continued) for the feature(s). Required Action B.1 features would be those that are initiated by Functions 1.a, 1.b, 2.a, 2.b, and 2.h (e.g., low pressure ECCS). The Required Action B.2 system would be HPCI. For Required Action B.1, redundant automatic initiation capability is lost if (a) two or more Function 1.a channels are inoperable and untripped such that both trip systems lose initiation capability, (b) two or more Function 2.a channels are inoperable and untripped such that both trip systems lose initiation capability, (c) two or more Function 1.b channels are inoperable and untripped such that both trip systems lose initiation capability, (d) two or more Function 2.b channels are inoperable and untripped such that both trip systems lose initiation capability, or (e), two or more Function 2.h channels are inoperable and untripped such that both trip systems lose initiation capability. 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, automatic initiation capability is lost if the combination of Function 3.a or Function 3.b channels that are inoperable and untripped result in the inability to energize the Function's trip relay; i.e., parallel pair logic channels are untrippable. 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.3 is not appropriate and the HPCI System 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 Aotion B.1), Required Aetion 8 .1 is oflly epplieeble ifl MODES 1, 2, end 3. In MODES 4 end 5, ti-le specitic initiation time ot the loi.v pr:essYr& ECCS is not assYm&d and the probability of a LOCA is lo't'ter. Thus, a total loss of iRitietioA eepebility for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Aotion B.3) is allowed during MODES 4 and 5 There ii no &in:iilar t!Jote provided for Required Action E3 2 since HPCI instrumentation is not rec:iYired in MODES 4 and 5; thus, a Note is not neoessary.

Notes are also provided (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 . Required Action B.1 (the Required Action for certain inoperable channels in the low pressure ECCS subsystems) is not applicable to Function 2.e, since this Function provides backup to administrative controls ensuring that operators do not divert LPCI flow from injecting into the core when needed. Thus, a total loss of Function Cooper B 3.3-109 02/22/16

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS (continued) must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 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.c, 1.e, 2.d, and 2.f, the affected portions are the associated low pressure ECCS pumps. As rmtee (Nete 1), Required Astion C.1 is only applisable in MODES 1, 2, and 3. In MODES 4 and a, the specitic initiation time oi the ECCS is not ass1.1med and the probability of a LOCA is lower. Thus, a total loss oi automatic initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (8s 811o"Oued by Required Action C.2) is allo"Oved durin~ MODES 4 and a.

IThe j1--->~Note-Z- states that Required Action C.1 is only applicable for Functions

1.c, 1.e, 2.c, 2.d, 2.f and 2.h. Required Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 9 and considered acceptable for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed by Required Action C.2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action C.1 , the Completion Time only begins upon discovery that the same feature in both subsystems (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.

Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 9) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events.

Cooper B 3.3-111 02/22/16

ECCS Instrumentation B 3.3 .5.1 BASES ACTIONS (continued)

Pressure Coolant Injection Pump Discharge Flow-Low 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.g (e.g., low pressure ECCS). Redundant automatic initiation capability is lost if (a) two Function 1.d channels are inoperable or (b) two Function 2.g 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 ),

Re~uireef Aetiel"I E.1 is el"lly applieable il"I MODES 1, 2, al"lef 3. 11"1 MODES 4 aAef 5, the specific iAitiatioA tifl'le of H~e EGGS is Aot assuA'leef aAef the probability of a LOCA is lo'floer. Thus, a total loss of initiation capability for Ther-~~~~~~~:

aRef 5. A Note is a so pro

~~ - ~~-~~~~ - ~~~

Note .Z 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 9 and considered acceptable for the 7 days allowed by Required Action E.2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

For Required Action E.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.

If the instrumentation that controls the pump minimum flow valve is inoperable, such that the valve will not automatically open, extended pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation, Cooper B3.3-113 02/22/16

RPV Water Inventory Control Instrumentation B 3.3.5.3 B 3.3 INSTRUMENTATION B 3.3.5.3 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation BASES BACKGROUND The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF.

If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.3 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 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 initiation or operation of the ECCS Cooper B3.3-132 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES BACKGROUND (continued) injection/spray subsystem required to be OPERABLE by LCO 3.5.2 and other functions that support automatic isolation of Residual Heat Removal subsystem and Reactor Water Cleanup system penetration flow path(s) on low RPV water level.

The RPV Water Inventory Control Instrumentation supports operation of core spray (CS) and low pressure coolant injection (LPCI). The equipment involved with each of these systems is described in the Bases for LCO 3.5.2.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 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.

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 judgement, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can be manually initiated to maintain adequate reactor vessel water level.

As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety.

Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36( c)(2)(ii).

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

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 Pressure - Low (Injection Permissive)

Cooper B 3.3-133 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES APPLICABLE SAFETY ANALYSES , LCO, and APPLICABILITY (continued)

Low reactor pressure signals are used as permissives for the low pressure ECCS injection/spray subsystems manual injection functions .

This function ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. While it is assured during MODES 4 and 5 that the reactor pressure will be below the ECCS maximum design pressure, the Reactor Pressure - Low signals are assumed to be OPERABLE and capable of permitting initiation of the ECCS.

The Reactor Pressure - Low signals are initiated from four pressure transmitters that sense the reactor dome pressure. The transmitters are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic.

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

The four channels of Reactor Pressure - Low Function are required to be OPERABLE in MODES 4 and 5 when ECCS manual initiation is required to be OPERABLE by LCO 3.5.2.

1.b, 2.b, Core Spray and Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass)

The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open . The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump.

One flow transmitter per ECCS subsystem is used to detect the associated subsystems' flow rates . The logic is arranged such that each transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded .

The LPCI minimum flow valves are time delayed such that the valves will not open for 3.5 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the Residual Heat Removal (RHR) shutdown cooling mode.

The Pump Discharge Flow - Low Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.

Cooper B3.3-134 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES APPLICABLE SAFETY ANALYSES , LCO, and APPLICABILITY (continued)

One channel of the Pump Discharge Flow - Low Function is required to be OPERABLE in MODES 4 and 5 when the associated Core Spray or LPCI pump is required to be OPERABLE by LCO 3.5.2 to ensure the pumps are capable of injecting into the Reactor Pressure Vessel when manually initiated.

RHR System Isolation 3.a - Reactor Vessel Water Level - Low, Level 3 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, Level 3 Function associated with RHR System isolation may be credited for automatic isolation of penetration flow paths associated with the RHR System .

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low, Level 3 Function are available , only two channels (both in the same trip system) are required to be OPERABLE.

The Reactor Vessel Water Level - Low, Level 3 Allowable Value was chosen to be the same as the Primary Containment Isolation Instrumentation Reactor Vessel Water Level - Low, Level 3 Allowable Value (LCO 3.3.6.1), since the capability to cool the fuel may be threatened .

The Reactor Vessel Water Level - Low, Level 3 Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME .

This Function isolates the Shutdown Cooling Isolation Valves.

Reactor Water Cleanup (RWCU) System Isolation 4.a - Reactor Vessel Water Level - Low Low, Level 2 The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close Cooper B 3.3-135 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation . The Reactor Vessel Water Level - Low Low, Level 2 Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.

Reactor Vessel Water Level - Low Low, Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low, Level 2 Function are available, only two channels (both in the same trip system) are required to be OPERABLE.

The Reactor Vessel Water Level - Low Low, Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level - Low Low, Level 2 Allowable Value (LCO 3.3.5.1), since the capability to cool the fuel may be threatened .

The Reactor Vessel Water Level - Low Low, Level 2 Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.

This Function isolates the Group 5 valves .

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.

Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.3-1 . The applicable Condition referenced in the Cooper B 3.3-136 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES ACTIONS (continued)

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 RHR System Isolation, Reactor Vessel Water Level - Low Level 3, and Reactor Water Cleanup System, Reactor Vessel Water Level - Low Low, Level 2 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.

Low reactor pressure signals are used as permissives for the low pressure ECCS injection/spay subsystems manual injection functions. If the permissive is inoperable, manual initiation of ECCS is prohibited.

Therefore, the permissive must be placed in the trip condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the permissive in the trip condition, manual initiation 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 trip.

If a Core Spray or Low Pressure Coolant Injection Pump Discharge Flow

- Low bypass function is inoperable, there is a risk that the associated low pressure ECCS pump could overheat when the pump is operating and the associated injection valve is not fully open. In this condition, the operator can take manual control of the pump and the injection valve to ensure the pump does not overheat.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities. The Completion Time is appropriate given the ability to manually ensure the pump does not overheat.

Cooper B 3.3-137 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES ACTIONS (continued)

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.

SURVEILLANCE REQUIREMENTS As noted in the beginning of the SRs, the SRs for each RPV Water Inventory Control instrument Function are found in the SRs column of Table 3.3.5.3-1 .

SR 3.3.5.3.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred . A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value . Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited ;

thus , it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST.

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

The 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.

Cooper B 3.3-138 xx/xx/xx

RPV Water Inventory Control Instrumentation B 3.3.5.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.5.3.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests .

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

The 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.

Cooper B 3.3-139 xx/xx/xx

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO , and APPLICABILITY (continued) switches that sense reactor dome pressure. Two channels of Reactor Pressure-High 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 both RHR shutdown cooling pump suction valves.

6 .b. Reactor Vessel Water Level-Low (Level 3)

Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water Level-Low (Level 3) 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 Level 3 supports actions to ensure that the RPV water level does not drop below fuel zone zero during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System .

Reactor Vessel Water Level-Low (Level 3) signals are initiated from four level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water Level-Low (Level

3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. -As neted (feetnete (b) te Table 3.3.6.1 1), enly ene trip system ef tt'le Reseter Vessel Wate1 Level*Louv (Level 3) Function is required to be OPERABLE in MODES 4 and 5, provieled the Rl-IR Shutdo*.vn Cooling System integrity is maintained . SystelTl integrity is maintained provided the piping is intaot and no ITlaintenance is being performed that has the potential for draining the reaetor vessel through the system.

The Reactor Vessel Water Level-Low (Level 3) Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low (Level 3) Allowable Value (LCO 3.3.1.1 ), since the capability to cool the fuel may be threatened.

Cooper B 3.3-149 11/22/16

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Reactor Vessel Water Level-Low (Level 3) Function is only required to be OPERABLE in MODES-3, 4, aRd 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2, another isolation (i.e., Reactor Pressure-High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path.

This Function isolates both RHR shutdown cooling pump suction valves and the inboard LPCI 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.

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 for Functions 2.a, 2.b, 5.d, and 6.b and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Functions other than Functions 2.a, 2.b, 5.d, and 6.b has been shown to be acceptable (Refs. 10 and 11) 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 8 .1 Bases). 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 A.1 .

Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. 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 isolation),

Condition C must be entered and its Required Action taken.

Cooper B 3.3-150 11125112 I

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS (continued) penetration flow path(s) accomplishes the safety function of the inoperable channels. Alternately, if it is not desired to isolate the affected penetration flow path(s) (e.g., as in the case where isolating the penetration flow path(s) could result in a reactor scram), Condition G must be entered and its Required Actions taken. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flow path(s).

G.1 and G.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or the Required Action of Condition Fis not met and the associated Completion Time has expired, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant 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 plant conditions from full power conditions in an orderly manner and without challenging plant systems.

H.1 and H.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated SLC subsystem(s) is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated SLC subsystems inoperable or isolating the RWCU System .

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System.

1.1 and 1.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed . However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdo*.vn Coolin9 System (i.e., l'rovide alternate deeay neat remo*.,.al capabilities so Cooper B 3.3-153 11125112 I

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS (continued) the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR ShutaowA GooliAg System is isolates.

SURVEILLANCE REQUIREMENTS As noted at the beginning of the SRs, the SRs for each 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 trip 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. 10 and 11) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path(s) when necessary.

SR 3.3.6 .1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels . It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value . Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION .

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.

Cooper B3 .3-154 11125112 I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES , LCO, and APPLICABIL TY function in harsh environments as defined by 10 CFR 50.49) are accounted for.

In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIVs and the SGT System are required.

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

1. Reactor Vessel Water Level-Low Low (Level 2)

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 Level-Low Low (Level 2) Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals.

The isolation and initiation systems on Reactor Vessel Water Level-Low Low (Level 2) support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

Reactor Vessel Water Level-Low Low (Level 2) signals are initiated from level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low (Level 2) Function are available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Low (Level 2) Allowable Value was chosen to be the same as the High Pressure Coolant Injection/ Reactor Core Isolation Cooling (HPCl/RCIC) Reactor Vessel Water Level Low Low (Level 2) Allowable Value (LCO 3.3.5.1 and LCO 3.3.5.2) since this could indicate that the capability to cool the fuel is being threatened).

The Reactor Vessel Water Level-Low Low (Level 2) 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 . IA additioA, tt:ie Cooper B 3.3-160 1112s112 I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABIL TY Function is also required to be OPERABLE dur i1 rg operatior rs uvitli a poteAtial for dreiAing tl'le reactor vessel (OPDRVs) because tl'le capability of isoletiAg poteRtiel seurees of leal<age must be provided to eRsure tl'lat offsite dose limits are Aot e>Eceeded if core damage occurs.

This function isolates the Group 6 valves listed in Reference 1.

2. Drvwell Pressure-High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. The Drywell Pressure-High Function associated with isolation is not assumed in any USAR accident or transient analyses, but will provide an isolation and initiation signal. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.

High drywell pressure signals are initiated from pressure switches that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.

The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LCO 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA).

The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.

3. Reactor Building Ventilation Exhaust Plenum Radiation-High High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from Cooper B 3.3-161 11125112 I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABIL TY the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident during refueling . When Reactor Building Exhaust Plenum Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the USAR safety analyses (Ref. 4).

The Reactor Building Exhaust Plenum Radiation-High signals are initiated from four radiation detectors that are located such that they can monitor the radioactivity of gas flowing through the reactor building exhaust plenum. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel in each trip system. Four channels of Reactor Building Ventilation Exhaust Plenum Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function .

The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.

The Reactor Building Ventilation Exhaust Plenum Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs, and movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded . Due to radioactive decay, this Function is 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 Cooper B 3.3-162 1112s112 I

CREF System Instrumentation B 3.3.7.1 BASES APPLICABLE SAFETY ANALYSIS, LCO, and APPLICABILITY (continued)

System. since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

Reactor Vessel Water Level-Low Low (Level 2) signals are initiated from level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low (Level 2) Function are available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude CREF System initiation.

The Reactor Vessel Water Level-Low Low (Level 2) Allowable Value was chosen to be the same as the Secondary Containment Isolation Allowable Value (LCO 3.3.6.2) to enable initiation of the CREF System at the earliest indication of a breach in the nuclear system process barrier, yet far enough below normal operational levels to avoid spurious initiation.

The Reactor Vessel Water Level-low Low (Level 2) Function is required to be OPERABLE in MODES 1, 2, and 3, and during operations with s potential for draining the rcaeter 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 the release of radioactive material to the environment is minimal. Therefore, this Function is not required in other MODES and specified conditions.

2. Drywell Pressure-High High drywelt pressure can indicate a break in the reactor coolant pressure boundary. A high drywell pressure signal could indicate a LOCA and will automatically initiate the CREF System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

Drywell Pressure-High signals are initiated from pressure switches that sense drywell pressure. Four channels of Drywell Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the initiation function . The Drywell Pressure-High Allowable Value was chosen to be the same as the ECCS Drywall Pressure-High Function Allowable Value (LCO 3.3.5.1).

The Drywall Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 to ensure that control room personnel are protected Cooper B 3.3-176 11125112 I

CREF System Instrumentation B 3.3.7.1 BASES APPLICABLE SAFETY ANALYSIS, LCO, and APPLICABILITY (continued) in the event of a LOCA. In MODES 4 and 5, the Drywell Pressure-High Function is not required since there is insufficient energy in the reactor to pressurize the drywell to the Drywell Pressure-High setpoint.

3. Reactor Building Ventilation Exhaust Plenum Radiation-High High radiation in the refueling floor area could be the result of a fuel handling accident. A refueling floor high radiation signal will automatically initiate the CREF System, since this radiation release could result in radiation exposure to control room personnel.

The Reactor Building Exhaust Plenum Radiation-High signals are initiated from radiation detectors that are located such that they can monitor the radioactivity of gas flowing through the reactor building exhaust plenum.

The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel in each trip system . Four channels of Reactor Building Ventilation Exhaust Plenum Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the CREF System initiation. The Allowable Value was chosen to promptly detect gross failure of the fuel cladd ing.

The Reactor Building Ventilation Exhaust Plenum Radiation-H igh Function is required to be OPERABLE in MODES 1, 2, and 3 and during movement of lately irradiated fuel assemblies in the secondary conta inment aAd ef3eratioFts witR a f30tcAtial fer draiAiAg ttic roaster vessel (OPDRVs), to ensure control room personnel are protected during e break resulting in significant releases of radioactive steam and g , fuel handling event, er vessel df'SiAdewA eveAt. During MODES 4 and 5, wheA tl"lese speeified eenditief'ls are f'let iA progress (e.g.,

OPDRVs), the probability of a pipe break resulting in significant releases of rad ioactive steam and gas or fuel damage is low; thus, the Function is not required . Due to radioactive decay, this Function is only required to initiate the CREF System during fuel handling accidents involving handling lately irradiated fuel (i.e., fuel that has occupied part of a critical

(

reactor core within the previous 7 days). During the movement of lately irradiated fuel, Reactor Building ventilation exhaust flow (provided by either a Reactor Building ventilation exhaust fan or SGT fan) is a required support function.

ACTIONS A Note has been provided to modify the ACTIONS related to CREF System instrumentation channels . Section 1.3, Completion Times, specifies that once a Cond ition has been entered, subsequent divisions, Cooper B 3.3-177 11125112 I

ECCS - Operating B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating

, RPVWATER INVENTORY CONTROL.

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.

The emergency condensate storage tanks (ECSTs) are capable of providing a source of water for the HPCI System.

On receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the ECSTs 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, if the ADS timed sequence is allowed to time out, the selected safety/relief valves (SRVs) would open, 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.

Cooper B 3.5-1 10/21/15

ECCS - Operating B 3.5.1 BASES APPL! CAB ILITY 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 s

dome pressure is 150 psig, ADS and HPCI are not require R OPERABLE because the low pressure ECCS su s can provi e sufficient flow below this pressure. EGGS re quirements for MODES 4 and 5 are specified in LCO 3.5.2, "eGGS Shutdown."

\_J.-R-P_V_W_at_e_r_ln_v_e_n-to_ry_C_o_n_tr_o_I- -

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.

If any one low pressure ECCS injection/spray subsystem is inoperable, or if one LPCI pump in both LPCI subsystems is inoperable, the inoperable subsystem(s) must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliab ility is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function . The 7 day Completion Time is consistent with the recommendations provided in a reliability study (Ref. 11) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service . The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed Completion Times .

B.1 and 8 .2 If the inoperable low pressure ECCS subsystem cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Cooper B 3.5-6 09/18/09

RPV Water Inventory Control B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control 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 active irradiated fuel at all times to prevent such elevated cladding temperatures .

APPLICABLE SAFETY ANALYSES 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 to the environment should an unexpected draining event occur.

A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses , and ductile piping systems. Instead, an event is considered in which single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e .g. ,

seismic event, loss of normal power, single human error). It is assumed ,

based on engineering judgement, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level.

As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety.

Therefore , RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36( c)(2)(ii).

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.

Cooper B 3.5-18 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES LCO (continued)

The Limiting Condition for Operation (LCO) requires the DRAIN TIME of RPV water inventory to the TAF to be~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A DRAIN TIME of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate 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.

One low pressure ECCS injection/spray subsystems is required to be OPERABLE and capable of being manually started to provide defense-in-depth should an unexpected draining event occur. A low pressure ECCS injection/spray subsystem consists of either one Core Spray (CS) subsystem or one Low Pressure Coolant Injection (LPCI) subsystem .

Each CS subsystem consists of one motor driven pump, piping , and valves to transfer water from the suppression pool to the RPV. Each LPCI subsystem consists of one motor driven pump , piping , and valves to transfer water from the suppression pool to the RPV. In MODES 4 and 5, the RHR System cross tie shutoff valve is not required to be closed .

The LCO is modified by a Note which allows a required 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 is not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is real igned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Because of the restrictions on DRAIN TIME, sufficient time will be available following an unexpected draining event to manually align and initiate LPCI subsystem operation to maintain RPV water inventory prior to the RPV water level reaching the TAF .

APPLICABILITY RPV water inventory control is required in MODES 4 and 5.

Requirements on water inventory control in other MODES are contained in LCOs in Section 3.3, Instrumentation, and other LCOs in Section 3.5, ECCS, RCIC, and RPV Water Inventory Control. RPV water inventory control is required to protect Safety Limit 2.1.1.3 which is applicable whenever irradiated fuel is in the reactor vessel.

Cooper B 3.5-19 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES ACTIONS A.1 and B.1 If the required low pressure ECCS injection/spray subsystem is inoperable, it must be restored to OPERABLE status with in 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 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 considers the LCO controls on DRAIN TIME and the low probability of an unexpected draining event that would result in loss of RPV water inventory.

If the inoperable ECCS injection/spray subsystem is not restored to OPERABLE status with in the required Completion Time , action must be initiated immediately to establish a method of water injection capable of operating without offsite electrical power. The method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The method of water injection may be manually initiated and may consist of one or more systems or subsystems, and must be able to access water inventory capable of maintaining the RPV water level above the TAF for 2': 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

If recirculation of injected water would occur, it may be credited in determining the necessary water volume .

C.1. C.2. and C.3 With the DRAIN TIME less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months but greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , compensatory measures should be taken to ensure the abil ity 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 fue l cladding and release of radioactive material. Add itional 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 Cooper B 3.5-20 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES ACTIONS (continued) secondary containment with respect to the environment. Verification that the secondary containment boundary can be established must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

Secondary containment penetration flow paths form a part of the secondary containment boundary. Required Action C.2 requires verification of the capability to isolate each secondary containment penetration flow path in less than the DRAIN TIME. The required verification confirms actions to isolate the secondary containment penetration flow paths are preplanned and necessary materials are available. Power operated valves are not required to receive automatic isolation signals if they can be closed manually within the required time.

Verification that the secondary containment penetration flow paths can be isolated must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases . Required Action C.3 requires verification of the capability to place one SGT subsystem in operation in less than the DRAIN TIME. The required verification confirms actions to place a SGT subsystem in operation are preplanned and necessary materials are available. Verification that a SGT subsystem can be placed in operation must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

D.1 , D.2, D.3, and D.4 With the DRAIN TIME less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , mitigating actions are implemented in case an unexpected draining event should occur. Note that if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action E.1 is also applicable.

Required Action D.1 requires immediate action to establish an additional method of water injection augmenting the ECCS injection/spray subsystem required by the LCO. The additional method of water injection includes the necessary instrumentation and controls , water sources, and pumps and valves needed to add water to the RPV or refueling cavity 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 Cooper B 3.5-21 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES ACTIONS (continued) offsite electrical power. The additional method of water injection may be manually initiated and may consist of one or more systems or subsystems. The additional method of water injection must be able to 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 a potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

The secondary containment provides a control volume in which fission products can be contained, diluted, and processed prior to release to the environment. Required Action D.2 requires that actions be immediately initiated to establish the secondary containment boundary. With the secondary containment boundary established, one SGT subsystem is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

The secondary containment penetrations form a part of the secondary containment boundary. Required Action D.3 requires that actions be immediately initiated to verify that each secondary containment penetration flow path is isolated or to verify that it can be manually isolated from the control room.

One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases . Required Action D.4 requires that actions be immediately initiated to verify that at least one SGT subsystem is capable of being placed in operation . The required verification is an administrative activity and does not required manipulation or testing of equipment.

If the Required Actions and associated Completion times of Conditions C or Dare 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.

Cooper B 3.5-22 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES ACTIONS (continued)

Note that Required Actions 0 .1, 0 .2, 0.3, and 0.4 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 REQUIREMENTS SR 3.5.2.1 This Surveillance verifies that the DRAIN TIME of RPV water inventory to the TAF is 2: 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The period of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate draining of reactor coolant. Loss of RPV water inventory that would result in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1 .1.3 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 sealing of the blade, the exposed cross-sectional area of the RPV penetration flow path is used .

The definition of DRAIN TIME excludes from the calculation those penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed , or otherwise secured in the closed position, blank flanges , or other devices that prevent flow of reactor coolant through the penetration flow paths. A blank flange or other bolted device must be connected with a sufficient number of bolts to prevent draining in the event of an Operating Basis Earthquake. Normal or expected leakage from closed systems or past isolation devices is permitted. Determination that a system is intact and closed or isolated must consider the status of branch lines and ongoing plant maintenance and testing activities.

The Residual Heat Removal (RHR) Shutdown Cooling System is only considered an intact closed system when misalignment issues (Reference 6) have been precluded by functional valve interlocks or by isolation devices, such that redirection of RPV water out of an RHR subsystem is precluded . Further, RHR Shutdown Cooling System is only considered an intact closed system if its controls have not been Cooper B 3.5-23 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) transferred to Alternate Shutdown , which disables the interlocks and isolation signals .

The exclusion of penetration flow paths from the determination of DRAIN TIME must consider the potential effects of a single operator error or 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 .

Surve illance 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 3.5.2.2 The minimum water level of 12 ft 7 inches required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the CS subsystem or LPCI subsystem pump , recirculation volume , and vortex prevention .

With the suppression pool water level less than the required limit, the required ECCS injection/spray subsystem is inoperable .

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

SR 3.5.2.3 The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the required ECCS injection/spray subsystems full of water ensures that the ECCS subsystem will perform properly. This may also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points .

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

Cooper B 3.5-24 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.4 Verifying the correct alignment for manual , power operated , and automatic valves in the required ECCS subsystem flow path provides assurance that the proper flow paths will be available 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 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.

SR 3.5.2.5 Verifying that the required ECCS injection/spray subsystem can be manually started and operate for at least 10 minutes demonstrates that the subsystem is available to mitigate a draining event. Testing the ECCS injection/spray subsystem through the recirculation line is necessary to avoid overfilling the refueling cavity. The minimum operating time of 10 minutes was based on engineering 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 (e.g., RHR, RWCU) actuates to the isolation position on an actual or simulate RPV water level isolation signal is required to prevent RPV water inventory from dropping below the TAF should an unexpected draining event occur.

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

Cooper B 3.5-25 xx/xx/xx

RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.7 The required ECCS injection/spray subsystem shall be capable of being manually operated from the Control Room . This Surveillance verifies that the required CS or LPCI subsystem (including the associated pump and valve(s)) can be manually operated to provide additional RPV water inventory, if needed.

The 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.

6. General Electric Service Information Letter No. 388, "RHR Valve Misalignment During Shutdown Cooling Operation for BWR 3/4/5/6," February 1983.

Cooper B 3.5-26 xx/xx/xx

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS {ECCS) AND REAC TOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5 .3 RCIC System , RPV WATER INVENTORY CONTROL, 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 lo ss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level. Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference 1 are satisfied.

The RCIC System (Ref . 2) consists of a steam driven turbine pump uni t , piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the sucti on source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Sucti on piping is provided from the emergency condensate storage tanks (ECSTs) and the suppression pool. Pump suction is normally aligned to the ECSTs to mini mize injection of suppression pool water into the RPV . However, if the ECST water supply is low , an automatic transfer to the suppression pool water source ensures a wate r 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 Sy stem is designed to provide core cooling for a wide range of steam inlet pressures, 150 to 1120 psia. Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water to the ECST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

(continued)

- -- -- - --- -*-**Ho**-*-*---****--*----***-----**--**-------- - - -** - - - - - - --- -----*****-*---**------***- *** - - - -***- *--

Cooper B3 . 5- ~ Rev*isiel'l 0

RCIC System B 3.5.3 BASES BACKGROUND (continued)

The RCIC pump is provided with a minimum flow bypass line, which discharges to the suppression pool. The valve in this line automatically opens to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge piping is kept full of water. The RCIC System is normally aligned to the ECSTs.

The RCIC discharge line is kept full of water using a "keep fill" system (Pressure Maintenance System).

APPLICABLE SAFETY ANALYSES The function of the RCIC System is to respond to transient events by providing makeup coolant to the reactor. The RCIC System is neither an ECCS nor 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, however, the system satisfies Criterion 4 of 10 CFR 50.36 (c)(2)(ii) (Ref. 3).

LCO The OPERABILITY of the RCIC System provides adequate core cooling such that actuation of any of the low pressure ECCS subsystems is not required in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC System has sufficient capacity for maintaining RPV inventory during an isolation event.

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 :S 150 psig, the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV. In MODES 4 and 5, RCIC is not required to be OPERABLE since RPV water inventory control is required by LCO 3.5.2, "RPV Water Inventory Control."

Cooper B 3.5-28 xx/xx/xx

PC I Vs B 3.6.1.3 BASES LCO MSIVs must meet additional leakage rate requirements. Other (continued) PCIV leakage rates are addressed by LCO 3.6.1.1, "PrimarJ 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 OBA 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, -mG£t. PCIVs are not required to be OPERABLE and the primary containment purge and vent valves when the are not required to be normally closed in MODES 4 and 5.

Certain valves, however, are required to be OPERABLE -re-..c.---

~reveAt iAa~vcrtcAt reacter vessel ~raiAclewA . T~cse valves are these whese associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "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)

Cooper B 3.6-18 Revision 0

PCIVs B 3.6.1.3 BASES ACTIONS (continued)

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 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 with in 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 afld F.2 If any Re(1uired Action ar 1d associated Completion Tin 1e ea1111ot be 1net for PGIV(s) requires te be OPERABLE duriAg MODE 4 er 5, ttie Uflit ffiUSt be plaseel in a soAelition in whish the LGO Elees Rot apply. l\otion A'IUSt be il'l'lA'lediately initiated to suspenEI eperatiofls with a potefltial fer draiAing the roaster vessel (OPDRVs) to l'l'linirnize the probability of a vessel elraindown and subsequent potential for fission produGt release . l\etions A'lust oontinue until OPDRVs are suspended aAd valve(s) arc rcstorcEI to OPERABLE status. If suspcFldirig ari OPDRV w*ould result in elosil"lg the residual heat reR1oval (Rl-lR) shutdo*Nn oooling isolation *1alvcs, an alternative Required Aetien is provided te iA'IA'lcdiately initiate aetioA to restore the 'O'Blve(s) to OPERABLE status. Tl 1is allows RI IR shutdovvn cooling te remain in service while aotions are being takeA to restore the valve.

SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR ensures that the 24 inch primary containment purge and vent valves are closed as required or, if open, open for an allowable reason. If a purge or vent valve is open in violation of this SR, the valve is considered inoperable. The SR is modified by Note 1 stating that the SR is not required to be met when the purge and vent valves are open for the stated reasons. Note 1 states that these valves may be opened in one supply line and one exhaust line for inerting, de-inerting, pressure control ,

ALARA or air quality considerations for personnel entry , or Surveillances that require the valves to be open. Note 2 modifies the SR by requiring both Standby Gas Treatment (SGT) subsystems OPERABLE and only one SGT subsystem operating when these purge and vent valves are open in accordance with Note 1.

Cooper B 3.6-23 November 28, 2001

Suppression Pool Water Level B 3.6.2.2 BASES APPLICABLE SAFETY ANALYSES (continued)

Suppression pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36 (c)(2){ii) (Ref. 2).

LCO A limit that suppression pool water level be ;;::: 12 ft 7 inches and s 12 ft 11 inches is required to ensure that the primary containment conditions assumed for the safety analyses are met. These limits equate to narrow range level instrument readings of-2" and +2", respectively. 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 OBA 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 requirement for maintaining suppression pool water level within limits in MODE 4 or 5 is addressed in LCO 3.5.2, "ECCS*Shutdow ~ RPV Water Inventory Control ACTIONS 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 drywell vents are covered, HPCI and RCIC turbine exhausts are covered ,

and SRV 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 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.

B.1 and B.2 If suppression pool water level cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Cooper B 3.6-61 02122116 I

Secondary Containment B 3.6.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)

Secondary containment satisfies Criterion 3 of 10 CFR 50.36( c)(2)(ii)

(Ref. 3).

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, following secondary containment isolation can be 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 tho reactor vessel (OPD~Vs), or d1::1rin9 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 critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

ACTIONS If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

Cooper B 3.6-71 02122116 I

Secondary Containment B 3.6.4.1 BASES ACTIONS (continued)

B.1 and 8.2 If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 BAd C.2 this case Movement of recently irradiated fuel assemblies in the second containment efld OPDR\'s can be postulated to cause sig * *cant fission product release to the secondary containment. In , the secondary containment is the only barrier to release of fission products to the environment. Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, eetioA fl"lust be ifl"lfl"ledietely iAitieted to suspeAd OPDRVs to fl"liAimize the probability of a

1essel draiF1dowA eAd subsequeAt poteAtiel for fissioA produet release.

AetieAs fl"lust eeAtiflue uAtil OPDRVs ere sus~eAded .

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.

Cooper B 3.6-72 02122116 I

SC IVs B 3.6.4.2 BASES APPLICABLE SAFETY ANALYSES (continued)

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment following secondary containment isolation so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3of10 CFR 50.36(cX2)(ii) (Ref. 5).

LCO SC IVs 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 are listed in Reference 6.

The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed or open in accordance with appropriate administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place. These passive isolation valves or devices are listed in Reference 6.

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 duriAg operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3. Due to radioactive decay, SCIVs are only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e. , fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Cooper B 3.6-76 02122116 I

SCI Vs B 3.6.4.2 BASES ACTIONS (continued) 0 .1, D.2, a11d 0 .3 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if apf'lieeble, eetions must be immediately initiated to sus,,end OPDRVs in order to minimize tne probability of a vessel di:aindewn and the s1:1eseq1:1ent potential fur fission product release. Actions must continue until OPDRVs ere suspended.

Required Action 0.1 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 irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE REQUIREMENTS SR 3.6.4.2.1 This SR verifies that each secondary containment manual isolation valve and blind flange that is 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 of the secondary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those SCIVs in secondary containment that are capable of being mispositioned are in the correct position .

Since these SCIVs are readily accessible to personnel during normal operation and verification of their position is relatively easy, the 31 day Frequency was chosen to provide added assurance that the SCIVs are in the correct positions. This SR does not apply to valves and blind flanges that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.

Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA Cooper B 3.6-79 02/22/16

SGT System 8 3.6.4.3 BASES LCO (continued)

When the required decay heat removal flow through the cross tie damper is not met, only ONE SGT subsystem may be considered 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 duriRg eperetieRs with a peteRtial fer eraiRiRg the reseter *tessel (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 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 subsystem and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Cooper B 3.6-84 02122116 I

SGT System B 3.6.4.3 BASES ACTIONS (continued) ~

C.1.. C.2.1. and C.2.2 During movement of recently irradiated fuel assemblies, in the secondary containment or dl:lring OPORVs, 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 have occurred , and that any other failure would be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position . Also, if applicable, actions ml:lst immediately be initiated to sl:lspend OPOR\ls in order to minimiile the probabi lity of a vessel dFaindown and sussequent potential foF fission preduet release. Aetiens must eentinue until OPDRVs are suspended.

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 irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

If both SGTS subsystems are inoperable in MODE 1, 2, or 3, the SGT system may not be capable of supporting the required radioactivity release control function. Therefore, actions are required to enter LCO 3.0.3 immediately.

Cooper B 3.6-85 02122116 I

SGT System B 3.6.4.3 BASES ACTIONS (continued)

E.1 and E.2 When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.

Also , if applioable, aotions must immediately be initiated to suspend OPDRVs in order to minimize tt'le J'robebility of 8 -vessel dreindotJvn and subsequent potential f-or fission preduet release. Aetierts must eoAtiAue until OPOR'Js are suspended.

Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE REQUIREMENTS SR 3.6.4.3.1 Operating each SGT subsystem, including each filter train fan, for<!:: 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for <!: 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

Cooper B 3.6-86 02122116 I

CREF System B 3.7.4 BASES APPLICABILITY (continued) required in MODE 4 or 5, except for ti 1e followi119 situatio11s u11der 1;ohieh sigriifieant radioactive releases eari be l'OStulated .

B. Duririg Ol'eratioris 'lluith a l'Oteritial for draining the reBet6r vessel (OPORVb), ar 1d

-&:- ~ring movement of lately irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the CREF System is only required to be OPERABLE during fuel handling involving handling lately irradiated fuel (i.e .* fuel that has occupied part of a critical reactor core within the previous 7 days).

ACTIONS When inoperable for reasons other than an inoperable CRE boundary, the inoperable CREF System must be restored to OPERABLE status within 7 days. With the unit in this condition, there is no other system to perform the CRE occupant protection function. The 7 day Completion Time is based on the low probability of a OBA occurring during this time period .

6.1 . 8.2, and B3 If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated dose of the licensing basis analyses of OBA consequences (allowed to be up to 5 rem whole body or its equivalent to any part of the body following a LOCA or 5 rem TEDE following a FHA),

or inadequate protection of CRE occupants from hazardous chemicals or smoke, the CRE boundary is inoperable. Protection of CRE occupants from hazardous chemicals or smoke is inadequate if they are unable to remain in the CRE and perform their duties. The CRE boundary must be restored to OPERABLE status within 90 days.

During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that in the event of a OBA, the mitigating actions will ensure that CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analyses of OBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke, i.e., the CRE occupants are able to remain in the CRE and perform their duties.

Cooper B 3.7-20 1112s112 I

CREF System B 3.7.4 BASES ACTIONS (continued)

These mitigating actions (i.e., actions that are taken to offset the consequences of the inoperable CRE boundary) should be preplanned for implementation upon entry into the condition, regardless of whether entry is intentional or unintentional. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion time is reasonable based on the low probability of a OBA occurring during this time period, and the use of mitigating actions. The 90 day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a OBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan, and possibly repair, and test most problems with the CRE boundary.

C.1 and C.2 In MODE 1, 2, or 3, if the inoperable CREF 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.

D.1, and D.2 ~

The Required Acti o ~ of Condition D 8fe modified by a Note indicating that LCO 3.0.3 does not apply. If moving lately irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of lately irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of lately irradiated fuel assemblies in the secondary containment or during OPDRVs, if the inoperable CREF System cannot be restored to OPERABLE status within the required Completion Time, or with the CREF System inoperable due to an inoperable CRE boundary, activities that present a potential for releasing radioactivity that might require isolation of the CRE must be immediately suspended. This places the unit in a condition that minimizes the accident risk.

If applicable, movement of lately irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of Cooper B 3.7-21 11125112 I

CREF System B 3.7.4 BASES ACTIONS (continued) these activities shall not preclude completion of movement of a component to a safe position. Also , if epplieeble, eetioAs A"IUst ee initiated iA"IA"leelietely to suspeAel OPDRVs te miAimize the !'}robability of a vessel draindown end the subsequent poteAtial for fission produot release.

Aetiel"ls must eel"ltinue until tne OPDRVs are suspeRdeel .

SURVEILLANCE REQUIREMENTS SR 3.7.4.1 This SR verifies that the CREF System in a standby mode starts on demand and continues to operate. The system should be checked periodically to ensure that it starts and functions properly. As the environmental and normal operating conditions of this system are not severe, testing the system once every month provides an adequate check on this system. Since the CREF System does not contain heaters, the system need only be operated for~ 15 minutes to demonstrate the function of the system. The 31 day Frequency is based on the known reliability of the equipment.

SR 3.7.4.2 This SR verifies that the required CREF testing is performed in accordance with the Ventilation Filter Tes ting Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test Frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.4.3 This SR verifies that on an actual or simulated initiation signal, the CREF System starts and operates. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.7.1, "Control Room Emergency Filter (CREF) System Instrumentation," overlaps this SR to provide complete testing of the safety function . The Frequency of 24 months is based on industry operating experience and is consistent with the typical refueling cycle .

Cooper B 3.7-22 11/25/12

AC Sources - Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources - Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources - Operating."

APPLICABLE SAFETY ANALYSES The OPERABILITY of the minimum AC sources during MODES 4 and 5, and during movement of 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.

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. Postulated worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences . These deviations from OBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1, 2, and 3, va rious deviations from the analysis assumptions and design requirements are allowed within the ACTIONS .

This allowance is in recognition that certain testing and maintenance activities must be conducted, provided an acceptable level of risk is not exceeded. During MODES 4 and 5, performance of a significant number of required testing and maintenance activities is also required. In MODES 4 and 5, the activities are generally planned and administratively Cooper B 3.8-23 02107113 I

AC Sources - Shutdown B 3.8.2 BASES APPLICABLE SAFETY ANALYSIS (continued) controlled . Relaxations from typical MODES 1, 2, and 3 LCO requirements are acceptable during shutdown MODES, based on:

a. The fact that time in an outage is limited . This is a risk prudent goal as well as a utility economic consideration.

b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operation MODE analyses, or both.

c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple systems.

d. Maintaining , to the extent practical, the ability to perform required functions (even if not meeting MODES 1, 2, and 3 OPERABILITY requirements) with systems assumed to function during an event.

In the event of an accident during shutdown, this LCO ensures the capability of supporting systems necessary for avoiding immediate difficulty, assuming either a loss of all offsite power or a loss of all onsite (diesel generator (DG)) power.

The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 1).

LCO One offsite circuit supplying the onsite Class 1E power distribution subsystem(s) of LCO 3.8.8, "Distribution Systems - Shutdown," ensures that all required loads are powered from offsite power. An OPERABLE DG, associated with a 4.16 kV critical 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 and reactor " &iiel draindown). Automatic initiation of the required DG during shutdown conditions is specified in LGO 3.3.5.1, EGGS IRstruffieF1tatioF1 , afld LCO 3.3.8.1, LOP Instrumentation.

The qualified offsite circuit must be capable of maintaining rated frequency and voltage while connected to its respective critical bus, and of accepting required loads during an accident. Qualified offsite circuits are those that are described in the USAR and are part of the licensing basis for the unit. The offsite circu it consists of incoming breaker and Cooper B 3.8-24 02107113 I

AC Sources - Shutdown B 3.8.2 BASES LCO (continued) disconnect to the startup or emergency station service transformer, associated startup or emergency station service transformer, and the respective circuit path including feeder breakers to all 4.16 kV critical buses required by LCO 3.8.8.

The required DG must be capable of starting, accelerating to rated speed and voltage, connecting to its respective critical bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 14 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 critical 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.

Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. The necessary portions of the Service Water System and Ultimate Heat Sink are also required to provide appropriate cooling to the required DGs.

It is acceptable during shutdown conditions, for a single offsite power circuit to supply both 4.16 kV critical buses. No fast transfer capability is required for offsite circuits to be considered OPERABLE.

APPLICABILITY The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment to provide assurance that: that provide core cooling are available

a. Systems ~~~;J...aael~~~oia~u::w.~~H+l~eUJ;)..a~

wailable for the irradiated fyel assemblies in tt:le core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident 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.

Cooper 8 3.8-25 02101113 I

AC Sources - Shutdown B 3.8.2 BASES ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since 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 irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving 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 or irradiated fuel assemblies would not be sufficient reason to require a reactor shutdown.

An offsite circuit is considered inoperable if it is not available to one required 4.16 kV critical bus. If two or more 4.16 kV critical buses are required per LCO 3.8.8, the remaining bus with offsite power available ~

may be capable of supporting sufficient required features to allew ~

continuation of CORE ALTERATIONS , fSel movement, aAd operatioAs witA a J3eteRtiel fer araiRiRQ tRe reseter *iiessel. By the allowance of the option to declare required features inoperable with no offsite power, appropriate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS . Required features remaining powered from a qualified offsite power circuit, even if that circuit is considered inoperable because it is not powering other required features, are not declared inoperable by this Required Action .

A.2.1. A.2.2. A.2.3. A.2.4, 8.1, 8.2. ~ and B.4= ~

With the required offsite circuit not available to all required 4.16 kV critical buses, 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 AL TERATIONS, ement of irradiated fuel assemblies in the secondary containment.-,:- ~e-a~~e&-~~~Je....l'eSt:flt..ffT-ffia~eR<effi.

areiRiRQ ef tRe reseter >tessel.

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.

Cooper B 3.8-26 02101113 I

AC Sources - Shutdown B 3.8 .2 BASES ACTION (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 4.16 kV critical 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 REQUIREMENTS SR 3.8.2.1 SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, and 3. SR 3.8.1.8 is not required to be met since only one offsite circuit is required to be OPERABLE. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

This SR is modified by a Note. The reason for the Note 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 deenergizing a required 4.16 kV critical bus or disconnecting a required offsite circuit during performance of SRs . With limited AC sources available, a single event could compromise both the required circuit and the 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 is requ ired to be OPERABLE.

Note 2 states SR 3.8.1.11 is considered to be met without the ECCS initiation signals OPERABLE when associated ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1 . This SR demonstrates the DG response to an ECCS signal in conjunction with a loss of power signal. When ECCS system(s ) are not required to be OPERABLE ~r LGO a.6.2, "EGGS Shukiewn ," the DG is not required to start in response to ECCS initiation signals. This is consistent with the Cooper B 3.8-27 02101113 I

AC Sources - Shutdown B 3.8.2 BASES SURVEILLANCE REQUIREMENTS (continued)

ECCS instrumentation requirements. However, the DG is still required to meet the other attributes of SR 3.8.1.11 when associated ECCS initiation signals are not required to be OPERABLE per Table a.a.e.1 1.

REFERENCES 1. 10 CFR 50.36(c)(2)(ii).

Cooper B 3.8-28 02101113 I

DC Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POW ER SYSTEMS B 3.8.5 DC Sources - Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources - Operati ng."

APPLICABLE SAFETY ANALYSES The initial conditions of Design Basis Accident and transient analyses in the USAR, Chapter XIV (Ref . 1) 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 auxil*iaries, and control and switching during all MODES of operation and during m ovem ent of 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 irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refue ling condition for extended periods ;

b. Sufficien t instrumentation and control capability is avaHable for monitori ng 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 l/066&1 or a fuel hand ling accident.

The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).

LCO The 125 V and 250 V DC electrical power subsystems , with each subsystem co nsisting of one batte ry, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus are required to be OPERABLE to support requi red DC distribution subsystems required OPERABLE by LCO 3.8.8, "Distribution Systems - Shutdown." This req uirement e nsures the Cooper B 3.8-47 0210711 3 I

DC Sources - Shutdown B 3.8.5 BASES LCO (continued) ava ilability 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 end inadvertent reactor vessel draiRdewn ).

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and dur.ing movement of irradiated fuel assemblies in the secondary containment provide assurance that:

~core cooling are available

a. Required features to provide t in*+<entory makeup ere available for the iFFadiated fuel assemblies in tfle eore iA case of an inad1io ertent draindown of tho reaGtor vessel ;

1

b. Required features needed to mitigate a fuel handling accident are avai,lable;

c. Required features n ecessary 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 cofd shutdown condition or refueling condiNon .

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 *is not applicable while in MODE 4 or 5. However, since irrad iated 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 irradiated fuel assemblies while in MOOE 4 or 5, LCO 3.0.3 wou ld not specify any action. If moving irradiated fue l assemblies while in MOOE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore , in either case, inability to suspend movement of irradiated fuel assemblies would not be sufficient reason to requ ire a reactor shutdown .

A.1 , A.2 .1, A.2 .2, A.2.3, and A. 2.4 If more than one DC distribution subsystem is requ ired according to LCO 3.8.8, the DC electrical power subsystems remaining OPERABLE , with one or more DC electrical power subsystems inoperable, may be capable of supporting sufficient required features to allow continuation of CORE Cooper 8 3.8-48 02101113 I

DC Sources - Shutdown B 3.8.5 BASES ACTIONS (continued) /8 AL TERA TIONS....fuel movement;-aRa operatioAs *Nitl'l a potential fer araining the reactor 't<ossel. By allowance of the option to declare required features inoperable with associated DC electrical power subsystems 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 AL TERATIONS.. ~ment of irradiated fuel assemblies in the secondary containment~ies result in inadvertent draining of the reactor ve~. [

l *r """'"

and _

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 untfl restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during wh ich the plant safety systems may be without sufficient power.

SURVEILLANCE REQUIREMENTS SR 3.8.5.1 SR 3.8.5.1 specifies applicability of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8. 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 DC sources from being discharged be1ow 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. USAR, Chapter XIV.

2. 10 CFR 50.36(c)(2)(ii).

Cooper B 3.8-49 02107113 I

Distribution Systems - Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems - Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution system is provided in the Bases for LCO 3.8.7, "Distribution Systems - Operating ."

APPLICABLE SAFETY ANALYSES The initia l conditions of Design Basis Accident and transient analyses in the USAR, Chapter XIV (Ref. 1) assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC and DC electrical power distrlbution system s are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the iuel, Reactor Coolant System, and containment design lim its are not exceeded.

The OPERABILITY of the AC and DC electrical power distribution system is consistent with tlie in itial assumptions of the accident ana lyses and the req uirements for the supported systems' OPERABtUTY.

The OPERAB1LITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refue'ling 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 aA inadvertent draindown of the \'essel or a fuel handling accident.

The AC and DC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).

LCO Various combinations of subsystems, equipment, and components are requ ired 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 Cooper B 3.8-63 0210711 3 I

Distribution Systems - Shutdown B 3.8.8 BASES LCO (continued) the portions of the electrical distribution system 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.

In addition, it is acceptable for required buses to be cross-tied during shutdown conditions, permitting a single source to supply multiple redundant buses, provided the source is capable of maintaining proper frequency (if required) and voltage.

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 aRd im.tdvefteRt reseter ivessel araiRde*i'ln ).

APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel that provide assemblies in the secondary containment provide assurance that:

core cooling are available a. Syste to pro*J4de .adequate coolant in~ntory makeup are available for tho irradiated fud iA U~e eor=e iA ease of an inadvertent drairnfoWfl of the reactor vessel.;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC and DC electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8.7.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since 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 irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving 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 irradiated fuel assemblies would not be sufficient reason to require a reactor shutdown.

Cooper B 3.8-64 02101113 I

Distribution Systems - Shutdown B 3.8.8 BASES ACTIONS (continued)

A.1 , A.2.1, A.2.2, A.2 .3, ~ 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 and s *

t requ ired features to allow continuation of CORE AL TERATl fuel movement, and operations with a potential for dFa in i!"I~ tne r:eaeto r ¥essel. By allowing the option to deciare 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 adm inistrative efforts. Therefore, the and r~~~~~o~r~s~u~ff~ic~ie:1n~tly conservative actions is made, (i.e., to suspend CORE ALTERA ... movement of irradia ted fue*I assemblies in the secondary containment, and any aetivities that eould 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 .i nitiate 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 Adiuns A.2.1 tti rough A. 2.4 do not adequately address the concerns relating to coolant circu lation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACT IONS wou ld not be entered . Therefore , Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable, which resul ts 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 requ ired distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.

Cooper B 3.8-65 02101113 I

lnservice Leak and Hydrostatic Testing Operation 83.10 .1 BASES BACKGROUND (continued)

Scram time testing required by SR 3.1.4.1 and SR 3.1.4.4 requires reactor steam dome pressure~ 800 psig. The hydrostatic and/or RCS leakage tests require pressure of approximately 1,000 psig.

Other testing (Excess Flow Check Valve testing for example) may be performed in conjunction with the allowances for inservice leak or hydrostatic tests and control rod scram time tests.

APPLICABLE Allowing the reactor to be considered in MOOE 4 when the reactor SAFETY ANALYSES 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 . 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 that could result in rvatively limit radia *an releases to the environme draining of the RPV any In the event of ary system leak, e reactor vessel would rapidly depressurizeJ allowing the low pressure oore oooling systeffis to operate . T capability of tho lmv PF066blr9 coolant injocti~ and corr I

~~~~~~~.~ requ ired in MODE 4 by LCO 3.5.2, ~ _RPV WIG .

~~*1--+Tftfflt:lffi'I~ ." would be more than adequate to keep the ~

makeup fleoded under this low decay heat load condition . Small system leaks RPV would be detected by leakage inspections before significant inventory water loss occurred . level above For the purposes of this test, the protection provided by normally required the TAF 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 .

Cooper B 3.10-2 11/06/06

v t e Letter Sequence Request
CAC:MG0138, (Approved, Closed)
Initiation Request Acceptance Administration Questions 24 January 2018 Answers Results Approval Other: ML17319A227, ML18171A161
MONTHYEAR2017-09-07 [Table View]

NLS2017048, Submittal of Application to Revise Technical Specifications to Adopt TSTF-542, Revision 2, Reactor Pressure Vessel Water Inventory Control

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Dear Sir or Madam:

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NLS2017048, Submittal of Application to Revise Technical Specifications to Adopt TSTF-542, Revision 2, Reactor Pressure Vessel Water Inventory Control

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Dear Sir or Madam:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

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