Text

Submittal of Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.
	ML17360A159
Person / Time
Site:	LaSalle
Issue date:	12/13/2017
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-17-127
	Download: ML17360A159 (229)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 Exelon Generation 630 65 7 2000 Office RS-17-127 10 CFR 50.90 December 13, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , D.C. 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (EGC) is submitting a request for an amendment to the Technical Specifications (TS) for LaSalle County Station (LSCS),

Units 1 and 2.

The proposed change replaces existing TS requirements related to operations with a potential for draining the reactor vessel (OPDRVs) with new requirements on reactor pressure vessel water inventory control (RPV WIC) to protect Safety Limit 2.1.1 .3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel (TAF). provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.

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

Approval of the proposed amendment is requested by December 13, 2018. Once approved , the amendment will be implemented prior to entry into Mode 4 following refueling activities during the LSCS, Unit 2 refuel outage in spring 2019 (i.e., L2R17), whM;h is currently scheduled to occur in February 2019.

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

December 13, 2017 U.S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing Is true and correct. Executed on the 13th day of December 2017.

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

2. Proposed Technical Specifications Changes (Mark-Up)

3. Revised Technical Specifications Pages

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

LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

Subject:

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

1.0 DESCRIPTION

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation 2.2 Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis 4.0 ENVIRONMENTAL EVALUATION

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

Exelon Generation Company, LLC (EGC) proposes a change to the LaSalle County Station (LSCS), Units 1 and 2 Technical Specifications (TS) requirements related to "operations with a potential fo r draining the reactor vessel" (OPDRVs) with new requi rements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation EGC has reviewed the safety evaluation provided to the Technical Specifications Task Force on December 20, 2016, as well as the information provided in TSiF-542. EGC has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the NRC are applicable to LSCS , Units 1 and 2 and justify this amendment for the incorporation of the changes to the LSCS TS.

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

1.1, Definitions 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation 3.3.6.1, Primary Containment Isolation lnstrumentatiotl 3.3.6.2, Secondary Containment Isolation Instrumentation 3.3.7.1, Control Room Area Filtration (CRAF) System Instrumentation 3.3.8.1, Loss of Power (LOP) Instrumentation 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring 3.5.1 , ECCS - Operating 3.5.2, ECCS - Shutdown 3.5.3, RCIC System 3.6.1 .3, Primary Containment Isolation Valves (PCIVs) 3.6.4.1, Secondary Containment 3.6.4.2, Secondary Containment Isolation Valves (SCIVs) 3.6.4.3, Standby Gas Treatment (SGT) System 3.7.4, Control Room Area Filtration (CRAF) System 3.7.5, Control Room Area Ventilation Air Conditioning (AC) System 3.8.2, AC Sources - Shutdown 3.8.5, DC Sources - Shutdown 3.8.8, Distribution Systems - Shutdown Page 1 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT 2.2 Variations EGC is proposing the following variations from the TS changes described in TSTF-542 or the applicable parts of the NRC's safety evaluation . These variations do not affect the applicability of TSTF-542 or the NRC's safety evaluation to the proposed license amendment.

In a few instances, the LSCS TS utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-542 was based . Specifically, the titles for the following LSCS , Units 1 and 2 TS vary from the STS discussed in TSTF-542 :

3.3.7.1, Control Room Area Filtration (CRAF) System Instrumentation 3.7.4, Control Room Area Filtration (CRAF) System 3.7.5, Control Room Area Ventilation Air Conditioning (AC) System These differences are administrative and do not affect the conclusion that TSTF-542 is applicable to the LSCS TS.

The LSCS TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement Frequencies for Limiting Conditions for Operation (LCOs) 3.3.5.2 and 3.5.2 are "In accordance with the Surveillance Frequency Control Program."

The LSCS TS differ from the STS on which TSTF-542 was based , but are encompassed in the TSTF-542 justification . Specifically, there are LSCS specific instrumentation functions that differ from the STS. Changes to these instrumentation functions are encompassed by the discussion in Section 3.3.4 of the TSTF-542 justification.

o LSCS TS Table 3.3.5.1-1 contains no function similar to STS Function 3.d (i.e., Condensate Storage Tank Level-Low) . This is a minor difference, due to the fact that the HPCS systems for LSCS , Units 1 and 2 are normally aligned to take suction from their unit's suppression pool , which provides the same function as the Condensate Storage Tank described in the STS (i.e., provides water source for the required HPCS system).

o In addition to the LPCI and LPCS subsystem injection permissive functions based on Reactor Steam Dome Pressure-Low, the LSCS low pressure ECCS subsystems require an additional injection permissive signal based on their associated Injection Line Pressure-Low (i.e., proposed Table 3.3.5.2-1 Functions 1.d and 2.c) . These permissive functions are utilized to protect the low pressure ECCS systems from pressures that exceed their design; therefore, their inclusion in the pr.ol!K)sed LCO 3.3.5.2, Table 3.3.5.2-1 is essentially the same as the justification for the inclusion of the Reactor Steam Dome Pressure-Low injection permissive functions.

Page 2 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

LSCS, Units 1 and 2 do not currently have the capabilit,y to perform Channel Checks for the following proposed Table 3.3.5.2-1 Functions:

1. Low Pressure Coolant Injection A (LPCI) and Low Pressure Core Spray (LPCS)

Subsystems:

a. Reactor Steam Dome Pressure-Low (Injection Permissive)

b. LPCS Pump Discharge Flow-Low (Bypass)

c. LPCI Pump A Discharge Flow-Low (Bypass)

d. LPCS and LPCI A Injection Line Pressure-Low (Injection Permissive)

2. LPCI B and LPCI C Subsystems:

a. Reactor Steam Dome Pressure-Low (Injection Permissive)

b. LPCI Pump Band LPCI Pump C Discharge Flow-Low (Bypass)

c. LPCI Band LPCI C Injection Line Pressure-Low (lrvection Permissive)

3. High Pressure Core Spray (HPCS) System:

a. HPCS Pump Discharge Pressure-High (Bypass)

b. HPCS System Flow Rate-Low (Bypass)

5. Reactor Water Cleanup (RWCU) System Isolation:

a. Reactor Vessel Water Level-Low Low, Level 2.

The current LSCS, Units 1 and 2 TS do not include Channel Checks for these functions; therefore, no Channel Check Surveillance Requirement (SR) was added for these functions.

LSCS LCO 3.3.8.1 , "Loss of Power (LOP) Instrumentation ," currently contains a footnote in Table 3.3.8.1-1 that is required to be modified along with the adoption of TSTF-542 as-proposed . Currently, Table 3.3.8.1-1 , Footnote a adds applicability for Functions 1.e and 2.e, "Degraded Voltage - Time Delay, LOCA." This footnote currently adds applicability for these Functions in Modes 4 and 5, when associated ECCS subsystem(s) are required to be operable by LCO 3.5.2, "ECCS-Shutdown." The purpose of this footnote is to ensure that the Degraded Voltage Time Delay, LOCA, Function is operable in Modes 4 and 5 when the associated ECCS subsystem is required to be operable for automatic initiation . EGC's justification for the proposed modification of Table 3.3.8.1-1, Footnote a is that following the adoption of TSTF-542, no ECCS subsystems will be required to start automatically in Modes 4 and 5; therefore, these Functions will no longer be required to be operable in Modes 4 and 5.

EGC proposes to delete a portion of the applicability for LSCS , Units 1 and 2 LCO 3.6.1 .3, "Primary Containment Isolation Valves (PCIVs) ," as shown in Figure 1 below, and LCO 3.6.1.3, Condition F and all of its associated Required Actions as shown in Figure 2 below. The Applicability for LCO 3.6.1.3 is: MODES 1, 2, and 3, and; When associated instrumentation is required to be OPERABLE per LCO 3.3. 6. 1, "Primary Containment Isolation Instrumentation. " These changes are justified since all OPDRV requirements are being deleted, and Mode 4 and 5 (i.e., the only non-Mode 1, 2, and 3 PCIV requirement in LCO 3.3.6.1) PCIV requirements have been relocated from LCOs 3.3.6.1 and 3.6.1.3 to the proposed LCOs 3.3.5.2 and 3.5.2. Thus, there are no longer any PCIVs required to be Page 3 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT operable by LCO 3.6.1.3 during OPDRVs, during Mode 4 or 5, or by LCO 3.3.6.1. These requirements are addressed by the proposed LCO 3.3.5.2 and 3.5.2 in their entirety.

Following the removal of OPDRV and relocation of Mode 4 and 5 requirements as discussed above, this portion of the LCO 3.6.1 .3 Applicability, and Condition F and associated Actions would never be applicable; therefore, are no longer necessary in LCO 3.6.1.3.

AP PLICAB ILI TY: MOD ES l, 2, and 3, When associated instrYmentation is reqYired to be QPERA8LE per LCO J.J.6,1, "Primary Containment Isolation Instrument~tion."

Figure 1: Proposed Variation That Deletes a Port of the LCO 3.6.1 .3 Applicability F"

Re El 1,.1 i re El Ac ti on a nEl .,...i;:__ n ,-*t-i-a-t.....e ~a....c-t-i....o++n-t-o 1+-----+-I... Immecliately associatecl Completion s1,.1spencl operations Time of Condition A, with a potential for B, C, or D not met for Graining the reactor PGl'H s) reEJHi reel to be vessel (OPDRl/s).

OPERABLE d1,.1ring MODE 4

~ M F.2 Initiate ac t ion to Immediately restore valve(s) to OPERABLE statHS .

Figure 2: Proposed Variation That Deletes LCO 3.6.1.3 Condition F

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

EGC proposes to revise Table 3.3.5.2-1 , "RPV Water Inventory Control Instrumentation," as described in TSTF-542 to reflect the LSCS, Units 1 and 2 design. Specifically, Function 1, "Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS)

Subsystems," Function 1.d, "Manual Initiation," Function 2, "LPCI Band LPCI C Subsystems," Function 2.c, "Manual Initiation," and Function 3, High Pressure Core Spray (HPCS) System, Function 3.a, "Reactor Vessel Water Level - High, Level 8," and Function 3.e, "Manual initiation," that appear in the BWR/6 Technical Specifications in TSTF-542 are not included in the LSCS Technical Specifications as proposed. This corrects an issue in TSTF-542 associated with the BWR/5 and BWR/6 emergency core cooling system (ECCS) instrumentation requirements.

Page 4 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT The purpose of the manual initiation functions is to allow manual actuation of the ECCS subsystems required by TS 3.5.2 to mitigate a drain ing evQJlt. Licensed operators in the Ma in Control Room have the capability to manually start ttle LPCI , LPCS , and HPCS pumps and to manually align valves to add water inventory, if needed. This can be accomplished without the "Manual Initiation" functions, and the "Reactor Vessel Water Level -High, Level 8" function associated with HPCS. If the water level is above Level 8, and HPCS is the required ECCS subsystem , the Level 8 function can be intentionally defeated to allow the HPCS injection valve to be opened , if needed to control inventory. All actions can be performed from the Main Control Room and can be accomplished well within the one-hour minimum drain time limit specified in TS 3.5.2, Condition E.

The Reactor Vessel Water Level High , Level 8 signal (i.e. , TSTF-542 , Table 3.3.5.2-1 ,

Function 3.a) prevents overfilling of the reactor vessel into the main steam lines by closing the HPCS injection valves when the water level is above the Level 8 setpoint. Therefore, if HPCS is the required ECCS subsystem and the water level is above Level 8, using the "Manual Initiation" Function 3.e will not result in inventory injection into the reactor vessel until the water level drops below the Level 8 setpoint. If the Level 8 function is retained in Table 3.3.5.2-1 , the function would need to be rendered inoperable in order to inject water when the water level is above the Level 8 setpoint.

Consequently, Table 3.3.5.2-1 , Functions 1.d, 2.c, 3.a, and 3.e, and TS 3.3.5.2, Condition E and associated Required Actions E.1, and E.2 as described in TSTF 542 are not needed to actuate the LPCI , LPCS , and HPCS subsystem components to mitigate a draining event, and are not included in the proposed Table 3.3.5.2-1 for LSCS . Since EGC proposes to not include the Manual Initiation logic functions in Table 3.3.5.2-1 , the need for including a Surveillance Requirement to perform a Logic System Function Test for any Table 3.3.5.2 function is eliminated . Therefore, SR 3.3.5.2.3 as described in TSTF-542 is not included in the proposed LCO 3.3.5.2 for LSCS . The remain ing functions and Cond itions have been renumbered accordingly.

Page 5 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT To address the changes associated with this variation , EGC also proposes that Surveillance Requirement (SR) SR 3.5.2.8 be modified to verify that the LSCS , Units 1 and 2 required ECCS injection/spray subsystem can be manually operated from the Main Control Room in accordance with the Surveillance Frequency Control Program as shown in Figure 3 below.

This will ensure that the requ ired ECCS injection/spray subsystem is Operable and can be manually aligned to provide RPV inventory makeup, if required to do so, without delay.

SR 3.5.2 . 8 - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Vessel injection/s pray may be excluded.

Verify the required ECCS injection/spray In accordance subsystem can be manually operated. with the Survei l lance Frequency Contro l Program Figure 3: Proposed SR 3.5.2.8

EGC proposes to modify LCO 3.8.2, "AC Sources-Shutdown," SR 3.8.2.1. The purpose for SR 3.8.2.1 is to define the LCO 3.8.1, "AC Sources-Operating ," SRs that are necessary for ensuring the operability of the AC sources in Modes or Conditions other than Modes 1, 2, and 3. SR 3.8.2.1 currently contains two notes. The purpose of the first note is to preclude rendering the require diesel generator inoperable for testing , and disconnecting a required offsite circuit during the performance of the listed SRs.

According to the TS Bases, it is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the DG and offsite circuit are required to be operable. Note 2 relaxes requirements for performing SR 3.8.1.12 and SR 3.8.1.19 when ECCS subsystems are not required to be operable in accordance with LCO 3.5.2, "ECCS-Shutdown ." The intent of Note 2 is to be consistent with other ECCS instrumentation requirements that are not required when the associated ECCS subsystem will not receive an initiation signal. Since, following the adoption of TSTF-542 as proposed, no ECCS initiation signal will be provided in Modes 4 and 5, EGC proposes to delete reference to LCO 3.5.2 from SR 3.8.2.1. This revision will continue to ensure that the required AC sources are adequately tested without unnecessarily rendering them inoperable during shutdown periods when the available AC sources are limited .

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Exelon Generation Company, LLC (EGC) requests adoption of Technical Specifications Task Force Traveler (TSTF)-542 "Reactor Pressure Vessel Water Inventory Control ," which is an approved change to the Standard Technical Specifications (STS) , into the LaSalle County Page 6 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT Station, Units 1 and 2 Technical Specifications (TS) . The proposed amendment replaces the existing requirements in the TS related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

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

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

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1 .3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e. , refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated.

RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

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

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

The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as Page 7 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a drain ing 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 The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC . The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting d~in 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 .

Page 8 of 9

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT Therefore , the proposed change does not involve a significant reduction in a margin of safety.

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

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

Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposE¥l change.

Page 9 of 9

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

TOC Page i 3.3.5.2-3 3.6.1 .3-1 TOC Page ii 3.3.5.2-4 3.6.1.3-5 1.1-2 3.3.5.3-1 3.6.4.1-1 1.1-3 3.3.5.3-2 3.6.4.1-2 1.1-4 3.3.5.3-3 3.6.4.2-1 1.1-5 3.3.5.3-4 3.6.4.2-3 1.1-6 3.3.6.1-9 3.6.4.3-1 1.1-7 3.3.6.2-4 3.6.4.3-2 1.1-8 3.3.7.1-1 3.6.4.3-3 1.1-9 3.3.8.1-3 3.7.4-1 3.3.5.1-2 3.3.8.2-1 3.7.4-2 3.3.5.1-3 3.3.8.2-3 3.7.4-3 3.3.5.1-4 3.5.1-1 3.7.5-1 3.3.5.1 -9 3.5.2-1 3.7.5-2 3.3.5.1-10 3.5.2-2 3.7.5-3 3.3.5.1-11 3.5.2-3 3.8.2-3 3.3.5.1-12 3.5.2-4 3.8.2-4 3.3.5.2-1 3.5.2-5 3.8.5-3 3.3.5.2-2 3.5.3-1 3.8.8-2

TABLE OF CONTENTS

1. 0 USE AND APPLICATION

1. 1 Definitions ........ . ............ . .. . .. ....... . ..... . ..... 1 . 1-1

1. 2 Logical Connectors . . . . ... .. .... . . . .. .. ....... .. .... .. . . .. 1.2-1

1. 3 Completion Times ... . . . .... . . . ....... .. ....... .. .. ..... .. . 1.3-1

1. 4 Frequency .. .. .. . . .............. .. .... . ... . ... .. ... ....... 1.4-1 2.0 SAFETY LIMITS CSLs) 2.1 SLs ....... . . ... . ........ . .... . . . . .................... . ... 2 . 0-1 2.2 SL Violations .. .... .... ... .. . . . . ... . ........ . ....... ..... 2 . 0-1 3.0 LIMITI NG COND IT ION FOR OPERATION (LCO) APPLICAB IL ITY . ... ... . 3.0-1 3.0 SURVEILLANCE REQUIREMENT CSR) APPLICABILITY . . . .... .. .. .. . ... 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS 3 . 1.1 SHUTDOWN MARGIN (SDM) ................ . .. . . ........ . ...... 3 . 1 . 1-1 3 .1. 2 Reactivity Anomalies .... . .. . . .. . .... . ......... . ..... .. . .. 3 . 1.2-1 3 . 1. 3 Control Rod OPERABILITY . . .... . . . . ... . . .. ...... . ... . ...... 3 . 1 . 3-1 3 . 1.4 Control Rod Scram Times ............... .... .... ........... 3.1.4-1 3 .1. 5 Control Rod Scram Accumulators .... . . .... . . ... ....... .. .. . 3 . 1 . 5-1 3 . 1. 6 Rod Pattern Control ............ . . . .... . . .. . ...... . ..... . . 3 . 1 . 6-1 3 . 1. 7 Standby Liquid Control CSLC) System . . .. . ....... ..... .. .. . 3 . 1 . 7-1 3 . 1. 8 Scra m Disc harge Vo lume (SD V) Vent and Drain Val ves ....... 3 . 1. 8-1 3.2 POWER DISTRIBUTION LIMITS 3 . 2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ..... . 3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ..... . .. .... .. ...... . . 3 . 2 . 2-1 3.2.3 LINEAR HEAT GENERATION RATE CLHGR) . .. .. . . ...... . .. . .. .. . 3 . 2 . 3-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation .... . . .... 3 . 3.1 . 1-1 3 . 3 . 1.2 Source Range Monitor CSRM) Instrumentation ..... .. . ... . ... 3.3.1.2-1 3 . 3 . 1.3 Oscillation Power Range Monitor (OPRM) Instrumentation ... 3 . 3 . 1.3-1

3. 3 . 2 .1 Control Rod Bl ock Instrumentation .. . .. ....... .. .. . ....... 3 . 3 . 2.1-1 3.3.2 . 2 Feedwater System an d Main Turbine Hig h Water Level Trip Instrumentation .. ... .. . . .. . . . . ............ ........ 3.3.2 . 2-1

3. 3. 3 . 1 Post Accident Monitoring (PAM) Instrumentation . .. ... . . ... 3 . 3.3.1-1 3.3.3.2 Remote Shutdown Monitoring System ..... . . .. . .... ... . ...... 3.3.3.2-1 3.3.4.1 End of Cycle Recirculation Pump Trip CEOC-RPT)

Instrument at i on ..... . . . .. . .. . ....... .............. ..... 3 . 3 . 4 . 1- 1 3.3.4. 2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ............ ..... . . 3 . 3 . 4 . 2-1 3.3. 5.1 Emergency Core Cooling System (ECCS) Instrumentation .... . 3.3.5.1-1

3. 3.5.2 Reactor Pressure Vessel (RPVJ Water Inventory Control Instrumentation . .. ...... . .... . . . ... . . . ................. 3. 3. 5. 2- 1 3 . 3 . 5. ~3 Reactor Core Isolation Cooling CRCIC) System Instrumentation ............ .. . ............... ... . .... 3.3 . 5. 3~-l 3.3.6.1 Primary Containment Isolation Instrumentation ..... ....... 3 . 3 . 6 . 1-1 (continued)

LaSalle 1 and 2 Amendment No.

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued )

3.3.6.2 Secondary Containment Isolation Instrumentation ...... .... 3.3 . 6.2-1 3 . 3.7 . 1 Control Room Area Filtration (CRAF) System Instrumentation .. ...... .. ........... .. . .... . . .......... 3.3.7.1 - 1 3.3.8.1 Loss of Power (LOP) Instrumentation .... . . .... ..... . .. ... . 3.3.8.1-1 3.3.8. 2 Reactor Protection System (RPS) Electric Power Monitoring .. . ......... . .... ......... ... ................ 3.3.8 . 2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3 . 4.1 Recirculation Loops Operating ....... .................. ... 3 . 4.1-1 3 .4. 2 Flow Contro l Valves (FCVs) . . ...... ..... . ................. 3 .4. 2-1 3 .4 . 3 Jet Pumps .. .. .... . .. ...... . .... ....... .... .. . .. .......... 3.4. 3-1 3.4.4 Safety/Relief Valves (S/RVs) ............................. 3.4.4 -1 3.4.5 RCS Operational LEAKAGE .................................. 3 . 4. 5- 1 3 . 4. 6 RCS Pressure Isolation Valve (PIV) Leakage ........ . .. .... 3 . 4. 6-1 3.4.7 RCS Leakage Detection Instrumentation .................... 3.4. 7- 1 3 . 4 .8 RCS Specif ic Activity ........................... ........ . 3 .4.8 -1 3 . 4.9 Residual Heat Removal (RHR) Shu tdown Cooling System-Hot Shutdown .............. ......... ... .... . .. . . . 3 . 4 . 9- 1 3 .4. 10 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown ......... ......... .. ..... . ......... 3 .4.10 -1 3 . 4 .11 RCS Pressure and Te mp erature (P/T) Limit s .............. .. 3. 4.11 -1 3 .4. 12 Reactor Steam Dome Pressure ............ . .. ... ........ .... 3 . 4.1 2-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) , REACTOR PRESSURE VESSEL (RPV)

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

SYSTEM 3.5.1 EC CS-0 per at i ng ........ . ...... ...... ...................... 3 . 5 . 1 - 1 3.5.2 EGGS Shutdm,*nRPV Water Inventory Control .... ...... . ... .. . 3 . 5. 2-1 3.5.3 RC IC-System .... .... . .. ............. .. .. . ................. 3. 5. 3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Pri mary Containment ........ . ................. ............ 3.6.1 . 1-l 3.6.1.2 Primary Conta inm ent Air Lock .. ... .. . ............. ........ 3.6.1.2-1

3. 6. 1.3 Primary Containment Isolation Valves (PCIVs) ........ . .... 3 . 6.1 . 3-1 3.6.1.4 Drywell and Suppression Chamber Pressure ......... . . ...... 3 . 6.1.4-1 3 . 6.1.5 Drywell Air Temperature ...... . . ..... . ...... . ............. 3 . 6 . 1.5-1 3 . 6 . 1.6 Suppression Chamber-to-Drywell Vacuum Breakers ........ ... 3 . 6.1 . 6-1 3 . 6. 2.1 Suppression Pool Average Temperature . .... .. ........ . ... .. 3 . 6.2.1-1 3 . 6.2.2 Suppression Pool Water Level ........ .. ..... .............. 3.6 . 2. 2-1 3 . 6.2 . 3 Residual Heat Removal (RHR) Suppression Pool Cooling ..... 3.6.2 . 3-1 3 . 6 . 2. 4 Residual Heat Removal (RHR) Suppression Pool Spray .... .. . 3.6.2 . 4-1 3.6.3. 1 Primary Containment Hydrogen Recombiners ... .... ... ....... 3 . 6 . 3 . 1-1 3.6.3.2 Primary Containment Oxygen Concentration ... .... .. ..... ... 3.6.3 . 2-1 3 . 6 . 4.1 Secondary Containment . ... .. . ..... . ......... .... .. ........ 3 . 6. 4. 1-1 3 . 6.4. 2 Secondary Co ntai nment Isolation Valves (SCIVs) ........... 3 . 6 . 4. 2-1 3 . 6.4 . 3 Standby Gas Treatment (SGT) System ...................... . 3.6.4 . 3- 1 (continued)

LaSalle 1 and 2 ii Amendment No.

Definitions

1. 1 1.1 Definitions (continued)

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

CORE ALTE RAT IO N COR E ALTERATION shall be the movement of any fuel, sources, or reactivity control compone nts, wit hin the reactor vessel with the ves sel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS :

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

and

b. Control rod move ment, provided there are no fuel assemblies in the associated core cell .

Suspension of CORE ALTERATIONS shall not preclude co mpleti on of movement of a component to a safe position .

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the cur rent reload cycle . These cycle specific limits shall be determined for each reload cyc l e in accordance with Spec ification 5 . 6 . 5 . Pl ant operation within these limits is addressed in individual Specifications .

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 ~hall be that concentration of I -131 Cmicrocuries/gram) that alone would produce the sa me thyroid dose as the quantity and isotopic mixture of I-131, I-132, I - 133, I-134, and I-1 35 actually present . The thyroid dose conversion factors used for thi s calculation shall be those listed in Table III of TID-14844, AEC , 196 2 , "Calculatio n of Distance Factor s for Power and Test Reactor Sites; " Table E-7 of Regulatory Guide 1.109, Rev . 1, NRC, 1977; or ICRP (continued)

LaSalle 1 and 2 1. 1- 2 Amendm ent No .

Definitions

1. 1 1.1 Definitions DOSE EQUIVALENT I-131 30 , Supplement to Part 1, pages 192-212, Table (continued) titled, "Committed Dose Equivalent in Target Organs or Ti ssues per Intake of Unit Activity . "

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

a. The water inventory above the TAF is divided by the limiting drain rate;

b. The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error), for all penetration flow paths below the TAF except:

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

2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when act ated by RPV water level isolation instrumentation; or (cont i nued)

LaSalle 1 and 2 1.1-3 Amendment No.

Definitions 1.1

1. 1 Definitions DRAIN TIME 3. Penetration flow paths with isolation (continued J devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation devices without offsite power .

c. The penetration f fow 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 .

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shal l be that t im e interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor unti l the ECCS equipment is capable of performing its safety function (i . e., the valves travel to their requ ir ed positions, pum p disc harge pressures reach the i r required values, etc . ) . Times s hall inc lu de diese l generator start i ng and sequence loading delays, where applicab l e. The response time may be measured by means of any series of sequential, over l apping, or tota l steps so that the entire respo nse ti me is meas ur ed. In lieu of measurement, response time may be verified for selected components provided that the components and method for verif1cation have been previously reviewed and approved by the NRC.

(continued)

LaSalle 1 and 2 1. 1-4 Amendment No .

Definitions

1. 1 1.1 Definitions (continued)

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial signal generation by (EOC-RPT) SYSTEM RESPONSE the associated turbine stop valve limit switch or TIME from when the turbine control valve hydraulic oil control oil pressure drops below the pressure switch setpoint to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker . The response time may be measured by means of any series of sequentia l , overlapp i ng, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and method for verification have been prev i ously reviewed and appro ved by t he NRC .

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

ISOLATI ON SYSTE M The ISOLATIO N SYSTE M RESPONSE TIME s hall be t hat RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions . The response time may be meas ured by means of any ser i es of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and method for verification have been previously rev i ewed and approved by the NRC .

(continued)

LaSalle 1 and 2 1. 1-5 Amendment No.

Definit i ons

1. 1

1. 1 Definitions (continued)

LEAKAGE LEAKAGE sha 11 be:

a. Identified LEA KAGE

1. LEAKAGE into t he drywe l 1 such as that from pump seals or val ve packing , th at is captured and conducted to a sump or col l ecting tank ; or

2. LEAKAG E i nto t he drywel l atmos ph ere fro m sources that are both specifical l y loca t ed and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Uni dent i f i ed LEAKAGE Al l LEAKAG E into the drywell that is not i dentified LEAKAGE ;

c. Tota 1 LEAKA GE Sum of the i dent i fied and unident i f i ed LEAKAGE ; and ct . Pressure Boundary LEAKAG E LEAKAGE th rough a noniso l ab l e fau l t i n a Reactor Coo l ant Sys t em (RCS) component body ,

pipe wal l , or vessel wall .

LINEA R HE AT GE NERA TI ON The LHGR s hall be the hea t ge neration r ate pe r RATE ( LHGR) uni t l ength of fuel ro d. It is the i ntegral of the heat flux over the heat tra nsfer area associated with the unit length .

LOGIC SYSTEM FUNC TIONA L A LOGIC SYSTEM FUNC TION AL TE ST s hall be a tes t TEST of all logic co mpone nt s requ i re d for OP ERAB IL ITY of a logic circuit, from as cl ose to the sensor as practicable up to, but not including , the actuated device, to verify OPERABILI TY. The LOGIC SYSTEM FUNC TIONAL TE ST may be perfo rm ed by mea ns of an y series of se que ntial, overlapp in g, or t ot al system steps so that t he entire logic syste m i s tested .

(co nt inu ed)

LaSalle 1 and 2 1. 1- 6 Amendment No.

Definitions

1. 1 1 .1 Definitions (continued)

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

MODE A MODE shall correspond to any one inclusive comb i nation 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.

OPERABL E-OPERAB IL ITY A system, subsystem, di visio n, component, or device shal l 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, coo l ing and seal water, lubr i cation, and other auxil i ary equ i pment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing the i r related support function(s).

RATED THERMAL POWER RTP shall be a total reactor core heat transfer

( RTP) rate to the reactor coolant of 3546 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that ti me interval SYSTEM (RPS) RESPONSE fro m when the monitored para meter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or tota l steps so that t he entire response time i s measured. In l ieu of measurement, response time may be verified for selected components provided that the components and method for verification have been previously reviewed and approved by the NRC .

(continued)

LaSalle 1 and 2 1. 1- 7 Amendment No.

Definitio ns 1.1 1.1 Definitions ( continued )

SHUTDOWN MARGIN (SOM ) SOM shall be the amount of reactivity by which the reactor i s subcritical or would be s ubcritical throughout the operating cy cle ass uming that:

a. The reactor is xenon free;

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

c. All control rods are fu ll y inserted except for the single control rod of highest reactivity worth, which i s assumed to be fully withdrawn.

With control rods not capable of being fu l ly inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channe l s, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components i n the associated function.

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

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

LaSalle 1 and 2 1.1 -8 Amendment No.

Definitions

1. 1 Table 1 . 1-1 (page 1 of 1)

MODES REACTOR MODE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLANT TEMPERATURE (OF) 1 Power Operation Run NA 2 Startup Refuel c,i or Startup/Hot NA Standby 3 Hot Shutdown c,i Shutdown > 200 4 Co 1d Shutdown Cal Shutdown :,; 200 5 Refuel i ng CbJ Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned .

(b) One or mo re reactor vesse l head closure bo lt s less t han fully tensio ned .

LaSalle 1 and 2 1.1 - 9 Amendment No.

ECCS Instrumentation 3 . 3 .5.1 ACT IONS CONDITION REQUIRED ACTION COMPLETION TIME B. As required by B.l - - - - - - - - NOTE&- - - - - - - -

Required Action A.l 1. Only applicable and referenced in in MODU) 1, 2, Table 3 . 3 . 5 . 1-1 . and 3 .

&.--Only applicable for Functions l.a, l.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 discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions B.2 - - - - - - - - NOTE&- - - - - - - -

1. Only app1icable in MODES, 1, 2, and 3.

&.--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 Core Spray (HPCS) disco very of System inoperable . loss of HPCS initiation capability B. 3 Place chan nel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip .

(cont inu ed)

LaSalle 1 and 2 3 . 3.5. 1-2 Amendment No.

ECCS Instrumentation 3 . 3 . 5. 1 ACT IONS CONDITION REQU I RED ACTION COMPLETION TIME C. As required by C. 1 - - - - - - - - NOTE&- - - - - - - -

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

&.--Only applicable for Functions l.c and 2 . C.

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 capabi lity capability for is inoperable. feature(s) in both divisions C. 2 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status .

(conti nued )

LaSalle 1 and 2 3 . 3.5.1- 3 Amendment No.

ECCS Instrumentation 3 . 3. 5.1 ACTIONS CONDIT I ON REQUIRED ACT ION COMPL ETION TIME D. As required by 0. 1 - - - - - - - - NOTE&- - - - - - - -

Required Action A. 1 1. Only applicable and referenced in in MODE~ 1, 2, Table 3.3 .5.1 -1. and 3 .

.i.:--only applicable for Functions l.d, 1.e, 1.f, 1.g, 2.d,

2. e, and 2. f .

Declare supported feature(s) inoperab l e 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from when its redundant disco very of feature ECCS loss of initiation capability initiation is inoperable. capabi 1 ity for feature(s) in bot h divi sion s 0.2 --------NOTE---------

Only applicable for Functions l . d and

2. d .

Declare supported feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from inoperab l e . disco very of loss of initiation capabi lity for feature(s) in one division (continued)

LaSalle 1 and 2 3.3 . 5. 1-4 Amendment No .

ECCS Instrumentation 3 . 3.5.1 Table 3.3.5.1-1 (page 1 of 4)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER RE QUI RED FROM SPEC IF I ED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION COND ITION S FUN CT I ON ACTION A.1 REQUIREMENTS VALUE

1. Low Pre ssure Coo lant Injection-A (LPCI) and Low Pressure Core Spray (LPCS)

Subsystems

a. Reactor Vesse l Water 1,2,3,. 2c*, 1 B SR 3.3.5.1.1 ;,, -1 47.0 Level-Low Low Low, SR 3. 3. 5. 1. 2 inches Leve 1 1 SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5 SR 3.3.5.1 .6

b. Drywell Pressure-High 1, 2 ,3 2c*, 1 B SR 3. 3. 5. 1. 2 ,; 1. 77 psig SR 3. 3. 5. 1. 4 SR *3.3.5.1.5 SR 3.3.5.1.6

c. LPCI Pump A 1,2,3,. C SR 3.3.5.1.2 ,; 5.5 seco nd s Start-Time Delay SR 3.3.5.1. 4 Relay SR 3.3.5.1.5

d. Reactor Stea m 1,2,3 2 D SR 3.3 .5.1. 2 ;,, 490 ps ig and Dome Pressure-Low SR 3. 3. 5. 1. 4 ,; 522 ps ig (I njection Permi ss i ve) SR 3.3.5.1.5 SR 3.3.5.1.6 SR LL§.1.2 2: 49Q ~sig aRe SR J.J.5.1.4 ,; §22 ~sig

~R J.J 5.1 5 SR J.J.5.1.e

e. LPCS Pump Di sc harge D SR 3 . 3. 5. 1. 2 ;,, 1240 gpm and Fl ow-Low (Bypass) SR 3. 3. 5. 1. 3 ,; 1835 gpm SR 3. 3. 5. 1. 5

f. LPCI Pump A Discharge 1,2, 3,. D SR 3. 3. 5. 1. 2 ;,, 1330 gpm and Flow-Low (Bypass) SR 3. 3. 5. 1. 3 ,; 21 44 gpm SR 3.3.5.1.5

g. LPCS and LPCI A 1, 2,3 1 per va l ve D SR 3.3.5.1.2 ;,, 490 psig and Injection Line SR 3. 3. 5. 1. 4 ,; 522 psig Pressure-Low SR 3.3.5.1.5 (Injection Permissive) SR 3. 3. 5. 1. 6 1 f!eF 11alve SR J.J.5.1.2 2: 49Q f!Sig aR9 SR J.J.5.1.4 ,; 5 22 f!S i g SR J. J. § . 1. § SR J.J.§.1.e

h. Manual Initiation 1,2, 3,. C SR 3. 3. 5. 1. 5 NA (continued)

(al WReR asseeiatee [CC& s~esyste111(sl aFe Fe~~iFee te Ile QP[RAQb[ f!eF bCQ J.5.2 , "[CC& &R~tee 11R,"

( &a ) Al so requir ed to initiate the associated diesel generator (DG).

LaSalle 1 and 2 3.3 . 5 . 1-9 Amendment No .

ECCS Instrumentation 3 . 3 . 5.1 Table 3 . 3 . 5 .1 - 1 (pag e 2 of 4)

Emergency Core Cooling Sys tem In s trumentati on APPLICABLE CONDITION S MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNEL S PER REQUIRED SURVEILLAN CE ALLOWABLE FUN CT I ON CONDIT ION S FUNCTION ACTION A. l REQUI REMENT S VALUE

2. LPCI B and LPC I C Subsys tems

a. React or Vess el Wate r l,2, 3T 2<*, 1 B SR 3 . 3.5 .1.1 ~ -147.0 Level - Low Low Low, SR 3 . 3 .5.1. 2 inche s Level 1 4,...~- SR 3 . 3 .5.1.4 SR 3 . 3 .5.1.5 SR 3 . 3 . 5.1.6

b. Drywell Pre ss ur e-High 1, 2 ,3 l "' ) B SR 3 . 3 .5.1 . 2 s 1 . 77 psig SR 3.3.5. 1.4 SR 3. 3 . 5. 1.5 SR 3 . 3 . 5 .1.6 C. LPCI Pump B l, 2 , 3T C SR 3 . 3 . 5 .1. 2 s 5. 5 second s Start- Time Del ay SR 3 . 3 . 5 .1.4 Relay 4-~- SR 3 . 3 .5.1.5

d. Reactor Stea m Dome 1,2,3 2 D SR 3 . 3.5 .1.2 ~ 490 ps i g and Pre ss ure- Low SR 3 . 3 .5.1.4 s 522 psig (Injection Permi ss ive) SR 3 . 3.5 .1.5 SR 3.3.5. 1.6 4.....,4.... ~R J.J,§.l.2 ~ 4QQ psig aREI SR J.J.5.1.4 ~ §22 ps 19 SR J.J.5.1.5 SR J.J.§.1.§

e. LPCI Pump Band LPCI 1, 2 . ~ 1 per pump D SR 3. 3 .5.1.2 ~ 1330 gpm and Pump C Di sc harg e SR 3 . 3.5. 1. 3 s 2144 gpm Flow-Low (Bypa ss ) 4....,4.... SR 3 . 3.5. 1.5

f. LPCI Band LPCI C 1, 2 , 3 1 per valv e D SR 3.3 .5 . 1. 2 ~ 490 ps ig and Injection Lin e SR 3 . 3 . 5 .1.4 s 522 ps ig Pre ss ure-Low SR 3 . 3 . 5 .1.5 (Inj ec tion Permi ss iv e ) SR 3 . 3 .5 . 1.6 4....,4... l peF 11a111e SR J,J,§,l,,l ~ 4QQ psig aREI SR J.J,!i.L 4 522 psig SR J.J.§.1.§ "

SR J.J.§.1.(i

g. Manua 1 Initiation l, 2 , 3T C SR 3 . 3 . 5. 1. 5 NA 4- .,4-(continued)

(al 1/~eR assesi ateEI [GGS swllsystem(s ) a Fe Feqwi FeEI te Ile QP[RAQb[ peF bGG ;i. 5, 2.

( &a ) Al so required to in itiate the ass ociated DG.

LaSalle 1 and 2 3.3.5.1-10 Am endment No.

ECCS Instrumentation 3 . 3.5.1 Table 3.3.5.1-1 (page 3 of 4)

Emergency Core Cooling System Instrumentation APPLICABLE COND lTl ONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUN CT JON CONDITIONS FUN CT I ON ACT I ON A. 1 REQUIREMENTS VALUE

3. High Pressure Core Spray (HPCSl System

a. Reactor Vessel Water B SR 3. 3. 5. 1. 1 ;,: -83 inches Level -Low Low, SR 3. 3. 5. 1. 2 Level 2 SR 3 . 3. 5. 1. 4 SR 3.3 . 5.1.5 SR 3.3 . 5. 1.6

b. Drywell Pressure-High 1,2,3 B SR 3. 3. 5. 1. 2 ,; 1. 77 psig SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5 SR 3. 3. 5. 1. 6

c. Reactor Vesse l Water 2 C SR 3. 3 . 5 . 1. 1 s 66.5 i nch es Level-High, Level 8 SR 3. 3. 5. 1. 2 SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5

d. HPC S Pump Discharge D SR 3. 3. 5. 1. 2 ;,: 113.2 psig Pressure-High SR 3. 3 . 5. 1. 4 (Bypass) SR 3.3 .5.1. 5

e. HPCS System Flow D SR 3. 3. 5. 1. 2 ;,: 1380 gpm and Rate-Low (Bypass) SR 3. 3. 5. 1. 3 s 219 4 gpm SR 3.3 . 5.1.5

f. Manual Initi ation C SR 3.3.5.1 . 5 NA

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water l, 2(.i, ),3(ob ) 2 SR 3. 3. 5. 1. 1 ;,: -147.0 Level -Low Low Low, SR 3.3.5. 1. 2 in ches Level 1 SR 3.3.5.1. 4 SR 3.3.5.1.5

b. Drywell Pressure-High 1, 2(ob ) , 3(ob ) 2 SR 3.3.5.1.2 ,; 1. 77 psi g SR 3.3.5.1.4 SR 3.3.5 . 1.5

c. ADS Initiation Timer l,2 (ob ),3 (ob ) F SR 3.3.5.1.2 s 118 seconds SR 3 . 3.5 .1.4 SR 3 . 3.5 .1. 5 (continued)

(al W~eR assaeiate~ EGGS sY~system(s) aFe Fe~YiFe~ ta ~e QPERA8LE ~eF LGQ 3.§.2.

(&a l Al so requi red to initiate the associated DG.

(eb ) With rea ctor steam dome pressure> 150 psig.

LaSalle 1 and 2 3 . 3.5.1-11 Amendment No.

ECCS Instrumentation 3 . 3.5.1 Table 3.3.5.1-1 (page 4 of 4)

Emergency Core Cooling System In stru mentati on APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECI FIED CHAN NELS PER REQUIRED SURVEI LLANCE ALLOWABLE FUN CT I ON CONDITIONS FUN CT I ON ACTION A.1 REQUIREMENTS VALUE

4. ADS Trip System A (continued)

d. Reactor Vesse l Water SR 3.3.5.1.1 ;2, 11.0 inches Level -Low, Level 3 SR 3.3. 5.1. 2 (Confirmatory) SR 3 . 3.5.1.4 SR 3.3.5.1. 5

e. LPCS Pump Di sc harge 1, 21<1> ). 31<1> ) 2 F SR 3. 3. 5. 1. 2 ;2, 131. 2 psig Pressure-Hig h SR 3. 3. 5. 1. 4 and SR 3.3.5.1.5 :s 271.0 psig

f. LPCI Pump A Discharge 1, 21<1> ). 3<<1> ) 2 F SR 3. 3. 5. 1. 2 ;2, 105.0 psig Pre ssure-Hig h SR 3. 3. 5. 1. 4 and SR 3. 3. 5. 1. 5 :s 128 . 6 ps ig

g. ADS Drywell Pressure 2 F SR 3. 3. 5. 1. 2 :s 598 seconds Bypass Ti mer SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5

h. Manual Initiation 1. 2<<1> ). 31<1> ) 2 SR 3. 3. 5. 1. 5 NA

5. ADS Tr i p System B

a. Reactor Vessel Water 1, 21**>. 31.i,J 2 SR 3.3.5.1.1 ;2, -147.0 Level -Low Low Low, SR 3.3.5.1 . 2 inches Level 1 SR 3.3.5.1 . 4 SR 3. 3. 5. 1. 5

b. Drywell Pre ss ure-High 1, 21<1> ). 31.i, ) 2 SR 3 . 3. 5. 1. 2 :s 1. 77 ps ig SR 3. 3. 5. 1. 4 SR 3.3.5.1 . 5

c. ADS Initiation Timer 1,21.t> ),31.0 ) F SR 3.3.5.1 . 2 :s 118 seconds SR 3.3.5.1.4 SR 3.3.5. 1.5

d. Reactor Vesse l Water 1, 2<.0 ). 31.0 ) SR 3.3.5. 1.1 ;2, 11.0 i nches Level-Low, Level 3 SR 3 . 3.5.1.2 (Confirmatory) SR 3.3.5.1.4 SR 3 . 3. 5.1.5

e. LPCI Pumps B & C 1 '2<<1> ). 31<1> ) 2 per pump F SR 3.3. 5.1. 2 ;2, 105.0 psig Di sc harge SR 3 . 3 . 5. 1. 4 and Pre ssure-High SR 3 . 3. 5. 1. 5 :s 128.6 psig

f. ADS Drywel l Pressure 2 F SR 3. 3. 5. 1. 2 :s 598 seconds Bypass Timer SR 3. 3. 5. 1. 4 SR 3. 3 . 5. 1. 5

g. Manual Initiation 1. 21<1> ). 31.i, ) 2 F SR 3.3.5.1 . 5 NA (Gb ) With reactor steam dome pre ss ure > 150 psig.

LaSalle 1 and 2 3 . 3 . 5 . 1-12 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5 . 2

3. 3 INSTRUMENTATION 3.3.5 . 2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCD 3.3.5.2 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3. 3.5 . 2-1.

ACTIONS

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

Separate Condition entry is allowed for each channel .

CONDIT JON REQUIRED ACTION COMPLETION TIME A. One or more channels A.I Enter the Condition Immediately inoperable . referenced in Table 3.3.5 . 2-1 for the channe I.

8. As required by B. 1 Declare associated Immediately Required Action A .1 penetration flow and referenced in path(s) incapable of Table 3.3 . 5.2-1. automatic isolation .

AND 8.2 Calculate DRAIN TIME. Immediately C. As required by C.1 Place channel in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action A. 1 and referenced in Table 3.3.5.2-1.

(continued)

LaSalle 1 and 2 3.3 . 5. 2-1 Amendment No .

RPV Water Inventory Control Instrumentation 3.3. 5. 2 ACTIONS (continued )

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 Re s tore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.I OPERABLE s tatus .

and referenced in Table 3.3.5.2-1 .

E. Required Action and E.1 Declare associated Immediately associated ECCS injection/ spray Completion Time of subsystem inoperable.

Condition C or D not met.

SURVEILLANCE REQUIREMENTS

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

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

SURVEILLANCE FREQUENCY SR 3. 3.5. 2.1 Perform CHANNEL CHECK . In accordance with the Surveillance Frequency Control Program SR 3. 3. 5. 2. 2 Perform CHANNEL FUNCTIONAL TEST . In accordance with the Survei / lance Frequen cy Control Program LaSa l le 1 and 2 3.3 . 5.2-2 Amend men t No.

RPV Water Inventory Control Instrumentation 3 . 3 . 5. 2 Table 3.3.5 . 2-1 (page 1 of 2)

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

1. Low Pressure Coolant Injection-A (LPCIJ and Low Pressure Core Spray (LPCS)

Subsystems

a. Reactor Steam 4,5 1/ 1 ) C SR 3.3.5.2.2 s522psig Dome Pressure-Low (Injection Permi s sive)

b. LPCS Pump 4,5 1 per D SR 3 .3 . 5.2.2 ~ 1240 gpm and Discharge pump<*J $ 1835 gpm Flow-Low (Bypass)

C. LPCI Pump A 4,5 1 per 0 SR 3.3.5.2.2 ~ 1330 gpm and Discharge pump 1*1 $ 2144 gpm Flow - Low (Bypass)

d. LPCS and LPCI A 4,5 1 per C SR 3 .3.5 . 2. 2 s522psig Injection Line va I ve <o Pressure-Low

( Injection Permissive)

2. LPCI Band LPCI C Subsystems

a. Reactor Steam 4,5 C SR 3.3.5.2.2 s 522 psig Dome Pressure-Low (Injection Permissive)

b. LPCI Pump B 4, 5 1 per D SR 3. 3. 5. 2. 2 ~ 1330 gpm and and LPCI Pump C pump1*1 $ 2144 gpm Discharge Flow-Low (Bypass)

c. LPCI B and LPCI C 4,5 1 per C SR 3.3.5.2. 2 s 522 psig Injection Line va I ve<*i Pressure-Low (Injection Permissive)

(continued)

(a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3. 5. 2, "RPV Water Inventory Contra I . "

LaSalle 1 and 2 3 . 3 . 5. 2-3 Amendment No.

RPV Water Inventory Control Instrumentation 3.3.5 . 2 Table 3.3.5.2-1 (page 2 of 2)

RPV Water Inventory Control Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION COND I TIONS FUNCTION ACTIONA.1 REQUIREMENTS VALUE

3. High Pressure Core Spray (HPCSJ System

a. HPCS Pump 4, 5 1(*) D SR 3.3 . 5. 2. 2 2: 113 . 2 psig Discharge Pressure-High (Bypass)

b. HPCS System Flow 4, 5 1(

J D SR 3.3.5.2.2 2: 1380 gpm Rate-Low (Bypass) and s; 2194 gpm

4. RHR Shutdown Cooling System Isolation

a. Reactor Vessel (b )

2 in one B SR 3.3.5 . 2.1 2: 11.0 Water Level-Low, trip SR 3.3.5.2 . 2 inches Level 3 system

5. Reactor Water Cleanup (RWCU) System Isolation

a. Reactor Vessel (b) 2 in one B SR 3.3.5.2. 2 2: -58.0 Water Level-Low trip inches Low, Level 2 system (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3. 5. 2, "RPV Water Inventory Control."

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

LaSalle 1 and 2 3 . 3 . 5.2-4 Amendm ent No.

RCIC System Instrumentation 3 . 3.5 .iJ 3.3 INSTRUMENTATION 3 . 3 . 5. ~ 3 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3 . 3.5 .iJ The RCIC System instru mentation for each Function in Table 3 . 3.5. iJ -l shall be OPERABLE .

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

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channe l s A. 1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5. iJ -1 for the channel .

B. As required by B. 1 Dec l are RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action A. 1 inoperable . discovery of and referenced in loss of RCIC Table 3.3.5. iJ -1. initiation capability AND B.2 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip .

(continued)

LaSalle 1 and 2 3. 3 . 5.3-1 Amendment No .

RCIC System Instrumentation 3 . 3.5. J.J ACTIONS CONDIT ION REQ UIRED ACT ION COM PL ETION TIME C. As required by C.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.l OPERABLE status.

and referenced in Table 3 . 3 . 5. ~3 -1 .

D. As required by D. l - - - - - - - -N OTE - - - - - - - - -

Re qui red Action A. 1 Onl y appl i cabl e i f and refere nced i n RC IC pu mp suct ion is Tab l e 3 . 3 . 5 .~3 -1. not al igned t o the suppressio n pool .

Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable . discovery of loss of RCIC init i ation capab il ity D. 2. 1 Place channe l in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip .

D. 2. 2 Ali gn RCIC pump 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months suction to the sup press i on pool .

E. Re quired Action and E. 1 Declare RC IC Syste m Immed i ately associated Completion inoperable.

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

LaSal l e 1 and 2 3 . 3. 5. 3-2 Amendment No .

RCIC System Instrumentation 3 . 3.5. i3 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3 . 3 . 5 .i3 -l to determine which SRs apply for each RCIC Function .

2. When a channe l is placed in an inoperable status solely for performance of required Surveillances, entry into associated Condit i ons and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 4; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1 and 3 provided the associated Function maintains RCIC ini tiation capabi lity.

SURVEILLANCE FREQUENCY SR 3 . 3 . 5 . i3 . 1 Perform CHAN NEL CHECK . In accorda nce wit h the Sur veillance Frequency Control Program SR 3 . 3 . 5 .i3 . 2 Perf orm CHANNE L FUN CTI ONAL TEST . In accordance with the Surveillance Frequency Co ntrol Program SR 3 .3 . 5 .i3 .3 Perform CHANNEL CALIBRATION . In accordance with the Surveillance Frequency Control Program SR 3 .3. 5. i3 .4 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Sur veill ance Frequency Control Program LaSalle 1 and 2 3.3.5 . 3-3 Amendment No.

RCIC System Instrumentation 3.3.5 .-2-3 Ta bl e 3.3.5. ~3 - l (page 1 of 1)

Reactor Core Isol ation Coo ling System In s trumentat i on CO NDITION S REQUIRED REFERENCED FROM CHANNELS PER REQUIRED SURVE ILLAN CE ALLOWABLE FUN CT ION FUN CT I ON ACT ION A. 1 REQUIREMENTS VALUE

1. Reactor Vesse l Wate r 4 B SR 3. 3.5. ~3 .2 " -83 in ches Level - Low Low, Level 2 SR 3.3 . 5. ~3 .3 SR 3. 3.5 .~3 . 4

2. Reactor Vesse l Water 2 C SR 3. 3.5. ~3 .l s 66.5 inches Level-High, Lev el 8 SR 3. 3.5. ~3 .2 SR 3.3 .5. ~3 .3 SR 3. 3 . 5 .~3 . 4

3. Condensate Storag e Tank 2 D SR 3. 3 . 5. ~3 . 2 " 71 3.6 ft Leve 1- Low SR 3.3.5. ~3 .3 SR 3. 3.5. ~3 .4

4. Manua l Initi at ion C SR 3. 3. 5 .~3 . 4 NA LaSal l e 1 and 2 3 . 3 . 5. 3-4 Amendme nt No .

Primary Containment Isolation Instrumentation 3 . 3.6 . 1 Table 3.3 .6.1 - 1 (page 4 of 4)

Primary Co ntainment Isolation Instrumentation APPLICABLE COND ITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPEC IF I ED CHANNELS PER REQUIRED SU RVEILLAN CE ALLOWABLE FUN CT ION CONDITIONS TRIP SYSTEM ACTION C.l REQUIREMENTS VALUE

4. RWCU Sy s tem Isolation (continued)

k. Reactor Vesse l Water 1, 2 ,3 2 SR 3.3.6.1.2 ~ -58.0 inche s Leve 1- Low Low, SR 3.3.6.1.4 Level 2 SR 3.3.6.1.5

l. Standby Liquid 1, 2, 3 2 fb) SR 3.3.6.1.5 NA Control System Initiation

m. Manual Initiation 1,2,3 G SR 3.3.6.1.5 NA

5. RHR Shutdown Cooling Sys tem Iso lation

a. Reactor Ve sse l Water 3-r4,--& J SR 3.3.6.1.1 ~ 11.0 inc hes Level - Low, Level 3 SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5

b. Reactor Ves se l 1, 2, 3 SR 3.3.6.1.2 ,s 14 3 psig Pre ss ure- High SR 3.3.6.1.4 SR 3.3.6.1.5

c. Manual Initiation 1, 2, 3 G SR 3.3.6.1.5 NA Cb) Only input s into one of two trip sys tems.

Cs) GRly eRe trip system reqYires iR MQll[~ 4 aRs 5 *iit~ RMR ~~Ytse11R GeeliRg ~ystem iRtegrity maiRtaiRes .

LaSalle 1 and 2 3.3.6 . 1-9 Amendment No.

Seconda ry Containment Isolation In strume nta t ion 3 . 3 . 6. 2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES AND REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUN CT ION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3.,.4-} 2 SR 3.3.6.2.2 ~ -58.0 inches Leve l-Low Low, Le vel 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; 1.93 psig SR 3.3.6.2 . 3 SR 3.3.6.2.4

3. Reactor Building 1,2 ,3, 2 SR 3.3.6.2 .1 s; 42 . 0 mR/hr Ventilation Exhaust Plenum (a)-,{-&+ SR 3.3.6.2.2 Rad i at i on-High SR 3.3.6.2.3 SR 3.3.6.2.4

4. Fuel Poo l Ventilation 1,2 ,3, 2 SR 3.3.6.2.1 s; 42 . 0 mR/hr Ex haust Radiatio n- Hi gh (al-,{-&+ SR 3.3.6 . 2. 2 SR 3.3.6.2.3 SR 3. 3.6.2. 4

5. Manual Initiation 1,2 ,3, SR 3.3.6.2.4 NA (al-,{-&+

(al Q~riR9 e~eratieR& "iti'l a ~eteRtial fer araiRiR9 ti'le reaster vessel.

( ba ) During CORE ALTER ATIONS, and during movement of irradiated fuel assemblies in the secondary containment.

LaSa l le 1 and 2 3. 3 . 6. 2-4 Am end men t No.

CRAF System Instrumentation

3. 3 . 7. 1 3.3 INSTRUMENTATION 3.3.7 . 1 Control Room Area Filtration (CRAF) System Instrumentation LCO 3 . 3 . 7.1 Two channels per trip system for the Control Room Air Intake Radiation-High Function shall be OPERABLE for each CRAF subsystem.

APPLICABILITY : MODES 1, 2, and 3, During movement of irrad i ated fuel assemb l ies in the secondary containment, During CORE ALTERATIONST During operations with a potential for draining the reactor vessel (OPDRVs) .

ACTIONS

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

Separate Condit i on entry is al l owed for each cha nn el.

CONDIT ION REQUIRED ACTION COMP LETION TIME A. One or more channe l s A. 1 Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable . CRAF subsystem discovery of inoperable . loss of CRAF subsystem initiation capability A. 2 Pl ace chan nel in 6 ho urs trip .

(continued)

LaSalle 1 and 2 3.3.7.1-1 Amendment No.

LOP Instrumentation 3.3.8 . 1 Tab l e 3 . 3 . 8 . 1- 1 (page 1 of 1 )

Lo ss of Powe r In st rument ation REQUIRED CHANNE LS PER SURV EILLANCE ALLOWABLE FUN CT I ON DIVI SION REQUI REMENT S VALUE

1. Di vi s ion s 1, 2 and Oppos ite Unit Divi s ion 2 - 4.16 kV Eme rge ncy Bus Undervoltage

a. Loss of Voltage - 4.16 kV 2 SR 3 . 3.8. 1. 3 ~ 2870 V and 5 3127 V Ba s i s SR 3 . 3 . 8 .1.4 SR 3 . 3.8. 1.5

b. Lo ss of Voltage - Time Delay 2 SR 3 . 3 .8.1.3 ~ 3 .1 s econd s and 5 10 .9 seconds SR 3. 3. 8 . 1.4 SR 3 . 3 . 8 .1.5

c. Degrad ed Voltage - 4 . 16 kV 2 SR 3 . 3.8. 1.1 ~ 3814 V and 5 3900 V Ba s i s SR 3 . 3.8 .1. 2 SR 3 . 3.8 .1.5

d. Degraded Voltage - Time 2 SR 3 . 3 .8.1.1 ~ 27 0 .1 second s and Delay , No LOCA SR 3.3 .8 . 1.2 5 329.9 second s SR 3 . 3 . 8 .1.5

e. Degraded Voltage - Time 2<*Hbl SR 3.3 . 8 .1.1 ~ 9. 4 second s and s 10.9 second s Delay, LOCA SR 3 . 3. 8 . 1.2 SR 3 . 3 .8.1.5

2. Divi s ion 3- 4 . 16 kV Emergen cy Bu s Underv olt age

a. Loss of Voltage - 4.16 kV 2 SR 3.3.8. 1. 3 ~ 2725 V and s 3172 V Basi s SR 3 . 3.8 .1.4 SR 3.3.8 .1.5

b. Loss of Volta ge - Time De l ay 2 SR 3 . 3 . 8. 1. 3 s 10 . 9 second s SR 3.3 . 8 .1.4 SR 3 . 3.8. 1. 5

c. Degrad ed Voltage - 4 . 16 kV 2 SR 3 . 3.8. 1.1 ~ 3814 V and 5 3900 V Ba s i s SR 3 . 3. 8 . 1. 2 SR 3 . 3 . 8 .1.5

d. Degrad ed Voltag e - Time 2 SR 3 . 3 . 8 .1.1 ~ 270 . 1 second s and Delay , No LOCA SR 3 . 3 . 8 .1. 2 s 329 . 9 second s SR 3 . 3 . 8 .1. 5

e. Degraded Voltage - Time 2<*l(bl SR 3 . 3 . 8 .1.1 ~ 9 . 4 second s and 5 10.9 second s Delay, LDCA SR 3 . 3 . 8 .1. 2 SR 3.3.8. 1.5 (al In MODE S 4 and 5, ,*~eR a;;seiates [GG~ 5~1l;y;telll(5) arenot requir ed t o be OPERABLE . ~er bGQ J.!i.2 , "[GGS S~~t88\IR."

( b) With no fuel in the reactor ve sse l, not required to be OPERABLE.

LaSalle 1 and 2 3.3 . 8 . 1-3 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 electr i c power monitoring assemblies sha l l be OPERABLE for each inservice RPS motor generator set or alternate power supply.

APPLICABILITY: MODES 1, 2, and 3, MODES 4 and 5 with res i dual heat remova l (RHR) shutdown cooling (SOC) isolation valves open, MODE 5, with any control rod withdrawn from a core cell containing one or more fuel assemblies, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONST D~ring operations with a potential for draining the reactor vessel (OPDRVs) .

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

inoperable.

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

inoperable .

(continued)

LaSalle 1 and 2 3.3.8.2-1 Amendment No.

RPS Electric Power Monitoring 3 . 3.8.2 ACT IONS CONDITION REOU I RED AC TION COMP LETION TIME F. Required Action and F. 1. 1 Isolate the Immediately associated Completion associated secondary Time of Condition A or containment B not met during penetration fl ow movement of irradiated path(s).

fuel assemblies in the secondary containmentT OR or during CORE ALT ERATIONS , or during F.1 . 2 Declare t he Immedi ate l y OPDRIJs . associated secondary containment isolation valve(s) inoperable.

F. 2 .1 Pla ce the associated Immediately sta ndby gas treatment (SGT) subsys tem( s) in operation .

F.2.2 Declare associated Immediately SGT subsystem(s) inoperable.

LaSalle 1 and 2 3 . 3.8. 2-3 Amendment No .

ECCS-Operating

3. 5.1 3.5 EMERGEN CY CORE COO LI NG SYS TE MS (ECCS) , REACTOR PRESSURE VESSEL (RPV)

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

SYS TEM 3.5.1 ECCS-Operating LCD 3.5 . 1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICA BIL ITY: MODE 1, MODES 2 and 3, except 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 HPCS .

CONDITION REQUIRED ACT ION COMPLETION TIME A. One low pressure ECCS A.l Restore low pressure 7 days injection/spray ECCS injection/spray subsystem inoperable . subsyste m to OPERABLE status .

(continued)

LaSalle 1 and 2 3. 5.1 -1 Amendment No .

RPV Water I nventory ControlECCS Shutdo1m 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) , REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOL ING (RCIC)

SYS TEM 3.5.2 RPV Water Inventory Contra lECCS Shutdown LCO 3. 5. 2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

One:+wB- ECCS inje cti on/sp ray su bsy stem~ s hall be OPERAB LE.

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

Gfl.e--A L+ow P~ressure C~oolant I+njection (LPCI) subsyste m ma y be considered OPERABLE dur ing alignment and operation for decay heat removal, if capa ble of being man uall y realigned and not otherwise inoperable .

APPLICABILITY: MODES 4 and, rn 5 except with the spent fuel storage pool gates removed and 1,ater 1evel

~ 22 ft over the top of the reactor pressure vessel flange .

ACTION S CONDITION REQUIRED ACTION COMP LETION TIME A. ~ Requ ired ECCS A. 1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperab l e. subsystem to OPERABLE statu s.

B. Required Action and B.1 Init i ate act i on to Immedi ately associated Co mpletion suspend operations Time of Condition A \\1 i th a potential for not met. draining the reactor vessel (OPDRVs)establish a method of water injection capable of operating without offsite power .

(continued)

LaSalle 1 and 2 3 . 5. 2-1 Amendment No .

RPV Water Inventory ControlEGGS Shbltdoi,,m 3 . 5. 2 ACTIONS (continued)

COND IT ION REQU I RED ACTION COMPL ETION TIM E G. +'110 reEJbli red EGGS G.1 IAitiate aEtioA to Immediately i AJ eEti OA,lspra~, SblSpeAd OPDRl/s .

Sblbsystems iAoperable.

Amt G.2 Restore OAe reEJbli red 4 hoblrs EGGS i AJ eEti o-Aispray Sblbs~,stem to 0 PE R/\B LE statbls.

C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. l 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.

A!:f12 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 .

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

(continued)

LaSalle 1 and 2 3 . 5. 2-2 Amendment No .

RPV Water Inventory ControltCCS Shutdo1m 3.5 . 2 ACT ION S (continued)

CONDITION REQUIRED ACT ION COMPLETION TIME D. DRAIN TIME 0.1 - - - - - - - NO TE - - - - - - - -

< 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .Req~irea Required ECCS Actien C.2 ana asseciatea injection/ spray subsystem Cempletien Time net met. or additional method of water injection shall be capable of operating without offsite electrical power.

Initiate action to establish an additional method of water injection Immediately with water sources capable of maint~ining RPV water level > TAF for

~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .restere secanaary cantainment te OP[RABb[ stat~s.

0.2 Initiate action to establish secondary containment Immediately boundaryrestere ene stanaby ~as treatment s~bsystem ta OP[ ~ABb~

-s-t-a-t-tl .

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

D. J4 Initiate action to verify one standby gas treatment subsystem is capable of Immediately being placed in operation.restere isalatien capability in each req~irea secanaary centainment penetratien fl m, patt:i net i sel a tea.

(con tinued )

LaSalle 1 and 2 3.5.2-3 Amendment No.

RPV Water Inventory Control't.CCS Shutdo*.m 3 . 5. 2 ACTIONS (continued)

CONDIT ION REQ UIRED ACT ION COMPLETION TI ME E. Required Action and E. l Initiate action to Immediately associated Completion restore DRAIN TIME to Time of Condition C or ~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D not met.

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

SURVE ILLANCE REQUI REME NTS 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 low pressure In accordance ECCS injection/spray subsystem, the with the suppression pool water level is~ - Surveillance 12 ft 7 in. Frequency Control Program SR 3.5.2. 3~ Verify, for ,t.l:\.e-a required High Pressure In accordance Core Spray (HPCS) Syste m, the suppression wi t h the pool water leve l i s~ -12 ft 7 i n. Surve i llance Frequency Control Program SR 3.5 . 2 . 4~ Verify , f or ~the re qu ired ECCS In accordance injection/spray subsyste m, locations with the susceptible to gas accumulation are Surveillance sufficiently filled with water . Frequency Control Program (continued)

LaSalle 1 and 2 3.5.2-4 Amendment No.

RPV Water Inventory ControlECCS Shutdo*.m 3.5 . 2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUE NCY SR 3.5.2. 54 - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - -

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

Verify , for the~ req uired ECCS In accorda nce injection/spray subsyste m, each manual, with the power operated, and automatic va l ve in the Surveillance f l ow pa th , that is not locked, sealed, or Fr equ ency otherwise secured in position, is in the Control Program correct position .

SR 3 . 5.2 . 6~ Operate the required ECCS injection/ spray In accordance subsystem through the recirculation line for with the

~ 10 mi nut es . Verify each reqbli red EGGS pblmp INSERVICE develops the specified flow rate against the TESTJrl.JG specified test line pressblre . PROGRl\MSurvei l l ance Frequency HST LI~IE Control Program SYSHM fLOW R.l\H PRESSURE LPGS c! 0ds0 gpm c! 290 psig LPGI ;, 7200 gpm ;, no psig FIPGS (Unit u ;, 02§0 gpm ;, d70 psig FIPGS (Unit 2) ;, 0200 gpm ;, dd0 psig SR 3.5.2.7 Verify each valve credited for automatically In accordance isolating a penetration flow path actuates with the to the isolation position on an actual or Surveillance simulated isolation signal. Frequency Control Program SR 3.5.2. 8-a. - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Vessel i njectio n/spray may be excluded .

Verify ~ t h e required ECCS In accordance i njection/spray subsystem actuates on an with the actual or simulated automatic initiation Surveillance signalcan be manually operated . Frequency Control Program 24 months LaSalle 1 and 2 3 . 5. 2-5 Amendment No .

RCIC System

3. 5. 3

3. 5 EM ERGENCY CORE COOLING SYSTEMS (ECCS) , REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISO LATION COOLING (RCIC)

SYSTEM 3.5.3 RCIC System LCO 3. 5.3 The RC I C Sy s t em s ha l l be OPERA BLE.

APPLICABI LITY : MODE 1, MODES 2 and 3 with reac t or stea m do me pressure> 150 ps i g .

ACTIONS

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

LCO 3.0.4. b is not app li cab l e to RCIC .

CON DITIO N REQ UIRED ACT IO N COM PLETION TI ME A. RC IC Syste m A. 1 Ver i fy by Immediately inoperable . adm i ni strat i ve means High Pressure Core Sp ray Syste m is OPERA BLE .

AND A. 2 Restore RCIC System 14 days to OPERAB LE status .

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Complet i on Tim e not met . AND 8.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months dome pressure to

~ 150 psig.

LaSalle 1 and 2 3.5.3-1 Amendme nt No .

PCIVs 3.6.1.3 3 . 6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3. 6 . 1. 3 Each PCIV shall be OPERABLE.

APPLICABILITY : MODES 1, 2, and 3, When associated instrymentation is reqYired to be OPERABLE per LCO 3. 3. 6.1, "Primary Containment Isolation Instrymentation . "

ACTIONS NOTES - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1. Penetration f l ow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetrat i on flow path.

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

4. Enter applicable Conditions and Required Actions of LCO 3.6 . 1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criter i a.

CONDIT ION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A. l Iso l at e t he affecte d 4 ho urs except Only applicable to penetration flow path for main steam penetration flow paths by use of at least line with two or more one closed and PCIVs . de-activated automatic va l ve, closed manua l valve, 8 hou r s for ma i n One or more blind flange, or steam line penetration flow paths check valve with flow with one PCIV through the valve inoperable for reasons secured.

othe r than Co nd ition D. AN D (continued)

LaSalle 1 and 2 3.6.1.3-1 Amendment No .

PC IVs 3 . 6.1.3 ACT IONS CONDITION REQUIRED ACT ION COMPLETION TIME 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 Ti me of Condition A, AND B, C' or D not met in MODE 1' 2' or 3. E. 2 Be i n MODE 4 . 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. ReEj!:li FeEI ActieA aAEI F .1 IAitiate actieA te Imme Eli atel y asseciateEI CempletieA S!:lSpeAEI epeFatieAs

+iFAe 9f CeAElitieA ,1, ti ',\Ii tR a peteAtial f9F g' C' 9F Q A9t met f9F EIFaiAiAg tRe FeacteF PCPH s) FeEJ!:li FeEI te se vessel ( OPQRVs).

OPERABLE 9!:lFi Ag MOQE 4

~ QR F.~ IAitiate actieA te Imme Eli a tel y FesteFe i,<ali,<e(s) te 0 PE R.t.,B LE stat!:ls.

LaSalle 1 and 2 3.6.1.3-5 Amendment No .

Secondary Containment 3.6 . 4.1

3. 6 CONTAINMENT SYS TEM S

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 irradiated fuel asse mblie s in the secondary contain ment, During CORE ALTERATIONST D1:1rin9 operations 11ith a potential for draining the reactor vessel COPDRVs) .

ACT IONS CONDITION REQUIRED ACTION COMPL ETION TIME A. Secondary containment A.1 Restore secondary 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s i noperable i n MODE 1, containment to 2' or 3. OPERABLE statu s .

B. Required Action and B.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Ti me of Co ndition A not met.

(con tinued )

La Sal le 1 and 2 3.6 .4.1 - 1 Amendm en t No.

Secondary Containment

3. 6. 4. 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment C. l - - - - - - - - NOTE- - - - - - - - -

inoperable during LCO 3.0. 3 is not movement of irradiated applicable .

fuel assemblies in the secondary containment, or during CORE Suspend movement of Immediately ALT ERATIONS , or durin~ irradiated fuel OPQRVs . assemblies in the secondary containment .

C. 2 Suspend CORE Immediately AL TE RAT IONS.

G.3 Initiate action to Immediately suspend OPQ RVs.

LaSalle 1 and 2 3.6 . 4. 1-2 Amendment No.

SC IVs 3 . 6. 4.2 3 . 6 CO NTAINMENT SYSTEMS 3 . 6.4 . 2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6 . 4.2 Each SCIV shall be OPERAB LE.

APPLICABILITY : MODES 1, 2, and 3 ,

During movement of irradiated fuel assemb lies in the secondary containment, During CORE ALTERATIONS, Q1,1rin9 operations 11ith a potential for draining the reactor vessel (OPQRVs) .

ACTIONS

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

1. Penetration flow paths may be unisolated intermittently under adm i nistrative controls .

2. Se parate Condition entry is al l owed for eac h penetration flow path.

3. Enter applicable Cond iti ons and Required Actions for systems made inoperable by SC IV s .

CONDIT ION REQUIRED ACT ION COMP LETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pe netration flow paths penetration flow path with one SC IV by use of at l east inoperable . one closed and de-activated automatic valve, clo sed ma nual valve, or bl ind fl an g.

AND (continued)

LaSalle 1 and 2 3 . 6. 4. 2-1 Amendment No.

SC IVs

3. 6. 4. 2 ACT IONS CONO IT ION REQUIRED ACTION COMPLETION TIME

0. Required Action and 0.1 - - - - - - - - NOTE - - - - - - - - -

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

or B not met during movement of irradiated fuel assemb l ies in the Suspend move ment of Immediately secondary containment, irradiated fuel or during CORE assemblies in the ALTERATIO NS, or durin~ secondary OPDRVs . co nt ai nment .

0. 2 Suspend COR E Immediately ALTE RA TI ON S.

D.3 Initiate action to Immediately suspend OPDRVs.

LaSalle 1 and 2 3. 6. 4.2-3 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 shal l be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary contain ment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vess el COPDRl/s) .

ACTIONS CONDIT ION REQ UIRED ACTION COMP LETION TIME A. One SGT s ub syste m A. 1 Restore SGT subsystem 7 days inoperable. to OPERABLE status .

B. Requ i red Act i on 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 MODE 1' 2' or 3.

C. Required Action and - - - - - - - - - - - - NOTE- - - - - - - - - - - - -

associated Completion LCO 3 . 0.3 is not app li cable .

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

not met during movement of irradiat ed C.l Pl ace OPERABLE SGT Immediately fuel assemblies in the subsystem in secondary contain ment, operation .

or during CORE ALTERATIONS , or during QR OPDRVs .

(continued)

LaSalle 1 and 2 3. 6.4.3-1 Amendment No .

SGT System 3.6 . 4 . 3 ACTIONS COND ITION REQ UIRED ACTION CO MPLETION TIME C. (continued) C. 2.1 Suspend movement of Immediately irradiated fuel assemblies in the secondary co nta in ment.

C. 2.2 Suspend CORE Immed i ately ALTERATIONS .

C.2.3 Initiate action to Immediately Sb1Spend OPDRVs.

D. Two SGT s ub systems 0. 1 Be in MODE 3 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s inoperable in MODE 1, 2 , or 3 .

E. Two SGT s ub systems E. l - - - - - - - - NOTE- - - - - - - - -

inoperable during LCO 3 . 0.3 is not movement of irradiated applicable .

fuel assemblies in t he secondary containment, or during CORE Suspend movement of Immediately ALTERATION S, or db1rin9 irradiated fuel OPDRVs . assembli es i n the secondary containmen t.

E.2 Suspend CORE Immediately AL TERA TI ONS.

E.3 Initiate action to Immediately Sb1Spend OPDRl/s.

LaSalle 1 and 2 3 . 6 . 4 . 3-2 Amendment No.

SGT System 3.6 . 4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4 . 3. 1 Operate each SGT subsystem for In accordance

~ 15 continuous minutes with heaters with the operating. Surveillance Frequency Control Program SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance wi th the Ve nt ilation Filter wit h the VFT P Testing Program (VFTP).

SR 3. 6.4.3 . 3 Verify each SGT subsystem actuates on an In accordance actua l or sim ul ated in i t i ation s i gnal. wi th the Surveillance Frequency Control Program LaSalle 1 and 2 3.6.4.3-3 Amendment No.

CRAF System 3.7 . 4

3. 7 PLANT SYSTEMS 3 . 7. 4 Control Room Area Filtration (CRAF) System LCO 3 . 7. 4 Two CRAF subsystems shall be OPERABLE .

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

The control room envelope (CRE) boundary may be opened intermittently under admin i strative control .

APPLICABILITY: MODES 1, 2, and 3 ,

During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONST D~ring operations with a potential for draining the reactor vessel (OPDRl/s ) .

ACTIONS CONDIT IO N REQ UIR ED ACTION CO MPL ETIO N TIM E A. One CRAF s ubsy stern A. 1 Restore CRAF 7 days inoperable for reasons subsystem to OPERABLE other than Condition status.

B.

B. One or more CRAF B.l Initiate action to Immediately subsystems inoperable implement mitigating due to inoperable CRE actions .

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

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

(continued)

LaSalle 1 and 2 3.7.4 - 1 Amendment No .

CRAF Sy stem 3 . 7. 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.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 or B not met in MODE 1, 2, or 3 .

D. Req ui red Action and - - - - - - - - - - - - NOTE- - - - - - - - - - - - -

associated Complet i on LCO 3.0.3 is not applicable.

Ti me of Condition A not met during movement of irradiated 0.1 Place OPERABLE CRAF Immediately fuel assemblies in the subsystem in secondary containment, pressurization mode .

or during CORE ALTERATIONS , or during OR OPDR'!s .

D. 2 . 1 Suspend movement of Immediately i rr ad i ate d fuel assembl ies i n the secondary containment.

0.2.2 Suspend CORE Immediately ALTERATIONS.

0. 2. 3 Initiate action to Immediately suspend OPDR'!s.

E. Two CRAF subsystems E .1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable in MODE 1, 2, or 3 for reasons other than Condition B.

(cont i nued)

LaSalle 1 and 2 3.7.4 -2 Amendment No.

CRAF System

3. 7. 4 ACT IONS CONDITION REQUIRED ACT ION COMPL ETION TIME F. Two CRAF subsystems - - - - - - - - - - - -NOTE - - - - - - - - - - - - -

inoperable during LCO 3. 0.3 is not appli cable .

movement of irradi ated fuel assemblies in the secondary containmentT F.l Suspend movement of Immediately or dur ing CORE irradiated fuel ALTERATIONS , or during assemblie s in the OPDRVs . secondary containment .

One or more CRAF subsy stems inoperable F. 2 Suspend CORE Immediately due to inoperable CRE AL TE RA TI ON S.

boundary during movement of irradiated Afil-0 fuel assemblies in the secondary containmentT ~F-.-3~~-I-n-i-t-i-a~t-e--a-c-t-io~n+-<t--o Immediately or during CORE suspend OPDRVs.

ALTE RATIO NS, or during OPDRVS .

SURVEILLAN CE REQUIREMENTS SURVE ILLAN CE FREQUEN CY SR 3.7 . 4.1 Operate each CRA F subsystem for In accordance

~ 15 cont inuou s minute s with the he aters with the operating . Sur veillan ce Fr eque ncy Contro l Program (continued)

La Sal le 1 and 2 3. 7. 4- 3 Amendment No.

Control Room Area Ventilation AC System 3.7.5 3.7 PLANT SYS TEMS 3 . 7. 5 Control Room Area Ventilation Air Conditioning (AC) Syste m LCO 3 . 7. 5 Two co ntrol room area ven t ilatio n AC subsystems shall be OPERABLE.

APPLICA BIL ITY : MOD ES 1, 2, an d 3, Dur ing moveme nt of irr adi ated f uel asse mbli es i n t he secondary contain ment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel COPDRl/s) .

ACTIONS CONDIT ION REQUIRED ACTION COMPLETION TIME A. One control room area A. 1 Res t or e co nt rol roo m 30 days ve ntilatio n AC area venti l ation AC subsystem inoperable . subsystem to OPERABLE status .

B. Tw o cont r ol room area B. l Verify co ntr ol room Once per 4 ve nt i lation AC area temperature hours subsystems inoperable . < 90°F .

B.2 Restore one control 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months roo m area ventilation AC subsystem to OPERABLE status .

C. Req uired Act i on and C.l Be i n 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, 2 , or 3.

(contin ued)

LaSalle 1 and 2 3.7.5- 1 Amendment No.

Control Room Area Ventilation AC System 3 . 7. 5 ACT IONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and - - - - - - - - - - - - NOTE- - - - - - - - - - - - -

associated Completion LCO 3 . 0. 3 is not app l icable .

Time of Condition A not met during movement of irrad i ated 0.1 Pl ace OPERABLE Immediately f uel assemb l ies in the control roo m area secondary containment, ventilation AC or during CORE subsystem in ALTERATIONS , or d1:Jrin~ operation .

OPDRl/s .

0. 2. 1 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment.

0. 2. 2 Suspend CORE Immediately AL TE RATIONS .

AN-Q.

D.2 . 3 Initiate action to Immediately Sl:JSpend OPDRl/s.

(continued)

LaSalle 1 and 2 3.7 . 5-2 Amendment No.

Control Room Area Ventilation AC System

3. 7. 5 ACT IONS CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and - - - - - - - - - - - - NOTE- - - - - - - - - - - - -

associated Completion LCO 3 . 0.3 is not applicable.

Time of Condition B not met during movement of irradiated E.1 Suspend movement of Immediately fuel asse mbli es in the irradiated fuel secondary containment, assemblies in the or during CORE secondary ALTE RATIO NS, or durin9 containment.

OPQRVs .

E.2 Suspend CORE Immediately AL TERA TI ONS .

E.3 Initiate action to Immediately suspend OPQRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Monitor control room and auxiliary electric In accorda nce equipment room temperatures . with the Surveillance Frequency Control Program SR 3.7.5.2 Verify correct breaker al ignment and In accordance indicated power are avai l able to the with the control room area ventilation AC Sur veillan ce subsystems . Frequency Co ntrol Program LaSalle 1 and 2 3.7.5-3 Amendment No.

AC Sources-Shutdown 3.8.2 ACTIONS COND IT ION REQUIRED ACTION COMPLETION TIME A. (continued) A.~ . J IRHiate i3Eti 9R te ImmeE!i atel;i,,

SYSf;)eREI ef;)erati eRs 1,<i tA i3 f;)eteRti al f:eF Elrai Ri Rg tAe Fei3Et9F vessel (OPDRl/s).

A@

A.2. 34 Init iate ac ti on to Immediately restore required offsite power circuit to OPERABLE status .

B. Required DG of LCO B.l Suspend CORE Immediately Item b. inoperable. ALTERA TIONS .

AND B.2 Suspend move ment of Immediately irradiated fuel asse mblies in secondary containment .

8:H:lt B.J IRitiate i3Eti9R te Imme Eli a tel J' SYSf;)eRE! OPORIJs AND B. 34 Initiate action to restore required DG Immediately to OPERABLE status .

C. Required DG of LCO C. l Declare High Pressure 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s It em C . inoperable . Core Spray System inoperable .

(continued)

LaSalle 1 and 2 3.8 . 2-3 Amendment No.

AC Sources-Shutdown 3.8. 2 ACT IONS CONDITION REQU I RED AC TI ON COMPLETION TIME D. Required offsite 0.1 Declare associated Immediately circuit or DG of LCO standby gas treatment Item d . inoperable . subsystem, control roo m area fi l tration subsystem, and control room area ventilation air cond i tioning subsyste m i noperab l e .

SURVEI LLAN CE REQ UIREMEN TS SURVEILLANCE FREQUENCY SR 3 .8 . 2. 1 - - - - - - - - - - - - - - - - - - - NOT ES- - - - - - - - - - - - - - - - - - -

1. The follow i ng SRs are not required to be performed : SR 3 .8.1.3, SR 3.8.1.9 through SR 3 . 8 . 1.11, SR 3 .8 . 1. 13 through SR 3 . 8.1 . 16, SR 3 . 8 . 1. 18, and SR 3 . 8 . 1.19.

2. SR 3 . 8.1 . 12 and SR 3.8.1.19 are not required to be met . when associated EGGS s~bsystem(s) are not req~ired to be OPERABLE per LGO 3.5.2, "EGGS Sh~tdo1,m."

For AC sources required to be OPERABLE, the In accordance SRs of Specification 3 . 8 .1, except wit h applicab l e SR 3 . 8.1 . 8, SR 3 . 8 . 1. 17 , and SR 3 . 8 . 1. 20, SRs are applicable.

LaSal l e 1 and 2 3.8.2-4 Amendment No .

DC Sources-Shutdown 3.8 . 5 ACTIONS CONDITION REQ UIRED ACTION COMP LETION TIME B. One or more requ i red B.l Declare affected Immediately DC el ectri cal power required feature (s )

subsystems inoperable inoperable.

for reasons other than Condition A. Q.R B. 2 . 1 Suspend CORE Immediately ALTE RA TI ON S .

Requi red Action and Co mpletion Tim e of Condition A not met.

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

B.2 . 3 Initiate action to Immediately sYspend operations with a potential for draining t he reactor vessel .

B.2. 34 Initiate action to Immediately restore required DC el ectrica l power subsystems to OPERABLE status.

LaSalle 1 and 2 3.8.5-3 Amendment No.

Distribution Systems-Shutdown 3.8 .8 ACT IONS CONDIT ION REQUIRED ACT ION COMPLETION TIME A. (continued) /\ . 2 . 3 Initiate action to Immediately s~spend operations with a potential for drainin~ the reactor vessel.

A.2. 34 In i t i ate ac t i ons to Immedi ate l y restore required AC and DC electrical power distribution s ubsystems to OPERA BLE st at us .

A. 2 . 4~ Declare associated Immediately req ui red s hu t do wn coo li ng subsystem(s) inoperable and not in operation .

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUE NCY SR 3.8.8.1 Verify correct breaker alignments and In accordance voltage to required AC and DC electrical with the power distribution subsystems . Sur veillance Frequency Contro l Program LaSalle 1 and 2 3 . 8.8-2 Amendment No.

LaSalle County Station, Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 3- REVISED TECHNICAL SPECIFICATIONS PAGES TOG Page i 3.3.5.2-3 3.6.1.3-1 TOG Page ii 3.3.5.2-4 3.6.1.3-5 1.1-2 3.3.5.3-1 3.6.4.1-1 1.1-3 3.3.5.3-2 3.6.4.1-2 1.1-4 3.3.5.3-3 3.6.4.2-1 1.1 -5 3.3.5.3-4 3.6.4.2-3 1.1-6 3.3.6.1-9 3.6.4.3-1 1.1-7 3.3.6.2-4 3.6.4.3-2 1.1-8 3.3.7.1-1 3.6.4.3-3 1.1-9 3.3.8.1-3 3.7.4-1 3.3.5.1-2 3.3.8.2-1 3.7.4-2 3.3.5.1-3 3.3.8.2-3 3.7.4-3 3.3.5.1-4 3.5.1-1 3.7.5-1 3.3.5.1-9 3.5.2-1 3.7.5-2 3.3.5.1-10 3.5.2-2 3.7.5-3 3.3.5.1-11 3.5.2-3 3.8.2-3 3.3.5.1 -12 3.5.2-4 3.8.2-4 3.3.5.2-1 3.5.2-5 3.8.5-3 3.3.5.2-2 3.5.3-1 3.8.8-2

TABLE OF CONTENTS

1. 0 USE AND APP LICATION

1. 1 Definitions ... . ........ . ...... . ..... . . .. . . .. . ...... .. . . .. 1 . 1- 1

1. 2 Logi ca 1 Connectors ..... ... ............ .. . . ..... . ......... 1. 2-1

1. 3 Completion Times . .. .... . . . ...... . .. . .. .... ... . . .... . ... . . 1.3-1

1. 4 Frequency ..... . ......................... .. ...... . ..... . .. 1 . 4- 1

2. 0 SAFETY LIMITS (SLs)

2. 1 SLs . .... ... . ........ . . .... . .. . ........ .... . . .. ..... . .. .. . 2.0-1

2. 2 SL Vi olatio ns .. . ... . ..................................... 2 . 0-1 3.0 LIMI TIN G CON DI TION FOR OPE RATION (LCO) APP LICAB ILI TY ........ 3 . 0- 1 3.0 SURVEIL LANCE REQUIREMENT (SR) APP LICAB IL ITY ................. 3 . 0-4 3.1 REACTIVITY CONTROL SYS TEMS 3 .1.1 SHUTDOWN MARGIN (SD M) ................... . . . .. .. .. ....... . 3.1. 1- 1 3 . 1. 2 React i vity Anomalies ........... . ......................... 3 . 1 . 2-1 3 . 1. 3 Con t ro l Rod OPERAB ILIT Y. . .... . . . ..... . .. .. .. . . . .......... 3 . 1.3-1 3 . 1.4 Control Rod Scram Ti mes ................. . ................ 3 . 1.4 - 1 3 .1. 5 Contro l Rod Scram Accumulators ....... .. . . . .. ... . . ....... . 3 . 1.5 - 1 3 . 1. 6 Rod Pattern Control ..................... . .. . .. . .. .. ... . .. 3 . 1.6-1 3 . 1. 7 Standby Li quid Contro l (SLC) System . ... . ... . . ............ 3 .1. 7-1 3 .1.8 Sc ram Disc har ge Vo lume (SD V) Vent and Drain Val ves .. . . . .. 3. 1 .8- 1 3.2 POWER DISTRIBUTION LIMITS

3. 2 .1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) .. . .. . 3.2.1 - 1

3. 2. 2 MINIMUM CR ITICAL POW ER RATIO (MCPR) ........ . ... . . . . ...... 3 . 2 . 2-1

3. 2. 3 LINEAR HEAT GENERAT ION RATE (LHGR) ...................... 3 . 2 . 3-1 3.3 I NSTR UME NTATIO N 3.3.1.1 Reactor Protection System (RPS) Instru mentation .... .. . .. . 3 . 3 . 1 . 1-1 3.3 . 1.2 Source Range Monitor (SRM) Instrumentation ............... 3 . 3.1.2-1 3 . 3 . 1.3 Oscillation Power Range Monitor (OPRM) Inst r umentation ... 3 . 3 . 1.3-1 3.3. 2. 1 Control Rod Bl ock In strume nt at ion ................. . ...... 3 . 3 . 2 . 1-1

3. 3. 2. 2 Feedwater System and Main Turbine Hig h Wate r Level Trip Instru mentat i on .............. ........ . ... . ..... ... 3 . 3 . 2 . 2-1 3 . 3. 3 . 1 Post Accident Monitoring (PAM) Instrumentation . . .. . . ..... 3 . 3 . 3 . 1-1 3.3.3.2 Re mote Shutdown Monitoring System ..... ...... . .. . . ........ 3 . 3 . 3 . 2-1 3.3.4 . 1 End of Cycle Recirculation Pump Trip (EOC-RP T)

In st rume ntation .... . .... . ......... ................. . ... 3.3 .4. 1-1 3.3 . 4.2 Ant i cip ate d Transie nt Wit hou t Scra m Recircul ation Pu mp Trip (ATWS-RPT) Instru mentation ..... .............. 3 . 3 . 4 . 2-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation . .... 3 . 3 . 5 . 1-1 3.3. 5. 2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ................... . ... . . ... ... .... . . . .. 3 . 3 . 5. 2-1 3.3. 5.3 Reactor Co r e Isolation Cool i ng (RC IC) System In st rumenta ti on . ... ............. .. .. .................. . 3 . 3. 5 . 3-1 3.3.6.1 Pri mary Containment Isolation Instrumentation . .... ....... 3 . 3 . 6 . 1-1 (continued)

LaSalle 1 and 2 Amendment No.

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6. 2 Secondary Containment Isolation Instrumentation .. .. . .. ... 3.3.6 . 2-1 3.3.7.1 Control Room Area Filtration (CRAF) System Instrumentation ......... . ........ . ... . ....... .. . . ...... 3 . 3 . 7. 1-1

3. 3. 8 . 1 Loss of Power (LOP) Instrumentation . .. .... ..... .. .... . ... 3 . 3 . 8 . 1-1 3.3.8 . 2 Reactor Protection System (RPS) Electric Power Mon i toring . ................ . ............. . ..... . . .. .... 3 . 3 . 8 . 2 - 1 3.4 REACTOR COOLANT SYSTEM (RCS) 3 . 4. 1 Recirculation Loops Operating ..... .. .. .. . ................ 3 . 4 . 1-1 3 . 4. 2 Flow Contro l Valves (FCVs) ................. .. .... .. .. . ... 3 . 4. 2-1 3.4.3 Jet Pumps ........... . ...... . .. . .... . ..................... 3 . 4. 3-1 3.4.4 Safe t y/Relief Va l ves (S/RVs) . .. ....... . .. . ... . . .......... 3 . 4 .4 - 1 3.4.5 RCS Operational LEAKAGE . . .............. . .. ... . .. . ..... ... 3 . 4. 5- 1 3.4 . 6 RCS Pressure Isolation Valve (PIV) Leakage ........ . ...... 3 . 4 . 6-1 3.4.7 RCS Leakage Detection Instrumentatio n ......... . . . ... .. . .. 3.4.7 - 1 3 . 4 .8 RCS Specif i c Activity . ... .. . .. ... . ....... .. .............. 3 . 4 . 8-1 3 . 4. 9 Residual Heat Remova l (RHR) Shutdown Cooling System-Hot Shutdown .................................... 3 . 4 . 9-1 3 . 4 . 10 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdo wn .................. .. ............... 3 . 4 .1 0-1 3 . 4 .11 RCS Pr ess ur e and Temp eratu r e ( P/T) Li mi ts . ... ... ......... 3 . 4 .11 -1 3 . 4 . 12 Re act or Steam Dome Pressure . . ............................ 3. 4 .1 2-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

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

SYSTEM

3. 5 . 1 EC CS-0 per at i ng ......................... . ........... .. . ... 3 . 5 . 1 - 1

3. 5. 2 RPV Water Inventory Control .... . .... . . .... ............... 3 . 5. 2-1 3.5.3 RC IC-System . . .... ........ . .... . . . ... . . .... .. .... . . . .. .... 3 . 5. 3-1 3.6 CONTAI NME NT SYS TE MS 3 . 6.1.1 Pr im ary Conta i nment ................. . .......... ......... . 3.6. 1. 1- 1 3.6 . 1.2 Primary Conta i nment Air Lock ... . ..... . ................ .. . 3.6.1.2-1 3.6 . 1.3 Primary Containment Isolation Valves (PCIVs) . . ..... ...... 3 . 6. 1. 3-1 3 . 6.1.4 Drywell and Suppression Chamber Pressure ... .. .. .. . ....... 3 . 6 . 1. 4-1 3 . 6 . 1.5 Drywell Air Temperature ...................... .. . .. . ... ... 3 . 6. 1.5-1 3.6. 1. 6 Suppression Chambe r- to-Drywe ll Vacu um Breakers .. ......... 3 . 6 .1. 6-1

3. 6 . 2 .1 Suppression Pool Ave r age Tem perature .. . .............. ... . 3.6.2.1-1 3 . 6 . 2. 2 Suppression Poo l Wate r Level . . . . ......................... 3.6.2 . 2-1 3 . 6. 2. 3 Residual Heat Remova l (RHR) Suppression Pool Cooling ..... 3.6 . 2. 3-1 3.6.2 . 4 Residual Heat Remova l (RHR) Suppression Pool Spray ....... 3 . 6 . 2. 4-1

3. 6. 3 . 1 Primary Containment Hydrogen Recombiners .............. ... 3.6.3.1-1 3 . 6. 3 . 2 Primary Containment Oxygen Concentration .. . .... .. ... . . ... 3.6.3.2-1

3. 6 . 4 . 1 Secondary Containment. . ...... . .. . .. ..................... . 3 . 6. 4 .1-1 3.6.4. 2 Secondary Containment Isolation Valves (SCIVs) . . . . . . . .... 3 . 6 . 4. 2-1 3.6.4.3 Sta ndby Gas Treatment (SGT) System ......... . ............. 3 . 6 .4. 3-1 (cont inu ed)

LaSalle 1 and 2 ii Amendment No .

r Definitions

1. 1 1.1 Definition s (continued)

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

CORE AL TE RAT ION CORE ALTERATION sha ll be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS :

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

and

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

Suspension of CORE ALTERATIONS shall not preclude comp l etion of movement of a component to a safe posit i on .

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle . These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie s/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present . The thyroid dose conversion factors used for this calculation shal l be those listed in Table III of TID-14844, AEC, 1962, "Cal cul ati on of Di stance Factors for Power and Test Reactor Sites ;" Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977; or ICRP (continued)

LaSalle 1 and 2 1. 1- 2 Amendment No .

Definitions

1. 1 1.1 Definitions DOSE EQUIVALENT I-131 30, Supplement to Part 1, pages 192-212, Table (continued) titled , "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity . "

ORA IN TI ME 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 i s div i ded by the l i miting 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 hu man error), for all penetrat i on flow paths be l ow the TAF excep t:

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 pat hs ;

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 iso l ation instrumentation; or (continued)

LaSalle 1 and 2 1.1-3 Amendment No .

Definitions

1. 1

1. 1 Definitions ORA IN TI ME 3. Penetration flow paths with isolation (continued) devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation devices wit hout offsite power .

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

d. No additional draining events occur ; and

e. Realistic cross-sectional areas and drain r ates are used .

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

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM ( ECCS) RESPONSE from when the monitored para meter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach the i r required values, etc.) . Times shall i nclude diese l generator sta r t i ng and seque nce loadi ng delays , where applicable . 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 . In lieu of measurement, response time may be verified for selected components provided t hat the co mponents and method for verification have been previously reviewed and approved by the NRC.

(continued)

LaSalle 1 and 2 1. 1-4 Amendment No .

Definitions

1. 1

1. 1 Definitions (continued)

END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial signal generation by (EOC-RPT) SYSTEM RESPONSE the associated turbine stop valve limit switch or TIME from when the turbine control valve hydraulic oil control oil pressure drops below the pressure switch setpoint to complete suppression of t he electr i c arc between t he fu ll y open co ntacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequentia l , overlapping, or total steps so that the entire response time is measured . In lieu of measurement, response time may be verified for selected components provided that the components and method for verificatio n have been previously reviewed and approved by the NRC .

INSERV ICE TESTI NG The I NSERVICE TEST I NG PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ISOLAT IO N SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be t hat RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the chan nel sensor until the iso l ation valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response ti me is measured . In lieu of measurement, response time may be verified for selected components provided that the components and method for verification have been previous l y reviewed and approved by the NRC.

(continued)

LaSalle 1 and 2 1. 1- 5 Amendment No .

Definitions

1. 1

1. 1 Definitions (continued)

LEAKAGE LEAKAGE sha 11 be:

a. Identified LEAKAGE

1. LEAKAGE into the drywel l 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. Uni dentified LEAKAGE A11 LEAKAGE into the drywel l that is not identified LEAKAGE; C. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and ct. Pressure Boundary LE AKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wal 1, or vessel wal 1.

LIN EAR HE AT GENERATION The LHGR shall be t he heat generatio n rate per RATE ( LHGR) unit l ength of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components requ i red 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 ser i es of sequential, overlapping, or total system steps so that the entire logic system is tested .

(continued)

LaSalle 1 and 2 1.1-6 Amendment No .

Definitions

1. 1 1.1 Definitions (cont inue d)

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

MO DE 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 ve sse l.

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, co ntrols, 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) .

RATED THERMAL POWER RTP shall be a total reactor core heat transfer

( RTP) rate to the reactor coolant of 3546 MWt .

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored para meter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids . 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 . In lieu of measurement, response time may be verified for selected components provided that the components and method for verification have been previously reviewed and approved by the NRC.

(continued)

LaSalle 1 and 2 1. 1-7 Amendment No.

Definitions 1.1

1. 1 Definitions (continued)

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~ 68°F, correspond i ng to the most reacti ve 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 .

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the test i ng of one of t he systems , subsystems ,

cha nn els, or ot her des i gnate d compo nents dur i ng the interva l specified by the Surveil l ance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels , or other designated components in the associated function .

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 TIM E sha l l be RESPONSE TIME that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions. The response time may be measured by means of any series of sequential ,

overlapping, or total steps so that the entire response time is measured .

LaSalle 1 and 2 1. 1-8 Amendment No .

Definitions

1. 1 Table 1 . 1-1 (page 1 of 1)

MODES REACTOR MODE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLAN T TEMPERATURE (OF) 1 Power Operation Run NA 2 Startup Refue1 <*> or Startup/Hot NA Standby 3 Hot Shutdown <*1 Shutdown > 200 4 Cold Shutdown <*1 Shutdown :,; 200 5 Refuel i ng <bJ Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bo lt s less than ful l y tensioned.

LaSalle 1 and 2 1. 1-9 Amendment No.

ECCS Instrumentation 3 . 3.5.1 ACT IONS CONDIT I ON REQUIRED ACTION COMPLETION TIME

8. As required by 8. 1 - - - - - - - - NOTE- - - - - - - -

Required Action A. 1 Only applicable for and referenced in Functions 1.a, 1. b, Table 3.3 . 5.1 -1. 2. a and 2. b.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months fro m feature(s) in opera ble discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable . feature(s) in both divisions 8.2 --------NOTE--------

Only appl i cable for Fu nctions 3.a an d 3 . b.

Declare High Pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Co re Spray ( HPCS) di sco very of Syste m inoperable . loss of HPCS initia ti on capability 8.3 Pl ace chan nel in 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s trip.

(continued)

LaSalle 1 and 2 3 . 3 . 5. 1-2 Amendment No.

ECCS Instrumentation 3 . 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.l - - - - - - - - NOTE- - - - - - - -

Req uired Action A.l Only app l icable for and referenced in Functions l . c and Table 3 . 3 . 5. 1-1. 2 . C.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inopera bl e disco very of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in bot h divis i ons C. 2 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPE RABLE status .

(continued)

LaSalle 1 and 2 3.3 . 5. 1-3 Amendment No.

ECCS In strumentati on 3.3.5.1 ACT IONS CONDIT ION REQUIRED ACTION COMPLETION TIME D. As required by D. l - - - - - - - - NOTE- - - - - - - -

Required Action A. l Only applicable for and referenced in Functions l . d , l.e, Table 3 . 3 . 5 . 1-1. l.f, l.g, 2. d, 2. e, and 2. f .

Dec l are s upp orted 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months fro m feat ure(s) in opera bl e disco very of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperab l e . feature(s) in both divisions D. 2 -------- NOTE---------

Onl y app l ic abl e fo r Fun cti ons l .d and 2.d .

Declare supported 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months fro m feature(s) disco very of inoperable . loss of initiation capability for feature(s) in one div is i on (continued)

LaSal l e 1 and 2 3.3 . 5.1 -4 Amendment No .

ECCS Instrumentation 3 . 3.5.1 Table 3 . 3.5.1-1 (page 1 of 4)

Emergency Core Cooling System In s trumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIF I ED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUN CT I ON CONDITIONS FUN CT I ON ACT I ON A. 1 REQUIREMEN TS VAL UE

1. Low Pre ss ure Coo l ant Injection-A (LPCJ) and Low Pressure Core Spray (LPCS)

Subsystems

a. Reactor Vesse l Water 1,2,3 B SR 3. 3. 5. 1. 1 ;,, -147.0 Level-Low Low Low, SR 3. 3. 5. 1. 2 i nches Level 1 SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5 SR 3. 3. 5. 1. 6

b. Drywell Pressure-High 1,2,3 21*> B SR 3 . 3.5.1 . 2 s 1. 77 psi g SR 3.3.5.1.4 SR 3. 3. 5. 1. 5 SR 3.3.5. 1. 6

c. LPC J Pump A 1,2,3 C SR 3 . 3.5 . 1.2 s 5.5 seconds Start-Time Delay SR 3.3.5.1.4 Relay SR 3. 3. 5. 1. 5

d. Reactor Stearn 1,2,3 2 D SR 3. 3. 5 . 1. 2 ;,, 490 psi g and Dorne Pressure-Low SR 3 . 3 . 5 . 1.4 s 522 ps i g (In ject i on Pe rmi ssive) SR 3.3.5. 1. 5 SR 3. 3. 5. 1. 6

e. LPCS Pump Discharge 1, 2 ,3 D SR 3. 3. 5. 1. 2 ;,, 1240 gprn and Flow-Low (Bypass) SR 3. 3. 5 . 1. 3 s 1835 gprn SR 3. 3. 5. 1. 5

f. LPC I Pump A Di sc harge 1,2,3 D SR 3 . 3 . 5 . 1. 2 ;,, 1330 gprn and Flow-Low (Bypass) SR 3.3 . 5.1.3 s 2144 gprn SR 3. 3. 5. 1. 5

g. LPCS and LPCI A 1,2,3 1 per valve D SR 3. 3 . 5. 1. 2 ;,, 490 ps ig and Injection Line SR 3.3.5.1.4 s 522 psig Pressure- Low SR 3 . 3.5.1.5 (I njection Permissive) SR 3.3.5.1.6

h. Manual Initiation 1,2,3 C SR 3.3.5.1.5 NA (continued)

(al Al so required to in i tiate t he associated diesel generator (DG).

LaSalle 1 and 2 3 . 3.5.1-9 Am endme nt No.

ECCS Instrumentation

3. 3. 5.1 Tabl e 3.3.5. 1* 1 (page 2 of 4)

Emergen cy Core Cooling System In st rumentation APPLICABLE CO NDITION S MODE S OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQU I RED SURVEILLANCE AL LOWABLE FUN CTI ON CONDITIONS FUNCTION ACTION A.1 REQUIREMENT S VALUE

2. LPCI B and LP CI C Subsys tems

a. Reactor Ve sse l Water 1,2, 3 21*> B SR 3.3 .5.1.1 ~ - 147.0 Leve l- Low Low Low, SR 3 . 3 .5.1.2 inche s Le vel 1 SR 3.3.5.1.4 SR 3 . 3. 5.1.5 SR 3.3.5.1.6

b. Drywel 1 Pre ss ure-High 1, 2 , 3 21*> B SR 3 . 3 .5.1. 2 s 1 . 77 ps ig SR 3.3.5. 1.4 SR 3.3.5.1.5 SR 3.3 .5.1.6 C. LPCI Pump B 1 , 2,3 C SR 3 . 3 .5.1. 2 s 5.5 second s Start-Time Delay SR 3.3 . 5.1.4 Relay SR 3 . 3 .5.1.5

d. Reactor Steam Dome 1, 2 , 3 2 D SR 3.3. 5.1.2 ~ 490 psi g and Pressure- Low SR 3.3 .5.1.4 s 522 ps ig (I njection Per missi ve) SR 3.3 .5.1.5 SR 3.3.5. 1. 6

e. LPCI Pump Band LPCJ 1, 2 ,3 1 per pump D SR 3 . 3.5 . 1. 2 ~ 1330 gpm and Pump C Di sc harge SR 3 . 3 .5.1 . 3 s 2144 gpm Flow- Low (Bypa ss) SR 3.3.5. 1.5

f. LPCI Band LPCI C 1, 2 , 3 1 per valve D SR 3.3.5. 1. 2 ~ 490 ps ig and Injection Line SR 3.3.5. 1.4 s 522 ps ig Press ure- Low SR 3 . 3 .5.1.5 (Injection Per mis s ive) SR 3 . 3.5. 1.6

g. Manual Initi ation 1, 2 , 3 C SR 3 . 3 . 5. 1.5 NA (co nt inued)

(a) Al so re qu ired to i nitiate t he assoc i ated DG.

LaSa l l e 1 and 2 3 . 3 . 5 . 1-10 Amendme nt No.

ECCS Instrumentation 3 . 3.5.1 Tab l e 3.3.5.1-1 (page 3 of 4)

Emergency Core Cooling System In str umentation APPLICABLE CONDITIONS MODES DR REFERENCED OTHER REQUIRED FROM SPECIF I ED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTl ON CONDITIONS FUN CT I ON ACTIONA.l REQUIREMENTS VALUE

3. Hig h Pre ss ure Core Spray (HPCS) System

a. Reactor Vessel Water 1 , 2 ,3 4<*) B SR 3. 3. 5. 1. 1 ~ -83 inche s Leve 1-Low Low , SR 3.3.5.1.2 Level 2 SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5 SR 3. 3. 5. 1. 6

b. Drywell Pressure-High 1, 2 ,3 4<*) B SR 3. 3. 5. 1. 2 s 1 . 77 ps i g SR 3. 3. 5. 1. 4 SR 3. 3. 5. 1. 5 SR 3.3.5.1.6

c. Reactor Vesse l Water 1,2,3 2 C SR 3.3.5.1. 1 s 66.5 in ches Level- Hig h, Level 8 SR 3.3.5.1.2 SR 3 . 3. 5. 1. 4 SR 3. 3. 5. 1. 5

d. HPCS Pump Di sc harge 1, 2 ,3 D SR 3.3.5.1.2 ~ 113.2 ps ig Pressure- Hig h SR 3. 3. 5. 1. 4

( Bypass) SR 3. 3. 5. 1. 5

e. HP CS System Flow 1,2,3 D SR 3. 3. 5. 1. 2 ~ 1380 gpm and Rate-Low (Bypass) SR 3. 3. 5. 1. 3 s 2194 gpm SR 3. 3. 5. 1. 5

f. Manual Initiation l, 2 ,3 C SR 3. 3. 5. 1. 5 NA

4. Automa ti c Depre ss urization System (AOS) Trip System A

a. Reactor Ve sse l Water 1, 2<b) , 3<b) 2 SR 3.3.5 . 1.1 ~ -147.0 Level - Low Low Low, SR 3 . 3.5.1.2 inches Level 1 SR 3.3.5.1.4 SR 3 . 3.5.1.5

b. Drywel l Pre ss ure-Hig h 1, 2<b) ,3 <b) 2 SR 3 . 3. 5 . 1. 2 s 1. 77 psi g SR 3. 3. 5. 1. 4 SR 3.3.5.1.5

c. ADS Initiation Timer 1, 2<b) ,3 <b) SR 3.3.5.1.2 s 118 seconds SR 3.3.5.1.4 SR 3.3.5.1.5 (cont inu ed)

(a) Al so required to initiate the associated DG.

(b) Wit h reactor s team dome pressure> 150 ps i g.

LaSal l e 1 and 2 3 . 3 . 5 .1 -11 Am end ment No.

ECCS In str umenta t i on 3 . 3 .5 . 1 Tab le 3 . 3 . 5 .1 - 1 ( pa ge 4 of 4 )

Eme rg ency Co re Coo lin g Sy s t em Ins trument at i on APPLI CAB LE CO NDITI ONS MODE S OR REFEREN CED OTHER REQ UI RED FROM SPEC IFIED CHANNE LS PER REQU I RED SURVEI LLAN CE ALLOWABLE FUN CT ION CONDITI ONS FUN CTION ACTION A.1 REQUIREMENT S VALUE

4. ADS Tr ip Sys t ern A (co ntinu ed)

d. Reacto r Vesse l Wat er 1 , 2101 . 31bl E SR 3 . 3.5 .1.1 ~ 11.0 in ches Leve 1- Low, Level 3 SR 3.3.5. 1. 2 (Conf i rmatory) SR 3.3.5 .1. 4 SR 3.3.5 .1.5

e. LPCS Pump Di sc harg e 1, 2101 , 3101 2 F SR 3.3.5. 1. 2 ~ 131 . 2 ps ig Press ure-Hi gh SR 3 . 3 . 5. 1.4 and SR 3 . 3.5 .1. 5 $ 271 . 0 ps i g

f. LPCI Pump A Di sc ha rg e 1, 2101 , 3101 2 F SR 3.3 . 5. 1. 2 ~ 10 5 .0 ps ig Pre ss ur e-High SR 3.3 . 5 .1.4 and SR 3.3 . 5 .1.5 s 128.6 ps ig

g. ADS Drywe 11 Press ur e 1, 2101 , 3101 2 F SR 3.3 .5.1. 2 s 598 sec ond s Bypa ss Ti me r SR 3 . 3 . 5. 1.4 SR 3.3.5. 1. 5

h. Manual Init i ati on 1, 2101 . 3101 2 F SR 3 . 3.5 .1.5 NA

5. ADS Tr i p Syst em B

a. Reactor Vess el Wat er 1 . 210) . 3<bl 2 E SR 3.3.5. 1.1 ~ - 147.0 Level - Low Low Low, SR 3.3.5. 1 . 2 in ches Level 1 SR 3. 3 .5. 1 . 4 SR 3 . 3 . 5 .1. 5

b. Drywe 11 Pres s ur e-Hi gh 1, 2101 . 3101 2 SR 3.3 . 5. 1. 2 $ 1. 77 ps ig SR 3.3.5. 1.4 SR 3.3 . 5. 1. 5 C. AD S In itia ti on Timer 1 , 2{b) , 3<bl F SR 3 . 3 . 5 .1. 2 $ 11 8 seco nd s SR 3 . 3.5 .1.4 SR 3 . 3 . 5 .1 . 5

d. Reactor Vesse l Wat er 1 , 2101 , 3101 SR 3.3.5 .1. 1 ~ 11.0 in ches Level - Low, Level 3 SR 3 . 3.5 .1. 2 (Confirmat ory) SR 3.3.5 .1.4 SR 3 . 3 . 5. 1. 5

e. LPCJ Pumps B & C 1 . 2101 . 3<b) 2 pe r pum p F SR 3.3 . 5.1. 2 ~ 105 .0 ps i g Di sc harge SR 3 . 3 . 5 .1.4 and Pressure-Hi gh SR 3 . 3 . 5 . 1. 5 $ 128 . 6 ps i g

f. ADS Drywe 11 Press ur e 1, 210) , 310) 2 F SR 3 . 3 . 5 .1. 2 :S 598 sec ond s Bypa ss Time r SR 3 . 3.5 .1 . 4 SR 3 . 3.5 .1. 5

g. Manual Init i ation 1. 2101 . 3101 2 F SR 3 . 3 . 5 . 1 . 5 NA

( b) Wit h reactor s t eam dome pre ss ure > 150 ps i g.

LaSa ll e 1 and 2 3 . 3 . 5. 1- 12 Am end men t No.

RPV Water Inventory Control Instrumentation 3 . 3.5.2 3.3 INSTRUMENTATION 3.3. 5. 2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5 . 2 The RPV Water Inventory Control instrumentation for each Function in Table 3 . 3. 5 . 2-1 shall be OPERABLE .

APPLICABILITY : According to Table 3 . 3 . 5. 2-1 .

ACTIONS

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NO TE- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

Separa t e Condit i on en t ry i s all owed for eac h chan nel.

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

B. As required by B. l Dec l are associated Immediately Required Action A. 1 penetration fl ow and referenced in path(s) incapable of Table 3 . 3 . 5 . 2-1. automatic isolation .

AND B. 2 Calculate DRAIN TIME . Immediately C. As required by C. l Place channel in trip . 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action A.1 and referenced in Tab l e 3 . 3 . 5. 2-1.

(continued)

LaSal l e 1 and 2 3.3.5.2-1 Amendment No .

RPV Water Inventory Control Instrumentation 3 . 3 . 5. 2 ACTIONS (continued)

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

and referenced in Table 3.3.5.2-1.

E. Required Action and E. 1 Declare associated Imm edia tel y associated ECCS injection/spray Completion Time of subsystem inoperable .

Cond it ion C or D not met.

SURVEI LLAN CE REQUIRE MENTS

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

Refer to Ta ble 3.3 .5. 2- 1 t o dete rmin e which SRs apply for each ECCS Fu nction .

SURVEILLANCE FREQUENCY SR 3.3 . 5 . 2.1 Perform CHANNEL CHECK . In accordance with the Surveillance Fre quen cy Co ntrol Program SR 3. 3 . 5.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Sur veillance Frequency Con trol Program LaSalle 1 and 2 3 . 3 . 5. 2-2 Am endment No.

RPV Water Inventory Control Instrumentation

3. 3. 5. 2 Table 3.3.5.2-1 (page 1 of 2)

RPV Water Inventory Control Instrumentation APPL! CABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECI FIED PER REQUIRED SURV EILLAN CE ALLOWABLE FUN CT ION CONDIT IONS FUN CT ION ACTION A.1 REQUIREMENTS VALUE

1. Low Pre ssure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCSJ Subsystems

a. Reactor Steam 4,5 1Cal C SR 3 . 3.5 . 2.2 s 522 psig Dome Pressure-Low

( Injection Permi ss ive)

b. LPCS Pump 4,5 1 per D SR 3.3.5.2.2 ?. 1240 gpm and Discharge pump Cal :S 1835 gpm Flow -Low (Bypass)

C. LPCI Pump A 4,5 1 per D SR 3. 3. 5.2.2 ?. 1330 gpm and Di sc harge pump c*> s 2144 gpm Flow-Low (Bypass)

d. LPCS and LPCI A 4,5 1 per C SR 3 . 3.5 . 2. 2 :S 522 psig Inje ction Line val ve c*>

Pressure-Low

( Injection Permissive)

2. LPCI B and LPCI C Subsystems

a. Reactor Steam 4,5 C SR 3.3.5.2.2 s 522 psig Dome Pressure-Low

( Inject ion Permissive)

b. LPCI Pump B 4,5 1 per D SR 3.3.5.2.2 ?. 1330 gpm and and LPCI Pump C pump c*> s 2144 gpm Discharge Flow -Low (Bypass)

c. LPCI Band LPCI C 4,5 1 per C SR 3. 3.5. 2. 2 s 522 psig Injection Line val ve c*>

Pressure-Low (Injection Permissive)

(continued)

(a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5. 2, "RPV Water Inventory Control .

La Sal le 1 and 2 3.3.5.2-3 Amendment No .

RPV Water Inventory Control Instrumentation 3 . 3.5.2 Tab1e 3.3.5.2-1 (page 2 of 2)

RPV Water Inventory Control Instrumentation APPL! CABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNE LS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUN CT ION ACTIONA.l REQUIREMENTS VALUE

3. High Pressure Core Spray CHPCS) System

a. HP CS Pump 4, 5 1Ca) D SR 3.3.5.2.2 ~ 11 3 . 2 ps i g Di scharge Pressure-High (Bypass)

b. HPCS System Flow 4, 5 1( a) D SR 3.3.5.2.2 ~ 1380 gpm Rate-Low (Bypass) and

$ 2194 gpm

4. RHR Shutdown Cooling System Isolation a . React or Ve ssel (b) 2 in one B SR 3.3.5.2. 1 ~ 11. 0 Water Level-Low, trip SR 3. 3 . 5.2.2 i nc hes Level 3 system

5. Reactor Water Cleanup (RWCU) System Isolation

a. Reactor Vessel ( b) 2 in one B SR 3.3.5.2.2 ~ -58.0 Water Level-Low trip inches Low, Level 2 system (a) Assoc i ated with an ECCS sub system required to be OPERABLE by LCO 3.5.2, "RPV Water Inventory Control . "

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

LaSalle 1 and 2 3 . 3. 5. 2-4 Amendment No.

RCIC System Instrumentation 3 . 3 .5 . 3 3.3 INSTRUMENTATION 3 . 3.5 . 3 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5 . 3 The RCIC System instrumentation for each Function in Table 3 . 3 . 5.3-1 shall be OPERABLE.

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

ACT IONS

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

Separate Condit i on entry is al l owed for each cha nnel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A. 1 Enter the Condition Immed i ately inoperable. referenced in Table 3 . 3 . 5. 3-1 for the channel.

B. As required by B. 1 Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5 . 3-1. initiation capability AND B.2 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip .

(continued)

LaSalle 1 and 2 3 . 3 . 5. 3-1 Amendment No .

RCIC System Instrumentation 3 . 3 . 5. 3 ACT IONS CONDIT ION REQUIRED ACT ION COMPL ETION TIME C. As required by C.l Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.l OPERABLE status .

and referenced in Table 3 . 3.5. 3-1.

D. As required by 0. 1 ---- ----NOTE---------

Required Action A. l On l y applicable if and referenced in RCIC pump suction is Table 3 . 3 . 5.3-1. not aligned to the suppression pool .

Declare RCIC Syste m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable . discovery of loss of RCIC initiation capability

0. 2. 1 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

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

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

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

LaSalle 1 and 2 3 . 3.5.3-2 Amendment No.

RCIC System Instrumentation 3.3.5.3 SURVEILLANCE REQUIREMENTS

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

1 . Refer to Table 3.3 .5. 3-1 to determine which SRs apply for each RCIC Function .

2. When a channel is placed in an in operable status sole l y for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 4; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.

SURVEILLANCE FREQUENCY SR 3 . 3 . 5.3.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3 . 3 . 5 . 3.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Su rv eillance Frequency Control Program LaSalle 1 and 2 3.3.5.3-3 Amendment No.

RCIC System Instrumentation 3.3 . 5.3 Table 3 .3.5 . 3* 1 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation CONO IT IONS REQUIRED REFERENCED FROM CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION FUNCTION ACT ION A. 1 REQUIREMENTS VALUE

1. Reactor Ve sse l Water 4 B SR 3 .3.5. 3.2 ~ -B3 inches Level-Low Low, Level 2 SR 3.3.5.3.3 SR 3.3.5.3.4

2. Reactor Ves se l Water 2 C SR 3.3.5.3.1 ,;; 66.5 inches Level-High, Leve l 8 SR 3.3 . 5.3.2 SR 3.3.5.3.3 SR 3.3 . 5.3.4

3. Condensate Storage Tank 2 0 SR 3.3.5.3.2 ~ 713.6 ft Leve 1-Low SR 3.3.5.3.3 SR 3.3.5.3.4

4. Manual Initiation C SR 3.3.5.3.4 NA LaSalle 1 and 2 3 . 3.5.3-4 Am endment No.

Primary Contain ment Isolation Instrumentation 3.3.6.1 Table 3 . 3.6.1- 1 (page 4 of 4)

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

4. RWCU System Iso lation (continued)

k. Reactor Ve ssel Water 1, 2, 3 2 SR 3.3 .6.1. 2 ~ - 58.0 inche s Leve 1-Low Low, SR 3. 3 . 6. 1. 4 Leve 1 2 SR 3.3.6.1.5

1. Standby Liquid 1, 2,3 2 (b) SR 3.3.6.1.5 NA Control Sys tem Init iation

m. Manual Initiation 1,2, 3 G SR 3.3 . 6.1.5 NA

5. RHR Shutdown Cooling System I so l ation

a. Reactor Ve sse l Water 3 2 J SR 3 . 3.6. 1.1 ~ 11.0 inche s Level-Low, Level 3 SR 3 .3.6 . 1. 2 SR 3.3.6.1.4 SR 3.3.6.1.5

b. Reactor Vessel 1, 2 , 3 SR 3.3.6.1.2 s 143 ps i g Pre ss ure- High SR 3.3.6.1.4 SR 3.3.6.1.5

c. Manual Initiation 1, 2 , 3 G SR 3.3.6.1.5 NA

( b) Only input s into one of two trip systems.

LaSalle 1 and 2 3 . 3 . 6. 1-9 Amendment No.

Secondary Containment Isolation Instrumentation 3 . 3.6.2 Table 3.3.6.2 - 1 (page 1 of ll Secondary Containment Isolation Instru mentation APPLICABLE MODES AND REQUIRED OTHER CHANNELS SPECIF IED PER TRIP SURVE ILLANCE ALLOWABLE FUNCTION CONDIT IONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vesse l Water 1, 2,3 2 SR 3.3.6.2.2 ~ -58.0 in ches Level-Low Low, Level 2 SR 3. 3 . 6.2.3 SR 3.3.6 . 2.4

2. Drywe 11 Pressure-High 1,2,3 2 SR 3.3.6.2.2 s; 1.93 psig SR 3.3 .6 . 2.3 SR 3.3.6.2. 4

3. Reactor Building 1, 2, 3, Col 2 SR 3.3.6 . 2.1 s; 42.0 mR /hr Ventilation Exhaust Plenum SR 3.3.6 . 2.2 Radiation-High SR 3.3 . 6 . 2.3 SR 3.3.6.2.4

4. Fue l Pool Ventilation 1, 2 ,3, Co) 2 SR 3.3.6.2. 1 ,; 42.0 mR/hr Exhaust Radiat ion- High SR 3.3.6 .2.2 SR 3.3.6.2.3 SR 3.3 .6.2.4

5. Manual Initi ation 1,2 ,3 , <*l SR 3.3.6.2. 4 NA (a) Dur i ng CORE ALTERATIONS, and during movement of irradiated fue l assemblies in the secondary contai nment.

LaSalle 1 and 2 3 . 3.6.2-4 Amendment No .

CRAF System Instrumentation 3 . 3 . 7. 1 3.3 INSTRUMENTATION 3 . 3 . 7.1 Control Room Area Filtration (CRAF) System Instrumentation LCO 3.3 . 7. 1 Two channels per trip system for the Control Room Air Intake Radiation-High Function sha ll be OPERABLE for each CRAF subsystem .

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

ACTIONS

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

Separate Condition entry is allowed for each channel .

CONDITION REQUIRED ACT ION COMPLETION TIME A. One or more channels A. 1 Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable . CRA F subsystem disco very of inoperable . loss of CRAF subsystem initiation capability A. 2 Pl ace channel in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months trip.

(continued)

La Salle 1 and 2 3.3.7.1-1 Amendment No .

LOP In strume ntation

3. 3 . 8 . 1 Ta ble 3 . 3 . 8 .1 - 1 ( page 1 of 1)

Loss of Power Ins t r ume nta ti on REQUIRED CHANNELS PER SURVEILLAN CE ALLOWABLE FUNCTION DIV IS JON REQUIREMENT S VA LUE

1. Di visio ns 1, 2 and Oppos i te Uni t Divi s ion 2 - 4 . 16 kV Emerg ency Bu s Underv oltage

a. Loss of Voltag e - 4 . 16 kV 2 SR 3 . 3.8 .1. 3 :?: 2870 V and $ 3127 V Ba s i s SR 3 . 3.8 .1.4 SR 3 .3. 8 .1.5

b. Lo ss of Voltag e - Time Delay 2 SR 3 . 3 . 8 . 1 . 3 ;,: 3 .1 second s and ,; 10. 9 second s SR 3 . 3.8 .1.4 SR 3 . 3. 8 .1.5

c. Degr aded Volt age - 4 . 16 kV 2 SR 3 . 3 . 8.1. 1 :?: 3814 V and s 3900 V Bas i s SR 3 . 3.8 .1. 2 SR 3.3 .8.1. 5

d. Degr aded Vo l t age - Time 2 SR 3.3 . 8 .1.1 ;,: 27 0.1 seco nd s and Delay, No LOCA SR 3.3 . 8. 1. 2 s 329 . 9 seco nd s SR 3 .3. 8 .1 . 5

e. Degrad ed Volt age - Time 2(a)(b) SR 3 . 3.8 .1.1 :?: 9 .4 s econd s and s 10. 9 sec ond s Delay , LOCA SR 3 . 3 . 8. 1. 2 SR 3.3.8. 1.5

2. Divi s ion 3 - 4 . 16 kV Eme rgen cy Bu s Und er volt age

a. Loss of Voltag e - 4.1 6 kV 2 SR 3 . 3 . 8. 1. 3 ;,: 272 5 V and $ 3172 V Bas i s SR 3 . 3.8. 1.4 SR 3 . 3 . 8. 1. 5

b. Loss of Voltag e - Time Del ay 2 SR 3 . 3 . 8 .1. 3 s 10.9 se cond s SR 3 . 3 . 8 .1.4 SR 3 . 3 . 8 . 1 . 5

c. Degrad ed Voltag e - 4 . 16 kV 2 SR 3 . 3 . 8 .1.1 ;,: 3814 V and $ 39 00 V Ba s i s SR 3 . 3 . 8 .1. 2 SR 3.3.8. 1.5

d. Degr aded Voltag e - Time 2 SR 3 . 3 . 8 .1.1 ;>: 27 0 .1 second s and Delay , No LOCA SR 3 . 3 . 8 .1. 2 s 329. 9 seco nd s SR 3.3.8 .1. 5

e. Degrad ed Voltag e - Time 2(a)(b) SR 3.3.8. 1. 1 ;,: 9 . 4 second s and ,; 10. 9 sec ond s De l ay , LOCA SR 3 . 3 . 8 .1. 2 SR 3 . 3 . 8 .1. 5 (a) In MODE S 4 and 5 , not requ i red to be OPERABLE.

(bl Wit h no fu e l i n t he r eactor vess el, not required to be OPERAB LE.

LaSa l le 1 and 2 3 . 3. 8 . 1-3 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, 2, and 3, MODES 4 and 5 with residual heat remova l (RHR) s hutd own cooling (SOC) isolation valves open, MODE 5, with any control rod withdrawn from a core cell containing one or more fuel assemblies, During movement of irradiated fuel assemblies in the secondary contain ment, During CORE ALTERATIONS.

ACTIONS CONDITION REQUIRED ACT ION CO MPLETION TIME A. One or both inservice A. 1 Remove associated 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months power supplies with in service power one electric power su pp ly(s) from moni toring assembly ser vice.

inoperable .

B. One or both inservice B. l Remove associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months power suppl i es with in service power both electric power supp l y(s) from monitoring assemblies service.

inoperable.

(continued)

LaSalle 1 and 2 3.3 . 8.2-1 Amendment No.

RPS Elec t ric Power Monitoring 3.3.8.2 ACT IONS COND IT ION REQUIRED ACTION COMPLETION TI ME F. Required Action and F. 1. 1 Isolate the Immediately associated Completion associated secondary Time of Cond i tion A or containment B not met during penetration fl ow movement of i rradiated pat h(s).

fuel asse mbl ies in the secondary containme nt or during CORE ALTERATIONS . F. 1. 2 Declare the Immediately associated seconda ry containment i solation va l ve(s) inoperable .

F. 2. 1 Pl ace the associa t ed Immediately standby gas treatme nt (SGT) subsystem(s) i n operation .

F.2.2 Dec l are associated Immediately SGT subsystem(s) i noperab l e .

LaSalle 1 and 2 3.3 . 8 . 2-3 Amendment No .

ECCS-Operating 3.5 . 1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

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

SYSTEM 3.5.1 ECCS-Operating LCO 3.5.1 Each ECCS injection/ spray subsystem and the Automati c Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICA BILITY : MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure~ 150 psig.

ACT IONS

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

LCO 3.0.4.b is not applicable to HPCS .

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.l Restore low pressure 7 days injection/spray ECCS injection/spray subsystem inoperable . subsystem to OPERAB LE status .

(continued)

LaSalle 1 and 2 3. 5.1-1 Amendment No .

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

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

SYSTEM 3.5.2 RPV Water Inventory Control LCO 3. 5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

One ECCS injection/spray subsystem shal l be OPERABLE.

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

A Low Pressure Coolant Injection (LPCI) subsystem may be considered OPERAB LE dur i ng al ign ment an d operatio n for decay heat removal, if capable of being manually realigned and not otherwise inoperable.

APPLIC ABILITY : MODES 4 and 5.

ACT IONS CONDITION REQ UIRED ACTION COMPLE TION TI ME A. Required ECCS A. 1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperable. subsystem to OPERABLE status .

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

(continued)

LaSalle 1 and 2 3.5.2-1 Amendment No.

RPV Water Inventory Control 3 . 5. 2 ACTIONS (continued)

CONDIT ION REQUIRED ACT ION 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 2: 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 .

C. 2 Ver ify eac h secondary 4 hours containment penetration flow path is capable of being i solated in l ess t han t he DRAI N TIME .

C. 3 Verify one standby 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months gas t reat ment s ubsystem i s capab l e of being placed in operation in less than the DRAIN TIME .

(co ntinu ed)

LaSalle 1 and 2 3. 5. 2-2 Amendment No.

RPV Water Inventory Control

3. 5. 2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. DRAIN TIME < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 0.1 - - - - - - - NOTE - - - - - - - -

Required ECCS injection/spray subsystem or additional method of water injection sh all be capable of ope r ating wit hout offsite electrical power .

Initiate action to Immediately establ i sh an additional method of water injection with water sources capable of maintaining RPV water leve l > TAF f or

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

0. 2 In i tiate action to Immediately establish secondary containment boundary.

D.3 Initiate act i on to Immediately iso l ate each secondary containment penetration flow path or verify it can be manually isolated fro m the control roo m.

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

(continued)

LaSalle 1 and 2 3 .5.2 -3 Amendment No .

RPV Water Inventory Control 3 .5.2 ACTIONS (continued)

CONDITION REQUIRED ACT ION COMPLETION TIME E. Required Action and E.l Initiate action to Immediately associated Completion restore DRA I N TIME to Time of Condition C or ~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D not met.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3 . 5 . 2.1 Verify DRAIN TIME~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In accordance with the Surveillance Frequency Control Program .

SR 3 . 5. 2 . 2 Verify, for a required low pressure ECCS In accordance injection/spray subsystem, the suppression with the pool water level is~ -12 ft 7 in. Surveillance Frequency Control Program SR 3 . 5 . 2.3 Verify, for a required High Pressure Core In accordance Spray (HPCS) System, the suppression pool with the water level is~ -12 ft 7 in . Sur veillance Frequency Control Program SR 3 . 5.2.4 Verify, for the required ECCS In accordance injection/spray subsystem, locations with the susceptible to gas accumulation are Surveillance sufficiently fi l led with water. Frequency Control Program (continued)

LaSalle 1 and 2 3. 5.2-4 Amendment No .

RPV Water Inventory Control 3.5 . 2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2 . 5 - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - -

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

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

SR 3. 5. 2.6 Operate t he required ECCS injection/spray In accordance subsystem through the recirculation line for with the

,

10 minutes. Surveillance Frequency Control Program SR 3.5 . 2.7 Verify each valve credited for automatically In accordance isolating a penetration flow path actuates with the to the isolation position on an actual or Surveillance simulated isolation sig nal. Frequency Contro l Program SR 3 . 5.2.8 - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -

Vessel injection/spray may be excluded .

Verify the required ECCS injection/spray In accordance subsystem can be manually operated. with the Surveillance Frequency Control Program LaSalle 1 and 2 3.5.2-5 Amendment No.

RCIC System

3. 5.3

3. 5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

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

SYSTEM 3.5.3 RCI C System LCO 3. 5. 3 The RCIC System shall be OPERABLE.

APPLIC ABIL ITY: MODE 1, MODES 2 and 3 wit h reactor stea m dome pressure > 150 psig.

ACTIONS

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

LCO 3. 0. 4 . b is not applicable to RCIC .

CONDITION REQUIRED ACT ION COM PL ETION TIME A. RCIC System A. 1 Verify by Immediately inoperable . administrative means High Pressure Core Spray Syste m is OPERAB LE.

AND A. 2 Restore RCIC System 14 days 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 not met . A.till.

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 .

LaSalle 1 and 2 3. 5.3-1 Amendment No .

PCIVs 3.6.1.3 3 . 6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves ( PCIVs)

LCO 3. 6 . 1. 3 Each PCIV sh all be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTIONS NOTES - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1. Penetration flow paths may be unisolated intermittently under ad ministrati ve controls .

2. Separate Condition entry is al l owed for each penetration flow path .

3. Enter applicable Conditions and Required Action s for systems made inoperable by PCIVs.

4. En t er appl i cable Con ditions and Required Act i ons of LC O 3 . 6 . 1 .1, "Pri mary Containment ," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.

CONDIT ION REQUIRED ACT ION COMPLETION TIME A. ---------NOTE--------- A. l Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except Only applicable to penetration flow path for main steam pe netratio n flow pat hs by use of at least line wit h two or more one closed and PCIVs . de-activated automatic valve, closed manual valve , 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main One or more blind flange , or stea m line pe netration flow paths check valve with f l ow wit h one PCIV through the valve inoperable for reasons secured.

other than Co ndition D. AND (continued)

LaSalle 1 and 2 3 . 6. 1.3-1 Amendment No .

PC IVs 3.6.1.3 ACT IONS CONDITION REQUIRED ACT ION COMPLETION TIME E. Required Action and E. 1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, C, or D not met in MODE 1, 2, or 3. E. 2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months LaSalle 1 and 2 3.6.1.3-5 Amendment No .

Secondary Conta i nment 3.6.4 . 1 3.6 CONTAINMENT SYSTEMS

3. 6. 4 . 1 Secondary Conta i nment LCO 3.6 . 4 . 1 The secondary containment sha l l be OPERABLE .

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel asse mblies in the secondary contain ment, During CORE ALTERATIONS .

ACTIONS COND ITI ON REQUIRED ACTION CO MPL ETION TIM E A. Secondary contain ment A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to 2' or 3 . OPERABLE status .

B. Required Action and B.1 Be in MODE 3 . 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Comp l et i on Ti me of Cond i tion A not met .

(continued)

LaSalle 1 and 2 3 . 6. 4. 1-1 Amendment No .

Secondary Containment 3.6.4.1 ACT IONS CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment C. l - - - - - - - - NOTE- - - - - - - - -

inoperable during LCO 3. 0. 3 is not movement of i rradiated applicable.

fuel assemblies in the secondary containment or during CORE Suspend movement of Immediately ALT ERA TI ON S. irradiated fuel assemblies in the secondary containment .

C. 2 Suspend CORE Immediately AL TERA TI ONS.

LaSalle 1 and 2 3. 6. 4 . 1-2 Amendment No .

SCIVs 3.6 . 4. 2 3.6 CONTAINMENT SYSTEMS 3 . 6.4 .2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6 . 4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary contain ment, During CORE ALTERATIONS .

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 appl i cable Conditions and Required Act i ons for systems made inoperable by SCIVs.

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

(continued)

LaSalle 1 and 2 3.6 .4. 2-1 Amendm ent No.

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.l - - - - - - - - NOTE- - - - - - - - -

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

or B not met during movement of irradiated fue l asse mbli es in t he Suspend mo vement of Immediately secondary containment irradiated fuel or during CORE assemblies in the ALTE RAT ION S. secondary containment .

D. 2 Suspend CORE Immediately ALTE RATIO NS.

LaSalle 1 and 2 3.6 . 4. 2-3 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 Tw o SGT subsystems sha ll be OPERABLE .

APPLICAB I LITY : MODES 1, 2, and 3, During mo veme nt of irrad i ated fu el asse mbl ies i n th e secondary containment, During CORE ALTERATIONS .

ACT IONS CONDITION REQUIRED ACT ION COMPLETION TIME A. One SGT subsystem A. l Restore SGT subsystem 7 days ino perable . to OPERABLE status .

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

C. Required Action and - - - - - - - - - - - - NOTE- - - - - - - - - - - - -

associated Completion LCO 3 . 0. 3 i s not applicab l e .

Ti me of Condition A no t met duri ng mo vement of irradiated C. l Pla ce OPERABLE SGT Immed i ately fuel assemblies in the subsystem in secondary containment operation .

or during CORE ALTERATIONS . Q.R (continued)

LaSalle 1 and 2 3.6.4.3-1 Amendment No.

SGT System 3.6 . 4. 3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 . 1 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment.

C. 2.2 Suspe nd CO RE Im medi at ely AL TE RATIONS .

D. Two SGT subsystems 0. 1 Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i noperable in MODE 1 ,

2, or 3 .

E. Two SGT s ubsystems E.1 - - - - - - - -NO TE - - - - - - - - -

i nopera bl e du r i ng LCO 3 . 0. 3 is not mov ement of i rrad i at ed ap plicable .

fuel assemblies in the secondary containment, or during CORE Suspend movement of Immediately ALT ERATIONS. irradiated fuel asse mblies in the secondary containment .

E. 2 Suspend CO RE Imm edi ate l y ALTERATIONS .

LaSal l e 1 and 2 3 . 6.4 . 3-2 Am endment No.

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6 . 4 . 3.1 Operate each SGT subsystem for In accordance

~ 15 continuous minutes with heaters with the operating. Surve illan ce Frequency Contro 1 Program SR 3 . 6 . 4.3.2 Perform required SGT fi l ter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP) .

SR 3.6.4.3.3 Ver i fy each SGT subsystem actuates on an In accordance actual or si mul ated initiation signal. wit h the Surveillance Frequency Contro 1 Program LaSalle 1 and 2 3 . 6.4.3-3 Amendment No.

CRAF System 3 . 7. 4 3 .7 PLANT SYSTEMS 3.7.4 Control Room Area Filtrati on (CRA F) Syste m LCO 3 .7 . 4 Two CRAF subsystems shall be OPERABLE .

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

The control room enve l ope (CRE) boundary may be opened intermittently under administrative control .

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRAF subsystem A.1 Restore CRAF 7 day s inoperable for reasons subsystem to OPERABLE ot her than Condition status.

B.

B. One or more CRA F B. 1 Initiate action to Immediately subsystems inoperable implement mitigating due to inoperable CRE actions.

boundary in MODE 1, 2' or 3 . AND B. 2 Verify mitigating 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months actions ensure CRE occupant exposures to radiologica l ,

chemical , and smoke hazards wil l not exceed limits .

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

(conti nued)

LaSalle 1 and 2 3 . 7.4 -1 Amendment No.

CRAF System 3 . 7. 4 ACT IONS CONDITION REQUIRED ACT ION COMPLETION TIME C. Required Action and C. 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 or B not met in MODE 1, 2 , or 3 .

D. Requ ired Act i on and - - - - - - - - - - - -NOTE- - - - - - - - - - - - -

associated Complet i on LCO 3.0 . 3 i s not app l icable .

Time of Condition A not met during movement of irradiated D. 1 Pl ace OP ERA BL E CRA F Immediately fu el asse mbl ies in the s ub system i n seco ndary co ntain men t pressur i zat i on mo de.

or during COR E ALT ERATIONS .

D. 2 . 1 Suspend movement of Immediately irr adiated f uel assemblies i n the secondary containment.

D.2.2 Suspend CORE Immed i ately AL TE RATIONS .

E. Two CRAF subsystems E. 1 Be i n MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ino perable i n MO DE 1, 2, or 3 for reasons other than Condition B.

(contin ued)

LaSalle 1 and 2 3.7.4-2 Amendment No .

CRAF System 3.7.4 ACT IONS CONDITION REQUIRED ACT ION COMPLETION TIME F. Two CRAF subsystems - - - - - - - - - - - -NOTE - - - - - - - - - - - - -

inoperable during LCO 3. 0.3 is not applicable .

movement of irradiated fuel assemblies in the secondary contai nment F. 1 Sus pend mo vement of Imm edi ate l y or during CORE irradiated fuel ALTERATIONS. assemblies in the secondary containment .

One or more CRAF AND subsystems inoperable due to inoperable CRE F. 2 Suspend CORE Immediate l y bou ndary dur ing ALTE RATIO NS.

movement of irradiated fuel assemblies in the secondary containment or during CORE ALT ERAT IO NS.

SURVEI LLANCE REQUI REME NT S SURVEILLANCE FREQUENCY SR 3.7.4 . 1 Operate each CRAF subsystem for In accordance

~ 15 continuous minutes with the heaters with the operat i ng . Surveillance Frequency Control Program (continued)

LaSalle 1 and 2 3. 7. 4-3 Amendment No .

Control Room Area Ventilation AC System 3.7.5

3. 7 PLANT SYSTEMS

3. 7. 5 Control Room Area Ventilation Air Condition ing (AC) Syste m LCO 3. 7.5 Two control room area ventilation AC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irrad i ated fuel assemblies i n the secondary containment, During CORE ALTERATIONS .

ACTIONS COND ITION REQUIRED ACTION COMPLETION TIME A. One control room area A.l Restore control room 30 days ventilation AC area ventilation AC sub system in operab l e . s ub system to OPERABLE status .

B. Two control room area B.l Verify control room Once per 4 ventilation AC area temperature hours subsystems inoperable. < 90°F .

ANO B.2 Restore one control 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months room area ventilation AC subsyste m to OPERABLE status .

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

(continued)

LaSalle 1 and 2 3. 7. 5-1 Amendment No.

Control Room Area Ventilation AC System 3 . 7. 5 ACTIONS COND IT ION 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 irradiated D.1 Pl ace OPERABLE Immediately fuel assemblies in the control room area secondary containment ventila tion AC or during CORE subsystem in ALTERATIONS. operation .

D. 2.1 Suspend movement of Immediately irradiated fuel assemblies in the secon da ry containment .

D. 2. 2 Suspend CO RE Immediately AL TERA TI ONS.

(continued)

La Salle 1 and 2 3.7 .5 -2 Amendment No .

Control Room Area Ventilation AC System 3.7.5 ACT IONS CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and - - - - - - - - - - - -NOTE - - - - - - - - - - - - -

associated Completion LCO 3. 0. 3 is not applicable.

Time of Condition B not met during movement of irradiated E. l Suspend movement of Immediately fuel assemblies in the irradiated fuel secondary containment assemblies in the or during CORE secondary ALTERATIONS . containment .

E. 2 Suspend CORE Immediately ALTERATIONS .

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3 . 7.5.1 Monitor control room and auxiliary electric In accordance equipment room temperatures . with the Surveillance Frequency Control Program SR 3. 7. 5. 2 Verify correct breaker al ignment and In accordance indicated power are available to the with the control room area ventilation AC Surveillance subsystems. Frequency Control Program LaSalle 1 and 2 3. 7. 5-3 Amendment No.

AC Sources- Shutdown

3. 8 . 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A. 2.3 Initiate action to Immediately restore required offsite power circuit to OPERAB LE status .

B. Required DG of LCO B. 1 Suspend CORE Immediately Item b. in operab l e . ALTE RAT IO NS .

AND B. 2 Suspend mo vement of Immediately irradiated fuel asse mblies in secondary containment.

AND B. 3 Initiate ac ti on to Immediately restore required DG to OPERABLE status.

C. Required DG of LCO C. l Declare Hig h Pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Item c . inoperable . Core Spray System inoperable.

(continued)

LaSalle 1 and 2 3. 8 . 2-3 Amendment No.

AC Sources-Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMP LETION TIME D. Required offsite 0.1 Declare associated Immediately ci rcuit or DG of LCO standby gas treatment Item d. inoperable . subsystem, control room area filtration subsystem , and control room area ventilation air co nditioning subsystem inoperab l e.

SURVEI LLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3 . 8 . 2.1 - - - - - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - - - - - -

1. The following SRs are not required to be performed: SR 3 .8 . 1 . 3, SR 3 . 8.1 . 9 through SR 3.8.1.11, SR 3.8.1 .1 3 through SR 3 .8.1 . 16, SR 3 . 8 . 1. 18, and SR 3 .8 . 1.19.

2. SR 3 .8.1 . 12 and SR 3. 8.1 . 19 are not required to be met .

For AC sources required to be OPERABLE, the In accordance SRs of Specification 3 . 8 . 1, except with applicable SR 3 . 8 . 1. 8, SR 3. 8 . 1. 17, and SR 3 . 8 . 1.20, SRs are applicable .

La Salle 1 and 2 3. 8.2-4 Amendment No.

DC Sources-Shutdown 3.8 . 5 ACT IONS CONDIT ION REQUIRED ACTION COMPLETION TIME B. One or more required B. l Declare affected Immediately DC electrical power requ i red feature(s) subsystems inoperable inoperable .

for reasons other than Condition A. .Q.R B. 2 . 1 Suspend CORE Immediately AL TERA TI ONS.

Required Action and Completion Time of Condition A not met .

B.2.2 Suspend mo vement of Immediately irradiated fuel assemblies in the secondary containment.

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

LaSalle 1 and 2 3.8.5-3 Amendment No .

Distribution Systems-Shutdown 3.8.8 ACT IONS CONDIT ION REQUIRED ACT ION COMPLETION TIME A. (continued) A. 2. 3 Initiate action s to Immediatel y restore requ ir ed AC and DC electri ca l power distribution s ub systems to OPERABLE status .

A.2.4 Declare associated Immed iat el y required sh utd own cooling subsystem(s) inoperable and not in operation .

SURVEILLAN CE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3 . 8.8.1 Verify correct breaker al ignments and In accordance voltage to required AC and DC el ectrical with the power distribution subsystems. Surveillance Frequency Con tr ol Prog r am LaSalle 1 and 2 3 . 8 . 8-2 Amendment No .

LaSalle County Station, Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (MARK-UP)

TOCi 8 3.3.5.3-12 8 3.6.4.3-3 TOCii 8 3.3.6.1-7 8 3.6.4.3-4 8 3.3.5.1-8 8 3.3.6.1-28 8 3.6.4.3-5 8 3.3.5.1-10 8 3.3.6.2-4 8 3.7.4-4 8 3.3.5.1-11 B 3.3.6.2-6 8 3.7.4-5 8 3.3.5.1-12 8 3.3.6.2-7 8 3.7.4-6 8 3.3.5.1-13 8 3.3.7.1-3 8 3.7.4-7 8 3.3.5.1-14 8 3.3.8.1-6 8 3.7.4-8 8 3.3.5.1-15 8 3.3.8.2-4 8 3.7.5-3 8 3.3.5.1-16 8 3.3.8.2-6 8 3.7.5-5 8 3.3.5.1-17 83.5.1-1 8 3.7.5-6 8 3.3.5.1-18 8 3.5.1-6 8 3.8.2-1 8 3.3.5.1-25 8 3.5.2-1 8 3.8.2-3 8 3.3.5.1-27 8 3.5.2-2 8 3.8.2-4 8 3.3.5.1-29 8 3.5.2-3 8 3.8.2-5 8 3.3.5.1-36 8 3.5.2-4 8 3.8.2-6 8 3.3.5.2-1 8 3.5.2-5 8 3.8.2-7 8 3.3.5.2-2 8 3.5.2-6 8 3.8.2-8 8 3.3.5.2-3 8 3.5.2-7 8 3.8.5-1 8 3.3.5.2-4 8 3.5.2-8 8 3.8.5-3 8 3.3.5.2-5 8 3.5.2-9 8 3.8.5-7 8 3.3.5.2-6 8 3.5.2-10 8 3.8.8-1 8 3.3.5.2-7 8 3.5.2-11 8 3.8.8-2 8 3.3.5.2-8 8 3.5.2-12 8 3.8.8-3 8 3.3.5.2-9 8 3.5.2-13 8 3.8.8-4 8 3.3.5.2-10 8 3.5.3-1 8 3.10.8-3 8 3.3.5.3-1 8 3.5.3-2 8 3.3.5.3-2 8 3.5.3-7 8 3.3.5.3-3 8 3.6.1.3-3 8 3.3.5.3-4 8 3.6.1.3-9 8 3.3.5.3-5 8 3.6.1.3-10 8 3.3.5.3-6 8 3.6.2.2-2 8 3.3.5.3-7 8 3.6.4.1-2 8 3.3.5.3-8 B 3.6.4.1-3 8 3.3.5.3-9 8 3.6.4.1-4 8 3.3.5.3-10 8 3.6.4.2-2 8 3.3.5.3-11 8 3.6.4.2-5

TABLE OF CONTENTS B 2.0 SAFETY LIMITS (SLs)

B 2. 1 . 1 Reactor Core SLs ....... ... . ...... .. .. .. . ............ B 2. 1.1-1 B 2 . 1. 2 Reactor Coolant System (RCS) Pressure SL ........... B 2. 1.2-1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ... B 3 . 0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ........ . ... B 3.0-11 B 3.1 REACTIVITY CONTROL SYSTEMS B 3 . 1.1 SHUTDOWN MARGIN (SOM) ....... . ... .. .................. B 3 . 1. 1-1 B 3 .1. 2 Reactivity An omal i es . . ..... ......................... B 3 .1. 2- 1 B 3.1.3 Co nt ro l Rod OPE RABILIT Y ... . . . . ...................... B 3 .1. 3-1 B 3 . 1. 4 Con t ro l Ro d Scram Ti mes ............ ...... . .......... B 3 . 1 . 4-1 B 3. 1.5 Control Rod Scram Accumulators ...... . .. . . ...... . .... B 3 . 1. 5-1 B 3. 1.6 Rod Pattern Control .. .. .... ... .. .. ..... .. ........... 8 3 . 1.6-1 B 3 . 1.7 Standby Li quid Control (SLC) System ...... . ... . .. ... . 8 3 . 1 . 7-1 B 3 . 1.8 Scram Disc harge Vo lum e (SD V) Vent and Drai n Valves . . B 3 . 1.8- 1 B 3.2 POWER DISTRIBUTION LIMITS B 3. 2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ......... . . . . ......... .. . ................. B 3 . 2. 1-1 B 3 . 2. 2 MINIMUM CRITICA L POW ER RATIO (MCPR) ................. B 3 . 2. 2-1 B 3. 2. 3 LIN EAR HEAT GE NERATION RATE ( LH GR) ................. B 3 . 2. 3- 1 B 3.3 INSTRUMENTATION B 3 . 3 . 1.1 Reactor Protection System (RPS) Instrumentation .. ... B 3.3 . 1. 1-1 B 3 . 3 . 1.2 Source Range Monitor (SRM) Instrumentation ....... . .. B 3 . 3 . 1. 2-1 B 3 . 3 . 1.3 Osc i llation Power Range Monitor (OPRM)

Instru mentatio n ............... . ........... . ...... . B 3 . 3 . 1.3-1 B 3. 3. 2 . 1 Control Rod Block Instrumentation .. .. .. ............. B 3. 3 . 2. 1-1 B 3 . 3 . 2. 2 Feedwater System and Main Turbine High Water Level Trip Instrumentation .. . . . .. .... . .. ................ B 3 . 3 . 2. 2-1 B 3. 3. 3 .1 Post Accident Monitoring (PAM) Instrumentation . .. . . . B 3 . 3 . 3 . 1-1 B 3.3.3. 2 Remote Shu tdown Mon itoring System ..... . . . ... . ..... .. B 3. 3.3 . 2-1 B 3 . 3. 4 .1 End of Cyc l e Rec ir cul atio n Pum p Trip (EOC-R PT )

Instrume ntation ... . ............................... B 3 . 3. 4 .1 -1 B 3. 3. 4. 2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation . . ............ B 3 . 3 . 4. 2-1 B 3.3.5 . 1 Emergency Core Cooling System (ECCS)

Instru mentation ....... .. .. . ... .......... ... ...... . B 3. 3.5 . 1-1 8 3.3.5 . 2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ........................... 8 3.3.5.2-1 B 3 . 3 . 5. 3~ Reactor Core Isolation Cooling (RCIC) System Instrumentation .................. .. . . .. .. . .. . ..... B 3 . 3.5. 3~ -l B 3.3 . 6.1 Primary Contain ment Isolation Instru mentat i on .. . .... B 3 . 3. 6 . 1-1 B 3.3.6. 2 Seco ndary Contain ment Iso l at i on Instrumentation ..... B 3 . 3 . 6 . 2-1 B 3 . 3 . 7.1 Con t rol Roo m Area Filt ratio n (CRAF )

System Instrumentation .. ...... ....... ...... .... .. . 8 3 . 3 . 7. 1-1 (continued)

LaSalle 1 and 2 Rev i sio n

TABLE OF CONTENTS B 3.3 INSTRUMENTATION (continued)

B 3.3.8. 1 Loss of Power (LOP) Instru mentat i on ........... . ..... B 3.3.8 .1 -1 B 3.3 .8.2 Reactor Protection System (RPS) Electric Power Monitoring . ... .. .. . ... . . .. . . . .... . ........ .. . .. .. . B 3.3 . 8.2-1 B 3. 4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operat i ng . ............... . .. ... . B 3 . 4 . 1-1 B 3.4.2 Fl ow Contro l Valves (FC Vs) ... ... . ................ ... B 3.4.2- 1 B 3.4.3 Jet Pumps ........ . ..... . .. . .... . . ................. .. B 3 . 4 . 3-1 B 3. 4. 4 Safety/Relief Valves (S/RVs) ............ .. . ......... B 3 . 4. 4- 1 B 3.4. 5 RCS Operat i onal LEAKAGE ...................... .. .. ... B 3. 4. 5- 1 B 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage ......... . B 3.4 . 6-1 B 3. 4. 7 RCS Leakage Detect i on Instrumentation ............... B 3 . 4 . 7-1 B 3 . 4 .8 RCS Specific Activity . ................. .. .. ... .. . ... B 3 . 4 .8-1 B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown ......................... . ..... B 3 . 4 . 9-1 B 3 . 4 .10 Res i dua l Heat Remo val (RHR) Shutdo wn Cooli ng System-Cold Shutdown .. . . .. . ......... . ...... . ... . .. B 3 . 4.10-1 B 3 . 4. 11 RCS Pressure and Temperature (P/T) Limits ........... B 3 . 4. 11-1 B 3.4 . 12 Reactor Steam Dome Pressure . . .... . ............ . . . .. . B 3 . 4.1 2-1 B 3.5 EME RGENCY CORE COOLI NG SYSTE MS (ECCS) , REACTOR PRESSURE VESSEL (RPVJ WATER INVENTORY CONTROL, AND REACTO R CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5. 1 ECCS-Operati ng . .. .. . ... . . . .. . . . .... .. . . .. ........... B 3 . 5 . 1-1 B 3.5. 2 ECG~ ~hl:.ltdovmRPV Water Inventory Control .. .. . ... . ... B 3. 5. 2-1 B 3 . 5. 3 RCIC System .. . . .. . . . . .. . . . .................... ..... . B 3 . 5 . 3-1 B 3.6 CONTAINMENT SYSTEMS B 3 . 6. 1.1 Primary Containment ................... . ... .. . ...... . B 3 . 6 . 1. 1-1 B 3 . 6. 1.2 Primary Containment Air Lock ... . ... ... . ......... . ... B 3.6 . 1.2-1 B 3 . 6. 1.3 Primary Containment Isolation Valves (PCIVs) ... .. ... B 3 . 6 . 1. 3-1 B 3.6.1.4 Drywell and Suppression Cha mber Pressure ... .. ....... B 3. 6 . 1. 4-1 B 3. 6.1. 5 Dry well Air Temperat ure ......... ... .. ..... .......... B 3 . 6.1. 5-1 B 3. 6. 1.6 Suppression Chamber-to-Drywell Vacuum Breakers . ..... B 3.6.1 . 6-1 B 3. 6. 2 .1 Suppression Pool Average Temperature .... . . . ... . . . . . . B 3 . 6 . 2. 1-1 B 3. 6 . 2. 2 Sup pression Poo l Water Le vel ....................... . B 3 . 6 . 2. 2-1 B 3. 6. 2. 3 Res i dual Heat Remo val (RHR) Suppression Poo l Cooling ............................. .... . .. ....... B 3 . 6 . 2. 3-1 B 3 . 6. 2. 4 Res i dual Heat Remo val (RHR) Suppress i on Poo l Spray .. B 3 . 6. 2. 4-1 B 3. 6 . 3 . 1 Primary Containment Hydrogen Recombiners .. .. .. ...... B 3 . 6.3 . 1-1 B 3.6.3. 2 Primary Containment Oxygen Concentration .. .......... B 3 . 6. 3 . 2-1 B 3 . 6. 4 .1 Seco ndary Conta in ment .... . ... ..... . ............ . .... B 3.6. 4.1 -1 B 3 . 6. 4. 2 Secondary Containment Isolation Valves (SCIVs) ...... B 3.6 . 4.2-1 B 3 . 6. 4. 3 Standby Gas Treatment (SGT) System .. ... . ............ B 3.6.4.3-1 (con t i nu ed)

LaSalle 1 and 2 ii Revision

ECCS Instrumentation B 3 . 3 . 5.1 BASES APPLICABLE ECCS instrumentation satisfies Criterion 3 of SAFETY ANALYSES, 10 CFR 50.36(c)(2)(ii) . Certain instrumentation Fun ct ion s LC O, and are retained for other reasons and are described below in APPLICABILITY the individual Functions discussion.

(continued)

The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions spec ified in Table 3.3 . 5. 1-1 . Each Function must have a required number of OPERAB LE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is cal ibrated consistent with applicable setpoint methodology assumptions.

Each ECCS subsystem must also respond within its assumed response time . Table 3.3 . 5. 1-1, ~Footnote ( at ), is added to show that certain ECCS instrumentation Fun ctions are also required to be OPERABLE to perform DG initiation.

Allowable Values are specified for each ECCS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The no mi nal setpoints are selected to ensure that the set points do not exceed the Allowable Value between CHAN NEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal trip setpoint , but within its Allowable Value, is acceptable. A channel is inoperable if it s actual trip setpo int is not within its required Allowable Value . Trip setpoints are those predetermined values of output at which an action should take place . The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the proce ss parameter exceeds the set point, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration , and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors .

These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip set point s and Allowable Values determined in this manner provide adequate protection (continued)

LaSalle 1 and 2 B 3 . 3.5.1-8 Revision

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE l.a. 2.a. Reactor Vessel Water Level-Low Low Low . Level 1 SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Reactor Vessel Water Level-Low Low Low, Level 1 signals are initiated from four differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level ( variable leg) in the vessel . The Reactor Vessel Water Level-Low Low Low, Level 1 A11 owabl e Value i s chose n to allow time for the l ow pressure core flooding systems to activate and provide adequate cooling.

Two channels of Reactor Vessel Water Level-Low Low Low, Level 1 Function per associated Division are only required to be OPERABLE when t he associated ECCS is requ i red to be OPERABLE for automatic initiation, to ensure that no single instrument failure can preclude the ECCS function. (Two channels input to LPCS, LPCI A, and the associated Division 1 DG, while the other two channels input to LPCI B, LPCI C, and Di vi s ion 2 DG). Refer to LGO 3.5.1 and LGO 3 . 5. 2, "EGGS Sh1,1tdo*,m," for Applicability Bases for the l 01.* press1,1re EGGS s1,1bsystems.

l.b. 2.b. Drywell Pressure-High High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB) . The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell Pressure-High Function in order to minimize the possibility of fuel da mage. The core cooling function of the ECCS, along with the scra m action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50 .46.

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

The Drywell Pressure-High Function is required to be OPERABLE when the associated ECCS is required to be OPERABLE in conjunctio n with times when the primary containment is (continued)

LaSalle 1 and 2 B 3 . 3 . 5. 1-10 Revision

ECCS Instrumentation B 3.3 . 5 . 1 BASES APPLICABLE 1.b, 2.b. Drywell Pressure-High (continued)

SAFETY ANALYSES ,

LCO, and required to be OPERABLE . Thus , four channels of the LPCS APPLICABILITY 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 initiation . (Two channels input to LPCS , LPCI A, and the Division 1 DG, while the oth er two channe l s i nput to LPC I B, LPCI C, and t he Di vis i on 2 DG . ) In MOD ES 4 and 5 , t he Drywe ll Press ure- Hig h Fun ctio n i s not re qu ire d s i nce t her e is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure-High setpoint . Refer to LGO 3.5 . 1 for Applicability Bases for the low pressyre EGGS SYbsysteFAS .

l . c , 2. c . LPCI Pump A and Pump B Start-Time Dela y Rela y The pu rpose of this t im e de l ay is to s t agger t he start of the t wo ECCS pump s t ha t are in eac h of Di vi s i ons 1 and 2 ,

thus l i miting t he sta rtin g tra nsients on the 4. 16 kV emergency buses . This Function is only necessary when power is being supplied from the standby power sources (DG). On ECCS initiation, the time delay is bypassed if the normal feed breaker to the Class lE switchgear is closed . The LPCI Pump Start-T ime Delay Relays are assu med to be OPE RABL E i n the accident and transient analyses requiring ECCS initiation . That is, the analysis assumes that the pumps will initiate when required and excess loading will not cause failure of the standby power sources (DG).

There are t wo LPCI Pum p Start-T im e De l ay Relays , one in each of the RHR "A" and RHR "B " 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 from the emergency bus , to perfo rm their i ntended functio n within the assu med ECCS RESPONSE TIMES (e.g., as in the case where both ECCS pumps on one emergency bus start simu l taneously due to an inoperable time delay relay) . This st il l leaves two of the four low press ure ECCS pumps OPERABLE ; thus, the sing l e fa i 1 ur e criter i on is met (i . e. , 1oss of one inst r ument does not pr ec l ude ECCS init i ation) . The Al l owable Val ue for the LPCI Pump Start-Time Delay Relays is chosen to be short enough so that ECCS operation is not degraded .

(continued)

LaSalle 1 and 2 B 3 . 3 . 5 . 1 -11 Revision

ECCS Instrumentation B 3 . 3.5 .1 BASES APPLICABLE l.c. 2.c . LPCI Pump A and Pump B Start-Time Delay Relay SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Each LPCI Pump Start-Time Delay Relay Fun ct ion is required to be OPERABLE when the associated LPCI subsystem is required to be OPERABLE . Refer to LGO 3.5.1 and LGO 3.5 . 2 for Applicability Bases for the LPGI subsystems.

l.d, l,q, 2.d , 2.f. Reactor Steam Dome Pressure-Low (Injection Permissive) and LPCS and LPCI Injection Line Pressure-Low (Injection Permissive)

Low reactor steam dome pres sure and injection line pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems maximum design pressure. The Reactor Steam Dome Pressure-Low (Injection Permissive) and LPCS and LPCI Injection Line Pre ssure- Low (Injection Permi ss ive) are two of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. In addition, the Reactor Steam Dome Pressure-Low (Injection Permissive) and LPCS and LPCI Injection Line Pressure-Low (Injection Permissive)

Functions are directly assumed in the analysis of the recirculation line break (Ref. 2). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50 .4 6 .

The Reactor Steam Dome Pressure-Low (Injection Permissive) signals are initiated from four pre ss ure switches that sense the reactor dome pressure . The LPCS and LPCI Injection Line Pressure-Low (Injection Permissive) signals are initiated from four pressure switches that sense the pressure in the inject i on line (one sw it ch for each low pressure ECCS injection line). The Allowable Values are low enough to (continued)

LaSalle 1 and 2 B 3. 3. 5. 1-12 Revision

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE l.d. l.q . 2. d . 2.f . Reactor Steam Dome Pressure-Low SAFETY ANALYSES, (Injection Permissive) and LPCS and LPCI Injection Line LCO, and Pressure-Low (Injection Permissive) (continued)

APPLICABILITY prevent overpressurizing the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the l i mits of 10 CFR 50 . 46.

Two channels of Reactor Steam Dome Pressure-Low ( Injection Permissive) Function per associated Division and one channel of LPCS and LPCI Injection Line Pressure-Low (Injection Permissive) per associated injection line are only required to be OPERABLE when the associated ECCS i s requ ired to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. (Two channels of Reactor Vessel Pressure-Low (Injection Permissive) are required for LPCS and LPCI A, while two other channels are required for LPCI B and LPCI C. In addition, one channel of LPCS Injection Line Pressure-Low (Injection Permissi ve) is required for LPCS, while one channel of LPCI Injection Line Pre ss ure is required for each LPCI subsystem) . Refer to LGO 3 . 5.1 and LGO 3.5.2 for Applicability Bases for the low pressyre EGGS sybsystems.

l.e . l . f. 2.e . LPCS and LPCI 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 i s not sufficiently open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump.

The LPCI and LPCS Pump Discharge Flow-Low (Bypass)

Functions are assumed to be OPERABLE and capable of closing the minimum flow valve s to ensure that the low pressure ECCS flows assumed during the transients and accidents analyzed in References 1, 2, and 3 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.

(continued)

La Salle 1 and 2 B 3. 3.5.1 -13 Revision

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE l . e . l . f. 2.e. LPCS and LPCI Pump Discharge Flow-Low SAFETY ANALYSES, (Bypass) (continued)

LCO, and APPLICABILITY One flow switch per ECCS pump is used to detect the associated subsystems flow rate . The logic is arranged such that each switch causes its associated minimum flow valve to open when flow is low with the pump running. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum f l ow valves are ti me delayed such that the valves will not open for approximately 8 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 (Bypass) Allowable Values are high enoug h to ens ur e that t he pump flow ra t e is s uff i cient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core .

Each channel of Pump Di scharge Flow-Low (Bypass) Funct i on (one LPCS channel and three LPCI channe l s) is on l y 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. Refer to LGO 3.5 . 1 and LGO 3. 5.2 for Applicability Bases for the low press~re Eccg s~bsystems.

l . h. 2.q. Manual Initiation The Manual Initiation push button channels introduce signals into t he appropriate ECCS logic to pro vide manua l initiation capability and are redundant to the automatic protective instrumentation. There is one push button for each of the two Divisions of low pressure ECCS (i . e . , Division 1 ECCS, LPCS and LPCI A; Division 2 ECCS, LPCI Band LPCI C).

The Manual Initiation Function is not assumed in any accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the low pressure ECCS function as required by the NRC in the plant licensing basis.

(continued)

LaSalle 1 and 2 B 3. 3. 5. 1-14 Revision

ECCS Instrumentation B 3 . 3 . 5. 1 BASES APPLICABLE l . h. 2. q. Manual Initiation (continued)

SAFETY ANALYSES, LCO , and There is no Allowable Value for this Function s i nce the APPLICABILITY channels are mechanically actuated based solely on the position of the push buttons . Each channel of the Manual Initiation Function (one channel per division) is only required to be OPERABLE when the associated ECCS is required to be OPERABL E for automatic ali gnment and injection . Refer to LGO 3 . 5.1 and LGO 3.5.2 for Applicability Bases for the low pressure EGGS subsystems.

Hi gh Pressure Core Spra y System

3. a . Reactor Vesse l Wat er Level - Low Low. Leve l 2 Low RPV water level indicates that the capability to cool the fuel may be threatened . Should RPV water le vel decrease too f ar , fuel damage could result . Therefore , the HPCS System and assoc i ate d DG i s i ni ti ate d at Leve l 2 t o maint ain l evel above t he top of the acti ve fuel . The Reactor Vessel Water Level-Low Low , Level 2 is one of the Functions assumed to be OPERABLE and capable of initiating HPCS during the transients analyzed in References 1 and 3 . The Reactor Vesse l Water Level-Lo w Low, Le vel 2 Fu nction associated wi t h HPCS is dir ect l y assu med i n the anal ysis of th e recirculation line break (Ref. 2) . The core cool i ng function of the ECCS , along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the li mits of 10 CFR 50.46.

Reactor Vesse l Water Le vel-Low Low, Le vel 2 sig nal s are initiated from four differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable l eg) in t he vesse l. The Reactor Vessel Water Le vel-Low Low, Le vel 2 All owable Va l ue is chosen such that for complete loss of feedwater flo w, the Reactor Core Isolation Cooling (RCIC) System flow with HPCS assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Reactor Vessel Water Leve l -Low Low Low , Level 1.

(continued)

LaSal l e 1 and 2 B 3 . 3 . 5. 1-15 Revision

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 3.a . Reactor Vessel Water Level-Low Low . Level 2 SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are only required to be OPERABLE when HPCS is required to be OPERABLE to ensure that no single instrument failure can preclude HPCS initiation . Refer to LCO 3 . 5. 1 and LCO 3.5.2 for HPCS Applicability Bases.

3.b . Drywell Pressure-High High pressure in the drywell could indicate a break in the RC PB. The HPCS Syste m and associated DG are i ni t i ated upon rece ipt of t he Dry well Press ur e-High Fun ction in orde r to minimize the possibi l ity of fue l damage . The core coo l ing function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains belo w the l im its of 10 CFR 50 . 46 .

Drywell Press ure-H i gh s i gnals are ini t iated fro m four pressure switches that sense drywell pressure . The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment .

The Dry well Pr essure-Hi gh Fun cti on i s r equire d to be OPERABLE when HPCS is required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE . Thus , four channels of the HPCS Drywell Pressure-High Function are required to be OPERAB LE in MO DE S 1 , 2, and 3 , to ensure that no single ins t rument fai l ure can prec l ude ECCS in it iation . 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 Functions setpoint.

Refer to LCO 3 . 5. 1 for the Applicabi l ity Bases for the HPCS System .

3. c . Reactor Vessel Water Level-High . Level 8 High RPV water level indicates that sufficient cooling water inventory exists in the reactor vesse l such that there i s no danger to the f uel. Th erefore , the Le vel 8 s i gnal is used to cl ose the HPCS injection val ve to prevent overflow into the main steam lines (MSLs). The Reactor Vesse l Water (continued)

LaSal l e 1 and 2 B 3 . 3 . 5. 1-16 Revision

ECCS Instrumentation B 3.3 . 5. 1 BASES APPLICABLE 3.c. Reactor Vessel Water Level-High . Level 8 (continued)

SAFETY ANALYSES, LCO, and Level-High, Level 8 Function for HPCS isolation is not APPLICABILITY credited in the accident analysis. It is retained since it is a potentially significant contributor to risk.

Reactor Vessel Water Level-High, Level 8 signals for HPCS are i ni tiated f r om two level t r ansmitte r s from t he narrow range water le vel measurement instru ment ation . The Reactor Vesse l Water Level -High, Level 8 Allowab l e Value i s chosen to isolate flow from the HPCS System prior to water overflowing into the MSLs .

Two channels of Reactor Vessel Water Le vel-High , Level 8 Function are only required to be OPERABLE when HPCS is required to be OPERABLE to ensure that no single instrument failure can preclude HPCS initiation . Refer to LGO 3 . 5.1 and LGO 3.5 . 2 for HPGS Applicability Bases.

3 . d . 3 . e . HPCS Pump Dischar ge Pressure-High (B ypass) and HPCS System Flow Rate-Low (Bypass)

The minimum flow instruments are provided to protect the HPCS pump from overheating when the pump is operating and the associated i njection valve is not sufficiently open .

The minimum flow line valve is opened when low flow and high pump discharge pressure are sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump or the disc harge pressure is low (indicating t he HPCS pump is not operating). Th e HPCS System Fl ow Rate- Low ( Bypass) and HPCS Pum p Disc harge Pressure-High Functions are assumed to be OPERABLE and capable of closing the minimum flow valve to ensure that the ECCS flow assumed during the transients and accidents analyzed in References 1, 2, and 3 are met . The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cl adding te mperature remains below the limits of 10 CFR 50 . 46 .

One f l ow switc h is used to detect the HPCS System ' s f l ow rate. The logic is arranged such that the switc h causes the minimum flow valve to open , provided t he HPCS pum p di sc harge pressure, sensed by anot her switch, is high enough (continued)

LaSalle 1 and 2 B 3.3.5 . 1-17 Revis i on

ECCS Instrumentation B 3.3.5 . 1 BASES APPLICABLE 3 . d . 3.e . HPCS Pump Discharge Pressure-High (Bypass) and SAFETY ANALYSES, HPCS System Flow Rate- Low (Bypass) (continued)

LCO , and APPLICABILITY (indicating the pump is operating) . The logic will close the mi nimum flow valve once the closure setpoint i s exceeded . (The valve will also close upon HPCS pump discharge pressure decreasing below the setpoint . )

Th e HP CS Syst em Flo w Rat e-Lo w ( Bypass ) All owabl e Val ues are hi gh enough to ensure th at pu mp flow r at e is su ffi cient t o 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. The HPCS Pump Discharge Pressure- High (By pass) Al lo wab l e Va lu e i s set high enough to ensure th at t he valv e will not be open wh en th e pump i s not ope r ati ng.

One channel of each Function is required to be OPERABLE when the HPCS is required to be OPERABLE . Refer to LCO 3 . 5.1 and LCO 3.5 . 2 for HPCS Applicability Bases.

3. f . Manual Initiation The Manual Initiation push button channel introduces a signa l into t he HPCS logic to pr ovide manua l i nitiatio n capab ili ty and i s re dund ant t o t he automati c pr ot ect iv e instrumentation . There is one push button for the HPCS System .

The Manual Initiation Function is not assumed in any accident or t ra nsient anal yses i n the UF SAR . However, th e Funct i on is re t ai ned f or overa ll redu nd ancy and di versity of the HPCS function as required by the NRC in the plant licensing basis .

There i s no All owab l e Value fo r this Function since t he chan nel is mec hanica ll y actuated based solely on the positio n of t he push button . One channe l of t he Manual Initiation Function is only required to be OPERABL E when the HPCS System is required to be OPERABLE . Refer to LCO 3 . 5. 1 and LCO 3 . 5. 2 for HPCS Applicability Bases .

(cont i nued)

LaSalle 1 and 2 B 3. 3. 5. 1-18 Revision

ECCS Instrumentation B 3 . 3 . 5.1 BASES ACT IONS B.1 . B.2. and B.3 (continued)

Divisions, this results in the affected portions in both Divisions of ECCS and DG being concurrently declared inoperable. For Required Action B. 2 , redundant automatic initiation capability (i.e., loss of automatic start capability for either Functions 3.a or 3. b) i s lost if two Function 3 . a or two Function 3.b parallel contacts (channels) are inoperable and untripped in the same trip system.

In this situa tion (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.3 is not appropriate and the feature(s) associated with the inoperable, untripped channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1 to Reqyired Action g_1 and Reqyired Action g _2), the tHo ReqYired Actions are only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the EGGS is not assymed and the probabi 1 i ty of a LOGA is 1 ower . Thys, a tot a 1 1os s of initiation capability for 24 hoyrs (as allowed by Reqyired Action g _3) is allo*,ied dYring MODES 4 and 5. Notes are a+-5 provided ( the Note~ to Required Action B.1 and Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable. This ensures that the proper loss of initiation capability check is performed.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Comp l etion Ti me also allows for an exception to the normal "time zero" for beginning the all owed outage ti me "clock. "

For Required Action B. 1, the Completion Time only begins upon discovery that a redundant feature in both Divisions (e.g ., any Division 1 ECCS and Division 2 ECCS) cannot be automatically initiated due to inoperable, untripped channels within the same variable as described in the paragraph above . For Required Action B. 2, the Completion Time only begins upon discovery that the HPC S System cannot be automatically initiated due to two inoperable, untripped channels (parallel contacts) for the associated Function in the same trip system . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it min i mize s risk while allowing time for restoration or tripping of channels.

(continued)

LaSalle 1 and 2 B 3.3 . 5. 1-25 Revision

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C. l and C.2 (continued) and 2. c, the affected portions of the Division are LPCI A and LPCI B, respectively. In addition, the specific inoperability of these Functions shou l d also be evaluated for impact on the DGs .

In this situat i on (loss of redundant automat i c in itiat i on capa bil ity) , th e 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months all owance of Required Acti on C. 2 i s not app ropr i at e and t he f eature(s) associated wi t h the inoperable channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1), the Required Action is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the EGGS is not assumed and the probability of a LOC.O. is lo11er . Thus, a total loss of automatic initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action G. 2) is allo11ed during MODES 4 and 5.

The Note---- states that Requ i red Act i on C. l is only app l i cable f or Fu nc ti ons l .c and 2.c . Th e Re qu i r ed Act i on i s not applic able to Fun ct i ons l . h, 2. g , and 3 . f (whic h al so require entry into this Condition if a channel in these Functions is inoperable) , since they are the Manual Initiation Functions and are not assumed in any accident or transient ana l ysis . Th us , a total loss of ma nu al in i tiation capa bil i ty fo r 24 hou r s (as al l owed by Require d Acti on C. 2) is all owed . Required Action C.l is al so 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 the Function was considered during the development of Refe r ence 4 an d cons id er ed acceptable f or the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months al l owed by Required Action C. 2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities . This Co mpletion Tim e also all ows for an exception to t he normal "time zero" for beginn i ng the al lowed outage t i me "clock . "

For Required Act i on C. l, the Co mpletion Ti me only begins upon discovery that the same feature in both Divisions (i.e . , any Division 1 ECCS and Division 2 ECCS) cannot be automatically i nitiated due to i noperab l e channels within the sa me variab l e as described in the paragraph above . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Complet i on Ti me from discovery of l oss of i nit i at i on capab il ity is acceptab l e because it min im i zes risk whi l e allowing time for restoration of channels .

(continued)

LaSalle 1 and 2 B 3. 3. 5.1-27 Revision

ECCS Instrumentation B 3 . 3.5.1 BASES ACTIONS D.l, D. 2. D.3 . and D.4 (continued) inoperable l.d channels or one inoperable l.g channel. For Function l . g, redundant automatic initiation capability is lost if two Function l.g channels are inoperable concurrent with either two inoperable Function 2.d channels or one inoperable Function 2.f channe l. For Fun ction 2. f, redundant automatic initiation capabi lity is lost if two Function 2.f channels are inoperable concurrent with two inoperable l . d channels or one inoperable l . g channel . Since each inoperable channel would have Required Action D. l applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected low pressure ECCS pu mp to be dec l ared i noperable . However, since channe l s for more than one low pressure ECCS pump are inoperable, an d 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 red und ant automatic initiat i on capability), the Completion Times of Required Actions D. 3 and D.4 are not appropriate and the feature(s) 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 after discovery of loss of initiation capabi li ty for feature(s) in both Divisions. As noted (Note 1 to Required Action D.1), Required Action D.l is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the lo*.1 pressure EGGS is not assumed and the probability of a LOGA is lower. Thus, a total loss of initiation capability for 7 days for Functions l.e, l.f, and 2. e (as allo*.1ed by Required Action D.4) is allowed during MODES 4 and 5. (This Condition is not entered when Functions l.d, l.g, 2.d or 2.f are inoperable in MODES 4 and 5.) A Note is -a-1--s-&-provided ( The Note~ to Required Action D.l) to delineate that Required Action D. l is only applicable to low pressure ECCS Fu nctio ns . Required Action D. l i s not app l icable to HPCS Functions 3.d and 3. e since the loss of one channel results in a loss of the Function (one-out-of-one logic) . This loss was considered during the development of Reference 4 and considered acceptable for the 7 days allowed by Required Action D. 4 .

Required Action D. 2 i s i ntended to ens ur e that appropr i ate (continued)

LaSalle 1 and 2 B 3.3.5.1-29 Revision

ECCS Instrumentation B 3 . 3.5.1 BASES SURVEILLANCE SR 3.3.5 . 1.2 (continued)

REQUIREMENTS clarifies what is an acceptable CHANN EL FUNCTIONAL TEST of a relay . This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions .

Any se t point adj ust ment s hall be cons i sten t wi t h t he assumptions of the current plant specific setpoint methodology .

The Surveil l ance Fre quency is control l ed under the Sur vei l lance Fr equency Co ntro l Progra m.

SR 3.3. 5. 1. 3 and SR 3. 3.5. 1.4 A CHANN EL CAL IBRAT ION i s a compl ete check of t he i nstr um ent l oop and t he se nsor. Thi s tes t verif i es the chann el responds to the measured parameter wit hin the necessary range and accuracy . CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calib r ations co nsiste nt with t he pla nt spec i fic setpoi nt met ho dol ogy .

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

SR 3.3 . 5 . 1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific chan nel . The system f unctiona l testi ng performed in LCD 3 . 5. 1, LGO 3. 5. 2, LCD 3 . 8 . 1 , and LCD 3.8.2 overlaps this Surve il lance to pro vi de co mplete testing of the assumed safety function .

(cont i nued)

LaSalle 1 and 2 B 3.3.5 . 1-36 Revision

RPV Water Inventory Control Instrumentation B 3 . 3.5. 2 B 3. 3 I NSTRUMENTATION B 3.3 . 5.2 Reactor Pressure Vessel (RPVJ 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 (LSSSJ for variables that have sign i ficant safety functions . LSSS are defined by the regulat i on as "Where a LSSS is specified for a variab l e on which a safety lim i t has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded . "

The Analytical Limit is the limit of the process variable at which a safety action is initiated to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The actual settings for the automatic isolation channels are the same as those established for the same functions in MODES 1, 2, and 3 in LCO 3.3.5.1, "Emergency Core Cooling System (ECCS) Instrumentation," or LCO 3 . 3.6.1, "Primary Containment Isolation Instrumentation . "

With the unit in MODE 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1 . 1. 3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur . Under the definition of DRAIN TIME, some penetration flow paths may be excluded from the DRAIN TIME calculation if they will (continued)

LaSal l e 1 and 2 B 3.3.5.2-1 Revis i on

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES BACKGROUND be isolated by valve s that will close automatically without (continued) 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 LCD 3.5.2, "RPV Water Inventory Control," and the definition of DRAIN TIME. There are functions that are required for manual operation of the ECCS injection / spray subsystem required to be OPERABLE by LCD 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 low pressure core spray (LPCS), low pressure coolant injection (LPCI), and high pressure core spray (HPCS). The equipment involve d with each of these systems is described in the Bases for LCD 3 . 5.2.

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is SAFETY not required to mitigate any events or accidents evaluated ANALYSES, LCD, in the safety analyses. RPV water inventory control is and APPLICABILITY required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur.

A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is postulated in which a single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e . g., seismic event, loss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one ECCS injection / spray subsystem can be manually operated to maintain adequate reactor ve s sel water level.

(continued)

LaSalle 1 and 2 B 3 . 3 . 5 . 2-2 Revision

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

(continued)

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

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

Low Pressure Coolant Iniection Systems and Low Pressure Core Wll l.a. l.d. 2.a. 2.c. Reactor Steam Dome Pressure-Low (Iniection Permissive) and LPCI and LPCS Iniection Line Pressure-Low Ciniection Permissive)

Low reactor steam dome pressure and LPCI and LPCS injection line pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure.

The Reactor Steam Dome Pressure-Low (Injection Permissive) signals are initiated from four pressure switches that sense the reactor dome pressure. The LPCI and LPCS Injection Line Pressure-Low (Injection Permissive) signals are initiated from four pressure switches that sense the pressure in the injection line (one switch for each low pressure ECCS injection line).

The Allowable Values are low enough to prevent overpressurizing the equipment in the low pressure ECCS.

One channel of Reactor Steam Dome Pressure-Low (Injection Permissive) Function per associated Division and one channel of LPCI and LPCS Injection Line Pressure-Low (Injection Permissive) per associated injection line are only required to be OPERABLE in MODES 4 and 5 when the associated subsystem is required to be OPERABLE by LCO 3.5.2, since these channels support the manual operation of these systems.

(continued)

LaSalle 1 and 2 B 3. 3 . 5 . 2-3 Revision

RPV Water Inventory Control Instrumentation B 3 . 3 . 5. 2 BASES APPLICABLE I . a. 1. d. 2.a. 2.c. Reactor Steam Dome Pressure-Low SAFETY Ciniect i on Permissive) and LPCI and LPCS Infection Line ANALYSES, LCD, Pressure-Low Ciniection Permissive) (continued) and APPLICABILITY In addition, one channel of LPCS Injection Line Pressure-Low (Injection Permissive) is required to be operable for a required LPCS system, and one channel of LPCI Injection Line Pressure-LOW (Injection Permissive) is required to be OPERABLE for a required LPCI subsystem.

l.b. I.e. 2.b. LPCS and LPCI Pump Discharge Flow-Low CBvoassJ 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 sufficiently open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump.

One flow switch per ECCS pump is used to detect the associated subsystem flow rate. The logic is arranged such that each switch causes its associated minimum flow valve to open when flow is low with the pump running. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for approximately 8 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 (Bypass) Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the RPV.

One channel of Pump Discharge Flow-Low (Bypass) Function is required to be OPERABLE when the associated LPCS or LPCI pump is required to be OPERABLE by LCD 3.5.2 to ensure that the pump is capable of injecting into the Reactor Pressure Vessel when manually operated.

(continued)

LaSalle 1 and 2 B 3 . 3 . 5 . 2-4 Revision

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE High Pressure Core Spray System SAFETY ANALYSES, LCD, and 3.a. 3.b . HPCS Pump Discharge Pressure-High (Bypass) and APPLICABILITY HPCS System Flow Rate-Low (Bypass)

(continued)

The minimum flow instruments are provided to protect the HPCS pump from overheating when the pump is operating and the associated injection valve is not sufficiently open.

The minimum flow line valve is opened when low flow and high pump discharge pressure are sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump or the discharge pressure is low (indicating the HPCS pump is not operating).

One fl ow switch is used to detect the HPCS System's fl ow rate. The logic is arranged such that the switch causes the minimum flow valve to open, provided the HPCS pump discharge pressure, sensed by another switch, is high enough (indicating the pump is operating). The logic will close the minimum flow valve once the closure setpoint is exceeded. (The valve will also close upon HPCS pump discharge pressure decreasing below the setpoint.)

The HPCS System Flow Rate-Low (Bypass) Allowable Values are high enough to ensure that 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 RPV.

The HPCS Pump Discharge Pressure-High (Bypass) Allowable Value is set high enough to ensure that the valve will not be open when the pump is not operating.

One channel of each Function is required to be OPERABLE when the HPCS is required to be OPERABLE by LCD 3.5.2 in MODES 4 and 5.

RHR Shutdown Cooling System Isolation 4.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 automatically isolated by RPV water level isolation instrumentation prior to the RPV water level being equal to (continued)

LaSal l e 1 and 2 B 3. 3 . 5 . 2-5 Revi sion

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE 4.a. Reactor Vessel Water Level-Low , Level 3 (continued)

SAFETY ANALYSES, LCD, and the TAF. The Reactor Vessel Water Leve l -Low, Level 3 APPLICABILITY Function is only required to be OPERABLE when automatic isolation of the associated RHR penetration flow path is credited in calculating DRAIN TIME .

Reactor Vessel Water Level-Low, Level 3 signals are initiated from differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (i . e., two channels per trip system) of the Reactor Vessel Water Level-Low, Level 3 Function are available, only two channels (all 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 RPS Reactor Vessel Water Level-Low, Level 3 Allowable Value (LCD 3.3 . 1. 1) since the capability to cool the fuel may be threatened .

This Function isolates the Group 6 valves.

Reactor Water Cleanuo CRWCUJ System Isolation 5.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 automatically isolated by RPV water level isolation instrumentation prior to the RPV water level being equal to the TAF. 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 differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two per tip system) of the Reactor Vessel Water (continued)

LaSalle 1 and 2 B 3 . 3 . 5 . 2-6 Revision

RPV Water Inventory Control Instrumentation B 3.3 . 5. 2 BASES APPLICABLE 5.a . Reactor Vessel Water Level-Low . Low . Level 2 SAFETY ANALYSES, (continued)

LCD, and APPLICABILITY Level-Low Low, Level 2 Function are available, only two channels (all 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.I directs entry into the appropriate Condition referenced in Table 3 . 3.5.2-1. The applicable Condition referenced in the Table is Function dependent .

Each time a channel is discovered inope rable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition .

(continued)

La Salle 1 and 2 B 3 . 3 . 5 . 2-7 Revision

RPV Water Inventory Control Instrumentation B 3 . 3 . 5. 2 BASES ACT IONS 8.1 and 8.2 (continued)

RHR Shutdown Cooling 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 8.1 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation. Required Action 8.2 directs calculation of DRAIN TIME . The calculation cannot credit automatic isolation of the affected penetration flow paths.

Ll Low reactor steam dome and LPCI and LPCS injection line pressure signals are used as permissives for the manual operation of the low pressure ECCS injection / spray subsystems. If these permissives are inoperable, manual operation of the affected subsystem is prohibited.

Therefore, the affected 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 a permissive in the trip condition, manual operation may be performed.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is intended to allow the operator time to evaluate any discovered inoperabilities and to place the affected channel in trip prior to declaring the affected subsystem inoperable.

If a LPCI or LPCS Discharge Flow - Low bypass function or HPCS System Discharge Pressure - High or Flow Rate - Low bypass function is inoperable, there is a risk that the associated 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 (continued)

LaSalle 1 and 2 B 3. 3 . 5 . 2-8 Revision

RPV Water Inventory Control Instrumentation B 3 . 3 . 5. 2 BASES ACTIONS D.1 (continued) inoperabilities prior to declaring the affected subsystem inoperable . The Completion Time is appropriate given the ability to manually start the ECCS pumps and open the injection valves as necessary to ensure the affected pump does not overheat.

With the Required Action and associated Completion Time of Conditions C or D not met, the associated ECCS injection / spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately .

SURVEILLANCE As noted at t he beginning of the SRs, the SRs for each RPV REQUIREMENTS Water Inventory Control instrumentatio n Function are found in the SRs column of Table 3.3.5 . 2-1.

SR 3.3 . 5. 2. 1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred . A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST.

Agreement cr i teria 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.

(continued)

LaSalle 1 and 2 B 3. 3 . 5 . 2-9 Revision

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5 . 2.1 (continued)

REQUIREMENTS 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 LCD.

SR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.

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.

LaSalle 1 and 2 B 3 . 3 . 5. 2-10 Revision

RCIC System Instrumentation B 3 . 3 . 5. 3-&

B 3.3 INSTRUMENTATION B 3 . 3 . 5. 3-& Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vesse l is iso l ated from it s primary heat sink (the main condenser) and normal coo l ant makeup flow from the Reactor Feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level precluding initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps. A more complete di scussion of RCIC Syste m operation is provided in the Bases of LCO 3 . 5 . 3, "RCIC System ."

The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for condit ion s of Reactor Vesse l Wat er Level Low-Low, Level 2. The variable is mo nit ored by four differential pressure transmitters that are connected to four trip un i ts . The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement . The logic can also be initiated by use of a manual push button. Once initiated, the RCIC logic seals in and can be reset by the operator only when the rea ctor ve ssel water level s ignal s have cleared .

The RCIC test line isolation va lve is closed on a RCIC initiation sig nal to allow full system flow to the reactor vessel .

The RCIC Syste m al so monitors the water level in the condensate storage tank (CST), since there are two sources of water for RCIC operation. Reactor grade water in the CST is the normal source . Upon receipt of a RCIC initiation signal, the CS T suction valve is automatically signaled to open (it is normal ly in the open position) unless the pu mp suction from the suppression pool valve is open . If the water level in the CST falls below a preselected level, first the sup pre ssion pool suction valve automatica ll y opens and then the CST suction valve automatically closes . Two level swi tches a re used to detect low wat er level in the CS T. Either switch can cause the sup pre ssion pool suction valve to open . To prevent losing suction to the pump, (continued)

LaSalle 1 and 2 B 3 . 3 . 5 . 3-& - 1 Revision

RCIC System Instrumentation B 3 . 3 .5. Ji BASES BACKGROUND the suction valves are interlocked so that one suction path (continued) must be open before the other automatically closes .

The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) trip (two-out-of-two logic), at which time the RCIC turbine steam inlet isolation valve closes (the injection valve also closes due to the closure of the RCIC turbine ste am inlet isolation valve). The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2).

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

The OPERABILITY of the RCIC System instrumentation is dependent on the OPERAB I LITY of the individual instrumentation channel Functions speci fied in Table 3.3 . 5 .Ji -l. Each Function must have a required number of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Allowable Values are specified for each RCIC System instrumentation Function specified in the Table . Nominal trip setpoints are spec ified in the setpoint calculations.

The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within it s Allowable Value, i s acceptable . A channel is inoperable if its actual trip setpo int is not within its required Allowable Value.

(continued)

LaSalle 1 and 2 B 3.3 .5. Ji -2 Revision

RCIC System Instrumentation B 3.3.5. 3 BASES APPLICABLE Trip setpoints are those predetermined values of output at SAFETY ANALYSES, which an action should take place . The setpoints are LCO, and compared to the actual process parameter (e.g., reactor APPLICABILITY vessel water level), and when the measured output value of (continued) the process parameter exceeds the setpoint, the associated device (e.g ., trip unit) changes state . The analytic limits (or design limits) are derived from the limiting values of the process parameters obtained from the safety analysis .

The trip se tpoint s are determined from th e analyt i c lim its, corrected for defined process, ca lib ration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors . These calibration based errors are limited to reference accuracy, instrument drift, errors assoc i ated with measurement and test equipment, and calibration to ler ance of loop components . The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties , process effects, calibration tolerances, instrument drift , and severe environment errors (for channe l s th at mu s t function in ha rs h environments as define d by 10 CFR 50 .49 ) are accounte d for and appropriately applied for the i nstrumentation .

The individual Functions are required to be OPERABLE in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure> 150 psig, since this is when RCIC is required to be OPERABLE . Refer to LCO 3.5 . 3 for Applicability Bases for the RCIC System .

The specific Applicable Safety Analyses, LCO , and Applicability discussions are l i sted below on a Function by Fun ction bas i s .

1. Reactor Vessel Water Level-Low Low , Level 2 Low reactor pressure vessel CRPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened . Should RPV water level decrease too far.

fuel damage cou l d result . Therefore, the RCIC System is initiated at Level 2 to assist in maintaining water l evel above the top of the active fue l.

(continued)

LaSalle 1 and 2 B 3 . 3 . 5. 3-2--3 Revision

RCIC System Instrumentation B 3 . 3.5. 3-&

BASES APPLICABLE 1. Reactor Vessel Water Level-Low Low . Level 2 SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from four differential pressure transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the ve ssel .

The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure core spray assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level 1.

Four channels of Reactor Vessel Water Level - Low Low, Level 2 Function are available and are required to be OPERABLE when RCIC i s required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation.

Refer to LCO 3.5.3 for RCIC Applicability Bases.

2. Reactor Vessel Water Level-High, Level 8 High RPV water level indicates that suff icient cooling water inventory exists in the reactor ve ssel such that there i s no danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC turbine steam inlet isolation valve to prevent overflow into the main steam lines (MSLs) . (The injection valve also closes due to the closure of the RCIC turbine stea m inlet isolation valv e.)

Reactor Vessel Water Level-High, Level 8 signal s for RCIC are initiated from two differential pressure transmitters from the narrow range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actua l water l evel (variable leg) in the ves sel.

The Reactor Vessel Water Level-High, Level 8 Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.

(continued)

LaSalle 1 and 2 B 3.3 .5. 3-&-4 Revision

RCIC System Instrumentation B 3 . 3 . 5 .JJ BASES APPLICABL E 2. Reactor Vessel Water Level - High , Level 8 (continued)

SAFETY ANALYSES ,

LCO, and Two channels of Reactor Vessel Water Level-High, Level 8 APPLICABILITY Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation . Refer to LCO 3 . 5. 3 for RCIC App l icabi l ity Bases .

3. Co ndensate St ora ge Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.

Normally the suction val ve bet ween the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection wou l d be taken from the CST . Howe ver, if the water le vel in the CST fal l s below a preselected le vel, first the suppression pool suction va l ve auto mati cally opens and t hen the CS T suction va l ve automatically cl oses . This ensures that an adequa t e supp l y of make up water is ava il able to t he RCIC pum p. To preven t l osing s uct ion to the pum p, the suction va l ves are interlocked so that the suppress i on pool suction va l ve must be open before the CST suction va l ve automatically closes .

Two level switches are used to detect low water level in the CST. The Condensate Storage Tank Level-Low Function Allowable Va l ue is set high enough to ensure adequate pu mp suctio n head wh i le water is being take n from the CST .

Two channels of Conde nsate Sto r age Tan k Level-Low Funct i on are availab l e and are require d to be OPE RABLE when RC IC i s requ i red to be OPERABL E to ensure that no single instrument failure can prec l ude RCIC swap to suppression pool source .

Refer to LCO 3. 5.3 for RCIC Applicability Bases .

4. Manual Initiation The Ma nual In i t i atio n push button channel introduces a signa l into th e RCIC System i nitiat i on l ogic t hat is redun dant to th e automat i c pro t ective in strume ntation and prov i des manua l initiat i on capab il ity . There is one pus h button channe l for the RCIC System .

(continued)

LaSa l le 1 and 2 B 3.3 . 5.JJ -5 Revis i on

RCIC System Instrumentation B 3.3.5. 3-&

BASES APPLICABLE 4. Manual Initiation (continued)

SAFETY ANALYSES, LCO, and The Manual Initiation Function is not assumed in any APPLICABILITY accident or tran s ient anal ys es in the UFSAR . However, the Function is retained for overall redundancy and diversity of the RCIC function as required by the NRC in the plant licensing basis.

There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button. One channel of Manual Initiation is required to be OPERABLE when RCIC is required to be OPERABLE . Refer to LCO 3.5.3 for RCIC Applicability Bases.

ACT IONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for in operable RCIC System instrumentation channels prov id e appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System in strumentation channe l.

Required Action A.l directs entry into the appropriate Condition referenced in Table 3 . 3.5. 3.i -l in the acco mpanying LCO. The applicable Condition referenced in the Table is Function dependent. Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition .

(continued)

LaSalle 1 and 2 B 3. 3 . 5. 3.i-6 Revision

RCIC System Instrumentation B 3 . 3 . 5 .Ji BASES ACTIONS B. l and B.2 (continued)

Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System . In this case , automatic initiation capability is l ost if two Fu ncti on 1 parall el co ntacts (c hanne l s) in t he sa me t rip syst em are in ope rable an d un tri pped . In t his si tu atio n Closs of auto mat i c init i at i on capabi lity), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B. 2 is not appropriate , and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after disco very of loss of RCIC initiation capability .

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

For Required Action B. 1, the Complet i on Time onl y beg i ns upon di scove ry th at t he RCIC Sys t em cann ot be autom atic all y i nit i at ed due t o two i noperab l e , unt ripp ed Reacto r Vesse l Water Level-Low Low , Level 2 channels (parallel contacts) in the same trip system . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it mini mizes risk while allowing time for restoration or trippi ng of channels .

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not credited in any accident or transient analysis , an al l owable out of serv i ce time of 24 hou r s has been show n to be acce pt able (Ref. 1) to permi t res t orat i on of any inoperab l e channel to OP ERABLE status . If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability , restore capabi l ity to accom modate a sing l e failure , and al low operation to continue .

Alternately , if it is not desired to place the channel in tr i p (e.g . , as in the case where plac i ng the inoperable channel in tr i p would resu l t i n an in i tiation), Condit i on E must be entered and its Required Action t aken .

(continued)

LaSalle 1 and 2 B 3.3.5. 3-& -7 Revision

RCIC System Instrumentation B 3 . 3.5. 3J BASES ACT IONS Ll (continued)

A risk based analysis was performed and determined that an allowab1e out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.l ) . A Required Action (similar to Required Action B.1), limiting the allowable out of service time if a loss of automatic RCIC initiation capability ex i sts, is not required . This Condition applies to the Reactor Vessel Water Level-High, Level 8 Function, whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation (high water level trip) capability . As stated above, this loss of auto mati c RCIC initiation (hig h water l evel tr ip ) capability was analyzed and determined to be acceptable. Thi s Condition also applies to the Manual Initiation Function .

Since this Function is not assumed in any accident or transient analysis, a total loss of manual initiation capability (Required Action C.1) for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed.

The Required Action does not allow placing a channel in trip since this action would not necessari l y result in the safe state for the channel in all events .

0.1. 0.2.1. and 0.2.2 Required Action 0.1 is intended to ensure that appropriate actions are taken if multiple inoperable, untripped channels within the same Function result in automatic component initiation (RCIC source swapover) capability being lost for the feature(s). For Required Action 0. 1, the RCIC Syste m is the only assoc iat ed feature. In this case, auto mat ic component initiation (RCIC source swapo ver) capability is lost if two Function 3 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Actions 0. 2. 1 and 0.2 . 2 is not appropriate, and the RCIC Syste m must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability . As noted, Required Action 0.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed .

(cont inu ed)

LaSalle 1 and 2 B 3 . 3 . 5 . 3J -8 Revision

RCIC Sy s tem In strumentation B 3 . 3.5. 3i!-

BASES ACTIONS D. l , D. 2.1 , and D.2.2 (continued )

The Completion Time is intended to allow the operator time t o evalua t e and re pair an y di scove r ed inoperabilities. Thi s Completion Time al so allows for an excepti on to the normal "time zero" for beginning the allowed outage time "clock . "

For Required Action D. l. the Completion Time only begins upon discovery that the RCIC System ca nn ot be automatically alig ned t o the s up pression pool due to two ino perable, untripped cha nnel s in the same Function. The 1 hou r Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ha s been shown to be acceptable (Ref . 1) to permit restoration of any inoperable chan nel to OPERABLE stat us. If the inoperable channel cannot be resto r ed to OPERABLE s ta tus within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D. 2.1, which perform s the intended function of the channel (shifting the suction source to the suppression pool). Alternatively, Required Action D.2.2 allows the manu al alignment of the RCIC suct i on to the suppression pool, which also perform s t he intended function . If Required Action D. 2.1 or D.2.2 i s performed ,

measures should be taken to ensure that the RCIC System piping remains filled with water . If it i s not desired to perform Required Actions D.2 . 1 and D.2 . 2 (e . g ., as in th e case wh ere shi ft ing the suctio n source co ul d drain dow n th e RCIC suction piping), Condition E must be entered and its Required Action taken .

Ll With any Required Action and associated Complet i on Time not met, the RCIC System may be incapable of performing the intended funct i on, and the RCIC System must be declared inoperable immediately.

(continued)

LaSalle 1 and 2 B 3.3.5. 3-& -9 Revision

RCIC System Instrumentation B 3.3 . 5. Ji BASES (continued)

SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3 .3.5. Ji -1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows :

(a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Funct i ons 2 and 4; and (b) for up to 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s for Functions 1 and 3 provi ded the associated Function maintains RCIC initiation capability . Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the rel iabil ity analysis (Ref . 1) assumption of the average time required to perform channel Surveillance . That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary .

SR 3.3.5. 3i .1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occu rred. A CHANNEL CHECK is nor mall y a co mpariso n of the parameter indi cated on one channel to a similar parameter on other cha nnel s . It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same valu e .

Significant deviations between the instrument channels could be an indication of excessive i nstrument drift in one of t he channels or so mething even more serious . A CHANNEL CHECK will detect gross channe l failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION .

Agreement cr iteria are determined by the plant staff based on a combinatio n of the channel instrument uncertainties ,

including indication and readability . If a channel is outside the criteria, it may be an indication that the in strument has drifted outside its limit .

(continued)

LaSalle 1 and 2 B 3 . 3 . 5. Ji -lO Revision

RCIC System Instrumentation B 3 . 3 . 5. Ji BASES SURVEILLANCE SR 3 . 3 . 5 .Ji . 1 (continued)

REQUIR EMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program . The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO .

SR 3 . 3 . 5. Ji . 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel wil l perform the intended funct i on. A successful test of the req ui red contact(s) of a channel relay may be performed by the verification of t he 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 Speci fi catio ns and no n-Tech ni ca l Spec ifi catio ns t ests at least once per r efue lin g i nte r va l wi t h appl ica bl e extensions. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology .

The Surveillance Frequency is controlled under t he Surveillance Frequency Control Program .

SR 3.3 . 5. Ji . 3 CHA NNEL CALI BRATI ON i s a complete chec k of the in strument loop and the sensor . This test verifies the chan nel responds to the measured parameter with the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint met hodology.

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

(conti nu ed)

LaSalle 1 and 2 B 3.3.5. 3~ -11 Revis i on

RCIC System Instrumentation B 3 . 3 . 5. 3i BASES SURVEILLANCE SR 3 . 3 . 5 . 3i .4 REQUIREMENTS (cont i nued) The LOGIC SYSTEM FUNC TIONAL TEST demonstrates the OP ERABILITY of the required in i tiation logic for a specific channel . The system f unctional testing performed in LCO 3 . 5. 3 overlaps this Surveillance to provide complete testing of the safety function .

The Sur veil l ance Frequ ency is contro ll ed under t he Sur veil lance Frequency Control Progra m.

REFERENC ES 1. GENE-770-06-2-A, "Addendum to Bases for Changes to Survei llan ce Test Inter val s and Al l owed Out-of-Se rvi ce Times fo r Selec t ed Instr umentat i on Technica l Specifications, " December 1992.

LaSal l e 1 and 2 B 3.3.5 . 3i-12 Revision

Primary Containment Isolation Instrumentation B 3.3.6 . 1 BASES APPLICABLE Allowable Value . Trip setpoints are tho se predetermined SAFETY ANALYSES, values of output at which an action should take place. The LCO , and setpoints are compared to the actual proce ss parameter APPLICABILITY (e .g., reactor vessel water level ) , and when the measured (continued) output value of the process parameter exceeds the setpoint, the associated device (e.g. , trip unit) changes state . The analytic limits are derived from the limiting values of the process parameters obtained fro m the safety analysis . The trip setpoints are determined from the analytic limits ,

corrected for defined process, calibration, and instrument errors . The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors. These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components . The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, calibration tolerances, instru ment drift, and severe env ir onment e rror s (for channe l s that mu st fu nct i on in harsh environ ments as defined by 10 CFR 50 . 49) are accounted for and appropriately applied for the instrumentation.

Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e . g ., minimum flow) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS and RCIC. Some instrumentation and ACTIONS associated with these signals are addressed in LCO 3 . 3.5 . 1, "ECCS Instrumentation," and LCO 3 . 3 . 5 . 3i , "RCIC System In st rumentation," and are not in cluded in this LCO .

In general, the individual Functions are required to be OPERABLE in MODES 1, 2, and 3 consistent with the Applicability for LCO 3 . 6 . 1.1, "Primary Containment . "

Functions that have different Applicabilities are discussed below in the individual Functions discussion .

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

(continued)

La Salle 1 and 2 B 3 . 3 . 6 . 1-7 Revision

Primary Containment Isolati on Instrumentation B 3.3. 6. 1 BASES APPLICABLE 5. a . Reactor Ve sse l Water Level-Low . Level 3 (c ontinued )

SAFETY ANALY SES, LCO, and fuel may be threatened . Should RPV water level decrea se too APPLICABILIT Y far, fuel damage could result . Therefore, i solation of som e reactor vesse l interfaces occurs to begin i solating the potential sources of a break . The Reactor Vessel Water Level-Low, Level 3 Function associated with RHR Shutdown Cooling Syste m iso l at i on is no t directly assumed in any tra ns i ent or accident analys i s, s i nce boundi ng ana l yses are performed f or l arge brea ks suc h as MSLB s . Th e RH R Shu tdown Cooling System isolation on Level 3 supports actions to ensure that the RPV water level does not drop below the top of the act i ve fuel during a vessel draindown event caused by a leak (e . g. , pipe break or in adverten t valv e open in g) in t he RHR Shutd own Coo l in g Syste m.

Reactor Vessel Water Level-Low, Level 3 signals are initiated from differential pressure transmitters that sense t he difference between the pressure due to a constan t column of water ( r ef erence l eg) and the press ure du e to t he actual wat er l evel ( variab l e l eg) i n t he vesse l . Four chann els (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 funct i on . As noted (foot note Cc) t o Table 3 . 3. 6. 1 1), on l y one tr i p syste m is reqyired to be OPERABLE in MODES 4 and 5 provided the RMR ShYtdown Cooling System integrity is maintained. System integrity is maintained provided the piping is intact and no maintenance is being performed that has the potent i al for draining the reactor vessel throygh the system .

The Reactor Vessel Water Level-Low , Level 3 Function is only required to be OPERABLE in MODE& 3, 4, and 5 to prevent this potential flow path from lowering reactor vessel level to the top of the fue l. In MOD ES 1 and 2 , the Reactor Vessel Pressure-High Function and admin i strative contro l s ensure that th i s flo w pat h re mai ns isolated to pre vent unexpected loss of inventory via this flow path.

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 Al l owable Va lu e (LCO 3 . 3 . 1 . 1) s i nce the capability to cool t he fuel may be threatened .

This Function isolates the Group 6 valves .

(continued)

LaSalle 1 and 2 B 3 . 3 . 6 . 1-2 8 Rev i sion

Secondary Containment Isolation Instrumentation B 3. 3. 6. 2 BASES APPLICABLE 1 . Reactor Vessel Water Level-Low Low , Level 2 SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from differential pressure transmitters that sense the difference between the pressure due to a constant column of water (ref erence leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vesse l Water Level-Low Low, Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Low, Level 2 A11 owabl e Value was chosen to be the same as the High Pressure Core Spray (HPCS)/Reactor Core Isolation Cooling (RCIC) Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LCO 3.3.5.1, "Emergency Core Cooling System (ECCS)

Instrumentation," and LCO 3 . 3 . 5. 3.i , "Reactor Core Isolation Cooling (RCIC) Syste m Instru ment ation " ) , since t his co ul d indicate the capability to coo l 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 the se MODES; thus, this Function is not required. In addition, the runction is also required to be OPER/\BLE during operations *.1ith a potential for draining the reactor vessel (OPDRVs) to ensure that offsite dose limits are not exceeded if core damage occurs .

2. Drywell Pressure-Hi gh High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary con tainment and actuation of the SGT System are init i ated in order to minimize the potential of an offsite dose release . The iso l ation and initiation of systems on Drywell Pressure-High supports actions to ensure that any offsite releases are within the limits calculated in the (continued)

LaSalle 1 and 2 B 3.3 . 6 . 2-4 Revision

Secondary Containment Isolation Instrumentation B 3 . 3 . 6. 2 BASES APPLICABLE 3 , 4 . Reactor Building Ventilation Exhaust Plenum and Fuel SAFETY ANALYSES, Pool Ventilation Exhaust Radiation-High (continued)

LCO, and APPLICABILITY Reactor Building Ventilation Exhaust Plenum Radiation-High signals are initiated from rad i ation detectors that are located in the reactor building return air riser above the upper area of the steam tunnel prior to the reactor building ventilation isolation dampers . Fuel Pool Ventilation Exhaust Radiation-Hig h signals are initiated from radiation dete ctors that are located in the reactor building exhaust ducting coming from the refuel floor. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel . Four channels of Reactor Building Ventilation Exhau st Plenum Radiation - Hi gh Func ti on and four channels of Fuel Poo l Ventilation Exhaust 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 detec t gro ss failure of the fuel cl adding .

The Reactor Building Ventilation Exhaust Plenum and Fuel Pool Ventilation Exhaust Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probab ili ty 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 req uired . In addition, the Functions are required to be OPERABLE during CORE ALTE RATIONS , OPORVs, and movement of 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.

5. Manual Initiation The Manual Initiation push button channels introduce s ign als into the secondary containment isolation logi c that are redundant to the automatic protective instrumentation channels, and provide manual isolation capability . There is (continued)

LaSalle 1 and 2 B 3 . 3 . 6. 2-6 Revision

Secondary Containment Isolation Instrumentation B 3 . 3 . 6. 2 BASES APPLICABLE 5. Manual Initiation (continued)

SAFETY ANALYSES, LCO, and no specif i c UFSAR safety analysis that takes credit for this APPLICABILITY Function . It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis .

Th ere is one manua l in itiation pus h butto n for the logic per trip system. Two channels of the Manual Initiation Function are available and are required to be OPERABLE in MODES 1, 2, and 3 and during CORE ALTERATIONS , OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, s i nce these are the MOD ES and other specified cond i t i ons in which the Secondary Containment Isolation automatic Functions are required to be OPERABLE . There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons .

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, compone nt s, or var i ables expressed in t he Condition di scovered to be inoperable or not within li mits 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 Comp l etion Times based on i nitial entry into the Conditio n. However, th e Required Actions for inoperab l e secondary containment i solation instrumentation channels provide appropriate compensatory measures for separate inoperable channels . As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containme nt isolation instrumentation channel .

A. 1 Because of the diversity of sensors available to provide i solation s i gnals and t he redu ndancy of the isolat i on design, an al lowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , depending on the Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those (continued)

LaSalle 1 and 2 B 3.3 . 6 . 2-7 Revision

CRAF System Instrumentation B 3 . 3.7.1 BASES LCO Each channel must have its setpoint set within the specified (continued) Allowable Value of SR 3.3.7.1.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Nominal trip setpoints are specified in the setpoint calculations . These nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint that is less conservative than the nominal trip setpoint , but wit hi n its Allowab l e Value, i s acceptable . A channel is inoperable if its actual trip setpoint is not within its required Allowable Value .

Trip setpoints are those predetermined values of output at whic h an act i on sho uld take pl ace . The setpo in ts are compared to t he actual process parame t er (e.g., control room air intake radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e . g., trip unit) changes state . The analytic limits are derived from the limiting values of the process para meters obta i ned f rom the sa f ety anal ysis . Th e trip setpoi nts are determined from the ana l ytic lim it s , corrected for defined process, calibration, and instrument errors.

The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors . These calibrat i on based errors are limited to refere nce acc ur acy, in st r ument drift , err ors assoc i ate d wi th measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, ca l ibration to l erances, instrument dri f t , and severe enviro nment errors (for channe l s that must function in harsh environments as defined by 10 CFR 50 . 49) are accounted for and appropriately applied for the instrumentation .

APPLICABILITY The Control Room Air Intake Radiation-High Funct i on is required to be OPERABLE in MODES 1, 2, and 3, and during CORE ALTERATIONS , OPDRVs, and movement of irradiated fuel in the secondary containment to ensure that control room personnel are protected during a LOCA or a fuel handling event , or a vessel drai ndo1m event . During MODES 4 and 5, when these specified conditions are not i n progress (e . g.,

CORE ALTERATIONS), the probability of a LOCA or fuel damage is low; thus, the Function is not required .

(continued)

LaSalle 1 and 2 B 3.3.7.1-3 Revision

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE l.c . l.d. l.e. 2.c. 2.d. 2.e. 4.16 kV Emergency Bus SAFETY ANALYSES, Undervoltage (Degraded Voltage) (continued)

LCO , and APPLICABILITY generated by either the Reactor Vessel Water Level - Low Low, Level 2 or Drywell Pressure - High ECCS Instrumentation . The required OPERABILITY of this instrumentation is ide ntif ied on Table 3 . 3 . 5.1- 1, "Emergency Core Cooling System Instrumentation . " Two foot note s have been provided for the Degraded Voltage Time Delay, LOCA, Function to modify its OPERABILITY consistent with the OPERABILITY requirements of the ECCS Instrumentation that generate the associated LOCA sign al . Per footnote (a), the Degra ded Voltage Time Delay, LOCA, Function is not requ i red to be OPERABLE in MODES 4 and 5 sinceWfl.-fl- the associated ECCS subsystems 4--&are not required to be OPER/1BLE for automatic initiationinitiate automatically in MODES 4 and 5 .

Additionally, footnote (b) states the Degraded Voltage Time Delay, LOCA, Function is not required to be OPERABLE when the reactor vessel i s defueled. These footnotes are acceptable because the Degraded Voltage Time Delay, No LO CA, Function provides adequate protection to ensure that other required systems powered from the DG(s) function as designed in any non-LOCA accident in which a loss of offsite power is assumed.

ACTIONS A Note has been provided to modify the ACTIONS related to LOP instrumentation channels . Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsyste ms, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not resu lt 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 LOP instrumentation chann els provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.

(continued)

LaSalle 1 and 2 B 3.3.8.1-6 Revision

RPS Electric Power Monitoring B 3 . 3. 8 . 2 BASES (continued)

APPLICABILITY The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS bus powered components from the i nservice MG set or alternate power supply during abnormal voltage or frequency conditions.

Since the degradation of a nonclass lE source supplying power to the RPS bus can occur as a result of any random single failure , the OPERABILITY of the RPS electric power mon i toring asse mblies i s req uired whe n the RPS bus powered co mpo nents ar e r equired to be OPE RAB LE. This r esults in the RPS Electric Power Mo ni toring System OPERABILITY being required in MODES 1, 2, and 3, MODES 4 and 5, with residual heat removal (RHR) shutdown cooling isolation valves open, MODE 5 with any control rod withdrawn from a core cell conta i ning one or more fuel asse mblies , during movement of i rra dia ted f uel asse mbli es i n th e seco ndary co nta i nme nt , and during CORE ALTERATIONS , and dYring operations with a potential for draining the reactor vessel COPDRVs) .

ACTIO NS A .1 If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inserv i ce power supp l y i s ino perable , t he OPERAB LE asse mbly wi l l still pro vi de protection t o the RPS bus po wered components under degraded voltage or frequency conditions .

However, the reliability and redundancy of the RPS Electric Power Monitoring System are reduced and only a limited time (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) is allowed to restore the inoperab l e asse mbly(s) to OPERABL E status . If the i noperable assemb l y(s) ca nn ot be resto red to OPE RABLE status, t he assoc i ated power supp l y must be removed from service (Required Action A. l) . This places the RPS bus in a safe condition . An alternate power supply with OPERABLE power monitoring assemblies may then be used to power the RPS bus.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time takes into account the rema1n1ng OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS Electric Power Monitoring protection occurring during this period. It allows time for plant operations personnel to take corrective act i ons or to place the plant in the required condition in an order l y manner and wit hout chal l enging pl ant systems .

(continued)

LaSalle 1 and 2 B 3.3 . 8 . 2-4 Rev i s i on

RPS Electric Power Monitoring B 3. 3.8 . 2 BASES ACT IONS Ll (continued) perfor m the nece ssary repairs to restore the system to OPERABLE status will be short . However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state . The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly ma nner and wit hout cha l lenging plant systems .

D.l and D. 2 If any Require d Action and associated Completion Time of Condition A or Bare not met in MODE 4 or 5 with RHR SDC isolation valves open, action must be immediately initiated to either restore one electric power monitoring assembly to OPERABLE status for the inservice power source supplying the required instrumentation powered from the RPS bus (Required Action D. l) or to iso l ate the RHR SOC System (Required Action 0. 2). Required Action D.l is provided becau se the RHR SDC System may be needed to provide core cooling. All actions must continue until the applicable Required Actions are completed .

.L1 If any Required Action and associated Completion Time of Condition A or Bare not met in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assembl i es, the operator mu st i mmediate l y initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies (Required Action E.l). Thi s Required Action results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g. , scra m of co ntrol rods) i s not required.

F.1.1 F.1.2 F.2.1 and F.2.2 If any Required Action and associated Completion Time of Condition A or Bare not met during movement of irradiated fuel assemblies in the secondary containment, or during CORE ALTERATION S, or d~rin9 OPDRVs, the ab ilit y to isolate the (continued)

La Sal le 1 and 2 B 3.3 .8.2-6 Revision

ECCS-Operating B 3 . 5. 1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) , REACTOR PRESSURE VESSEL (RPV)

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

SYSTEM B 3.5.1 ECCS-Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and seco ndary co ntainm ent, to limit the re l ease of radi oact iv e materials to t he env ir onment following a l oss of coolant accident (LOCA) . The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA . The ECCS network is composed of the High Pressure Core Spray (HPCS) Syste m, the Low Pressure Core Spra y (L PC S) System ,

and the l ow pressure coo lant inje ctio n ( LPCI) mode of t he Residual Heat Removal (RHR) System. The ECCS al so consists of the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS.

On receipt of an initiation signa l, ECCS pumps automatically start ; the system aligns, and the pumps inject water, taken from the suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of t he ECCS pumps. Although the system is initiated, ADS action i s de l ayed, al l owing t he operator to interrupt th e timed sequence if the system is not needed . The HPCS pu mp discharge pressure almost immediately exceed s that of the RCS , and the pump injects coolant into the spray sparger above the core . If the break is smal l, HPCS will maintain coo lant inventory, as well as vessel lev el, while the RCS is sti ll pr ess urized. If HP CS fails, it i s backed up by ADS in combination with LPCI and LPCS. In this event, the ADS timed sequence would be allowed to time out and open the selected safety/relief valves (S/RVs), depressurizing the RCS and allo wing the LP CI and LPCS to ove rcome RCS pressure and in ject coolant into the vessel . If the break is larg e, RCS pressure initi al l y dr ops rapidly, and the LPCI and LP CS systems cool the core.

Water from the break returns to the suppression pool where it is used aga i n and again. Water in th e suppression pool is circulated through a heat exchanger cooled by the Residual Heat Removal Ser vi ce Wat er (R HRSW) System .

Depending on the location and size of the break, portions of (continued)

LaSalle 1 and 2 B 3 . 5.1 -1 Revision

ECCS-Operating B 3.5.1 BASES (co ntinued)

APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerab le energy in the reactor core and core cooling wou l d be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, the ADS function is not required when pressure i s~ 150 psig because the low pressure ECCS subsystems (LPCS and LPCI) are capable of providing flow i nto the RPV below t hi s pres sure. EGGS rRequirements for MODES 4 and 5 are specif i ed in LCO 3.5.2, " RPV Water Inventory Control EGGS Shutdo*,m ."

ACT IONS A Note prohibits the application of LCO 3.0 . 4.b to an inoperable HPCS subsys tem. There is an increa sed ris k associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCS subsys tem 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, sho uld not be applied in this circumstance .

A .1 If any one low pressure ECCS injection/spray subsys tem i s inoperable, the inoperable subsystem 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 reliability is reduced because a s ingle failure in one of the remaining OPERABLE subsystems concurrent with a LOCA may result in the ECC S not being able to perform its intended safety function. Th e 7 da y Completion Time is based on a reliability study (Ref. 12) that evaluated the impact on ECCS availability by assuming that various components and subsystems were taken out of serv i ce . The results were used to calculate the average availability of ECCS equipment needed to mitigate t he conse quences of a LOCA as a function of allowed outage times (i.e . , Completion Time s) .

8.1 and B.2 If the HPCS System i s inoperable, and the RCIC Syste m is immediately verified to be OPERABLE (when RCIC is required to be OPERABLE), the HPCS System must be restored to OPERABLE status within 14 days. In this Condition, adequate (continued)

LaSalle 1 and 2 B 3.5 . 1-6 Revision

RPV Water Inventory Contra/EGGS ShutdO'.\/n B 3 . 5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) , REACTOR PRESSURE VESSEL (RPV)

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

SYSTEM B 3.5 . 2 EGGS ShutdownRPV 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 the active irradiated fuel at all times to prevent such elevated cladding temperatures.A description of the ~igh PressHre Core Spray CMPCS) System, Lo*.1 PressHre Core Spray (LPCS) System, and loH pressHre coolant injection (LPCI) mode of the ResidHal ~eat Removal (RMR) System is provided in the Bases for LCD 3.5.1, "EGGS Operating . "

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is SAFETY ANALYSES not required to mitigate any events or acidents evaluated in the safety analyses.The EGGS performance is evaluated for the entire spectrum of break sizes for a postulated loss of coolant accident (LOG/\). 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.Hie long term cooling analysis follo*,1ing a design basis LOCA (Ref . 1) demonstrates that only one EGGS injection/spray SHbsystem is reqHired, post LOCA, to maintain adeqHate reactor vessel *.:ater 1 evel in the event of an inadvertent vessel draindol.'n. It is reasonable to assHme, based on engineering jHdgment, that while in MODES 4 and 5, one EGGS injection/spray sHbsystem can maintain adeqHate reactor vessel 1,1ater 1evel . To provide redHndancy, a mini mHm of t*.:o EGGS injection/spray SHbsystems are reqHired to be OPERABLE in MODES 4 and 5.

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 (continued)

LaSalle 1 and 2 B 3.5.2-1 Revision

RPV Water Inventory Contra/EGGS Shutdo1,*n B 3.5.2 BASES APPLICABLE path with the highest flow rate, or the sum of the drain SAFETY ANALYSES rates through multiple penetration flow paths susceptible to (continued) a common mode failure (e.g. , se i smic event, loss of normal power, single human error) . It is assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection / spray subsystem can maintain adequate reactor vessel water level.The EGGS satisfy Criterion 3 of 10 GfR 50 . 36(c)(2)(ii).

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 act i ve irradiated fuel as required by Safety Limit 2. 1. 1.3.

The Limiting Condition for Operation (LCO) requires the DRAIN TIME of RPV water inventory to the TAF to be

~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A DRAIN TIME of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate unexpected draining of reactor coolant . An event that could cause l oss 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.

Onel-.w ECCS injection/spray subsystem-& i s~ required to be OPERABLE and capable of being manually started to provide defense-in-depth should an unexpected draining event occur .

An.:i:.h--e ECCS injection/spray subsystem-& is~ defined as either one of the three Low Pressure Coolant Injection

( LPCI ) subsystems, the Low Pressure Core Spray ( LPCS )

System, or-a-oo th e High Pressure Core Spray ( HPCS ) System.

The LP CJ.S. System subsystem and the-a-&J:l. LP CS+ 1:1-9-&System consist of one motor driven pump, piping, and valves to transfer water from the suppression pool to the reactor pressure vessel ( RPV ) . The HPCS System consists of one motor driven pump, piping, and valves to transfer water from the suppression pool t o the RPV. The necessary portions of the Diesel Generator Coo l ing Water System are also requ i red to provide appropriate cooling to each required ECCS (continued)

LaSalle 1 and 2 B 3.5.2-2 Revision

RP V Water Inventory Cont ro l'i.GGS Sh 1:1td 01m B 3 . 5. 2 BASES LCO injection/spray subsystem. Management of gas voids is (continued) important to ECCS injection/spray subsystem OPERABILITY .

A required ECCS subsystem may be aligned with the pump control switch in pull - to - lock and associated ECCS subsystem injection valves may be configured to allow throttling to control RPV makeup flow rates. Operators must be able to take manual action from the control room to provide makeup to the RPV as-necessary with the pump and associated injection valve in this alignment without delay.

The LCD is modified by a Note which allows a required A-&

noted, one LPCI subsystem (A or B) toflttl-;Y- 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:

a) when the system is realigned to or from the RHR shutdown cooling mode and; b) when the system is in the RHR s hutdown cooling mode, whether or not the RHR pump is operating.

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, low press1:1re and low temperat1:1re conditions in MODES 4 and 5, sufficient time will be available to manually align and operate the required initiate LPCI subsystem operation to maintain RPV inventory provide core cooling prior to RPV water level reaching the TAFpost1:1lated f1:1el 1:1ncovery .

APPLICABILITY RPV water inventory control OPER/\BILITY of the EGGS injection/spray s1:1bsystems is required in MODES 4 and 5-t-B-ens1:1re adeq1:1ate coolant inventory and s1:1fficient heat removal capability for the irradiated fl:lel in the core in case of an inadvertent draindown of the vessel .

Requirements on water inventory control for EGGS OPER/\BILITY during other MODES 1, 2, and 3 are disc1:1ssed contained in LCOs in Section 3.3, Instrumentation, and other LCOs in Section 3.5, ECCS, RPV Water Inventory Control, and RCIC----t-R-e

/\pplicability section of the Bases for LGO 3 . 5. 1 . 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. ECC~ si,esystems are Rat re(li,i res ta ee OPfRABLf s1,riR9 MOQE § wit~ t~e s~eRt fi,el stera9e ~eel 9ates removes aRs (continued)

LaSal l e 1 and 2 B 3.5.2-3 Revision

RPV Water Inventory ControlEGGS Sh1:.1tdo*,m B 3.5.2 BASES (continued)

ACTIONS A. land B. l the water level maintained at> 22 ft above the RPV flange. This provides syfficient coolant in*,entory to allo11 operator action to terminate Hie inventory loss prior to f1:.1el 1:.1ncovery in case of an inadvertent draind01m .

The A1:.1tomatic Depress1:.1rization System is not re~1:.1ired to be OPERABLE dYring MODES 4 and 5 because the RPV pressure is

< 150 psig, and the LPGS, ~PCS, and LPGI s1:.1bsystems can provide core cooling *.1ithout any depress1:.1rization of the primary system . If the required ECCS injection / spray subsystem is inoperable, it must be restored to OPERABLE s tatus within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In this Condition, the LCO controls on DRAIN TIME minimize the possibility that an unexpected draining event could necessitate the use of the ECCS injection / spray subsystem, however the defen se - in -depth provided by the ECCS injection / spray sub sy s tem 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 ECCS injection / spray subsystem to OPERABLE s tatus is based on engineering judgment that considers the LCO controls on DRAIN TIME and the low probability of a an unexpected draining event that wou Id re su It in a I oss of RPV water inventory. If any one reqYired [CCS injection/spray sybsystem is inoperable, the reqyired inoperable [CCS injection/spray sybsystem myst be restored to QP[RABL[

statys 1.1itt:1in 4 ROYrs. In tRis Condition, tRe remaining QP[R.I\BL[ sybsystem can provide syfficient RPV flooding capability to recover from an inadvertent vessel draindo1.1n . Mo1.1ever, overall system reliability is redYced becayse a single failyre in the remaining QP[RABL[ sYbsystem conCblrrent 11ith If the i noperab I e ECCS injection / spray sub system is not restored to OPERABLE statu s within 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 operated manually, and may consist of one or more systems or subsystems, and must be able to acces s water inventory capable of maintaining the RPV water level above the TAF for ~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . If recirculation of injected water would occur, it may be credited in determining the necessary water volume.a vessel draindo*.*n coblld resYlt in the [CCS not being able to perform its intended fynction. The 4 hoblr Completion Time for restoring the reqblired [CCS injection/spray sybsystem to QP[RABL[

statbls is based on engineering jydgment that considered the availability of one sybsystem and the 10 1,1 pr obability of a vessel draindoHn event.

(continued)

LaSal l e 1 and 2 B 3.5.2-4 Rev i sion

RP V Water Inventory Cont ro 1EGGS Shutdo1m B 3 .5. 2 BASES ACT IONS C. l. C. 2 . 0.1 0. 2 and C~ . 3 (co ntinued )

With the inoper able subsystem not restored to OP~RABL~ st at us 11 ithin the required Completion Time, action must be init i ated immediate l y to suspend operations *.11th a potential for drainin9 the reactor vessel COPDRVs) to minimize the probability of a vessel draindo*..*n and the subsequent potential for fission product release . Actions must continue until OPDRVs are suspended .

With the DRAIN TIME less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months but greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , compensatory measures should be taken to ensure the ability to implement mitigating actions should an unexpected draining event occur. Should a draining event 1ower the reactor coo I ant 1eve 1 to be 1ow the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material . Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

The secondary contai nment provides a controlled vo lume in wh i ch fission products can be contained, di luted, and processed prior to release to the environment. Required Action C.1 requires verification of the capability to establish the secondary containment boundary in less than the DRAIN TIME . The required verification confirms actions to establish the secondary containment boundary are preplanned and necessary materials are available. The secondary containment boundary is considered established when one Standby Gas Treatment (SGT) subsystem is capable of maintaining a negative pressure in the secondary containment with respect to the environment .

Verification that the secondary containment boundary can be established must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment. Secondary containment penetration flow paths form a part of the secondary containment boundary . A secondary containment penetration flow path can be considered isolated when one barrier in the flow path is in place . Examples of suitable barriers include, but are not limited to, a closed secondary containment isolation valve (SCIV), a closed manual valve, a blind flange, or another sealing device that sufficiently seals the penetration flow path. The planned actions are not required to restore secondary containment to an OPERABLE status, only sufficiently sealed to allow one division of (continued)

LaSalle 1 and 2 B 3.5 . 2-5 Revision

RPV Water Inventory Contra I EGGS Shutdoi,.m B 3.5.2 BASES ACT IONS C.l . C.2. D.l D.2 and C~ .3 (continued)

SGT to maintain a negative pressure with respect to the environment. 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 dampers 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. If both of the required EGGS injection/spray subsystems are inoperable, all coolant inventory makeup capability may be unavailable . Therefore, actions must be initiated immediately to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release .

Actions must continue until OPDRVs are suspended . One EGGS injection/spray subsystem must also be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time to restore at least one required EGGS injection/spray subsystem to OPERABLE status ensures that prompt action 11i 11 be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

If at least one required EGGS injection/spray subsystem is not restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE; one standby gas treatment subsystem is OPERABLE; and secondary containment isolation capability is available in each secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity 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 (continued)

LaSalle 1 and 2 B 3.5 . 2-6 Re vision

RPV Water Inventory Control't.CCS Shutdovm B 3.5.2 BASES ACTIONS C.l . C.2. 0. 1 0. 2 and C~ .3 (continued) subsyst em 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.releases (i.e., sne secsnaary csntainment isslatisn valve ana asssciateG instr~mentatisn are OPERABLE sr stAer acceptable aaministrative csntrsls ts ass~re isslatisn capability .

TAe aaministrative csntrsls csnsist sf statisning a aeaicateG speratsr, *.1As is in csnti n~s~s csmm~ni cati sn 'di tA tAe csntrsl rssm at tAe csntrsl s sf tAe i ssl ati sn aevi ce. In tAi s *.1ay, tAe penetratisn can be rapialy isslatea \JAen a neea fsr secsnaary csntainment isslatisn is inaicatea.) TAis may be perfsrmea by an aaministr i tive cAeck, by examining legs sr stAer infsrmatisn, ts aetermine if tAe csmpsnents are s~t sf service fsr maintenance sr stAer reassns. It is nst necessary ts perfsrm tAe s~rveillances neeaea ts aemsnstrate OPERABILITY sf tAe csmpsnents . If, Aswever, any req~ i rea csmpsnent is insperable, tAen it m~st be restsree ts OPERABLE stat~s. In tAis case, tAe s~rveillances may neea ts be perfsrmea ts restsre tAe csmpsnent ts OPERABLE stat~s. Actisns m~st csntin~e ~ntil all req~irea csmpsnents are OPERABLE.

D.l, 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.l requires immediate action to establish an additional method of water injection augmenting the ECCS injection / spray subsystem required by the LCD. The additional method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The Note to Required Action D.1 states that either the ECCS injection / spray subsystem or the additional method of water injection must be capable of operating without offsite electrical power . The additional method of water injection may be manually initiated and may consist of one or more system or subsystems . The additional method of water injection must be able to access water inventory capable of (continued)

LaSalle 1 and 2 B 3.5.2-7 Revision

RPV Water Inventory ControlEGGS Shutdo1m B 3.5.2 BASES ACT IONS D.1. D.2, D. 3, and D.4 (continued) being injected to maintain the RPV water level above the TAF for~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The additional method of water injection and the ECCS injection / spray subsystem may share all or part of the same water sources. If recirculation of injected water would occur, it may be credited in determining the required water volume.

Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material.

Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

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

The seco ndary 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. Examples of manual isolation from the control room could include the use of manual isolation pushbuttons, control switches, or placing a sufficient number of radiation monitor channels in trip from either unit. A secondary containment penetration flow path can be considered isolated when one barrier in the flow path is in place. Examples of suitable barriers include, but are not limited to, a closed secondary containment isolation valve (SCIV), a closed manual valve, a blind flange, or another sealing device that sufficiently seals the penetration flow path. The actions are not required to restore secondary containment to an OPERABLE status, only sufficiently sealed to allow one division of SGT to maintain a negative pressure with respect to the environment .

(continued)

LaSalle 1 and 2 B 3 . 5. 2-8 Re vis i on

RPV Water Inventory Contra/EGGS Shutdo*,.*n B 3.5 . 2 BASES ACT IONS D. l. D.2. D.3 . and D. 4 (con tinued )

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

E. l 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 . Note that Required Actions D. 1, D. 2, 0.3, and D.4 are also applicable when DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLA NCE SR 3. 5. 2.1 and ~R 3.5 . 2.2 REQUIR EMENTS This Survei I lance verifies that the DRAIN TIME of RPV water inventory to the TAF is ~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The period of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate draining of reactor coolant. Loss of RPV water inventory that would result in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1.1.3 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 (continued)

LaSalle 1 and 2 B 3.5 . 2-9 Revision

RP V Water Inventory Cont ro H.CCS S hutdQ1,m B 3.5.2 SURVEILLANCE SR 3 . 5.2 . 1 and SR 3.5.2.2 (continued)

REQUIREM ENTS adjust or verify seating of the blade, the exposed cross-sectional area of the RPV penetration flow path is used.

The definition of DRAIN TIME excludes from the calculation those penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths. A blank flange or other bolted device must be connected with a sufficient number of bolts to prevent draining in the event of an Operating Basis Earthquake. Normal or expected leakage fromclosed 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 transferred to Remote Shutdown, which disables the interlocks and isolation signals.

The exclusion of penetration flow paths from the determination of DRAIN TIME must consider the potential effects of a single operator error or initiating event on items supporting maintenance and testing (rigging, scaffolding, temporary shielding, piping plugs, snubber removal, freeze seals, etc.). If failure of such items could result and would cause a draining event from a closed system or between the RPV and the isolation device, the penetration flow path may not be excluded from the DRAIN TIME calculation.

Surveillance Requirement 3.0.1 requires SRs to be met between performances . Therefore, any changes in pl ant 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.

(continued)

LaSalle 1 and 2 B 3.5 . 2-10 Re vi s i on

RPV Water Inventory Controlt.CCS Shutdo1:n B 3.5.2 BASES SURVEILLANCE SR 3.5.2.2 and SR 3.5.2.3 REQUIREMENTS (continued) The m1n1mum water level of -12 ft 7 in (referenced to a plant elevation of 699 ft 11 in) required for the suppression pool, equivalent to a contained water volume of 70,000 ft 3 , is periodically verified to ensure that the suppression pool wil l provide adequate net positive suction head (NPSH) for the ECCS pumps, recirculation volume, and vortex prevention. With the suppression pool water level l ess than the required limit, all ECCS injection/spray subsystems are inoperable.

The Surveil l ance Frequencies are control l ed under the Surveillance Frequency Control Program .

SR 3.5.2. 43, SR 3.5.2.5, and SR 3.5.2.6 The Bases provided for SR 3.5.1.1 , SR 3.5.1.5, and SR 3.5.1.6 are applicable to SR 3.5.2. 43, SR 3.5.2.5, and SR 3.5.2.6, respectively .

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

SR 3.5.2. 54 Verifying the correct alignment for manual, power operated, and automatic valves in the required ECCS subsystem flow path~ provides assurance that the proper f l ow 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 (continued)

LaSalle 1 and 2 B 3.5 . 2-11 Revision

RPV Water Inventory Controlf.CCS Shutdo1m B 3.5.2 BASES SURVEILLANCE SR 3 . 5.2. 54 (cont i nued)

REQUIREMENTS reposition i n the proper stroke time. This SR does not require any testing or valve manipulation ; rather, it involves verification that those valves capab l e of potentia ll y being mi spositioned are in th e correct pos i tion .

Thi s SR does not ap pl y to valves t hat cannot be in adverte ntly misaligned, suc h as check valves . The Sur vei l la nce Freque ncy i s co nt ro ll ed under t he Surve il lance Fre quency Co ntrol Progra m.

The Surve il lance is modified by a Note wh i ch exempts system vent f l ow pat hs ope ned under adm i nistrat i ve contro l. The adm inistrat i ve cont r ol should be procedu r al ized an d include st ationing a dedicated indivi dua l at the system ve nt flow pa th who i s in conti nuous commun i cation wi t h the ope r ators i n the co nt r ol roo m. Thi s i ndiv i dual wi l l have a meth od to rap idly cl ose the system vent f low path if directed.

SR 3.5.2.6 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 full flow test recirculation line is adequate to confirm the operational readiness of the required ECCS injection/spray subsystem. The minimum operating time of 10 minutes was based on engineering judgement.

The Surveillance Frequency is controlled under the Survei I lance Frequency Control Program.

SR 3. 5.2.7 Verifying that each va Ive credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated RPV water level isolation signal is required to prevent RPV water inventory from dropping below the TAF should an unexpected draining event occur.

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

(continued)

LaSalle 1 and 2 B 3 . 5.2-12 Rev i sion

RPV Water Inventory Controlt.CCS Shutdo 1fJn B 3.5.2 BASES (continued)

SURVEILLANCE SR 3.5 . 2 . 8 REQUIREMENTS (cont i nued) The required ECCS subsystem shall be capable of being manually operated from the main control room . This Surveillance verifies that the required LCPI subsystem, LPCS System, or HPCS System (including the associated pump and valve(s)) can be manually operated, including throttling injection valves, as necessary, to pro vi de additional RPV Water Inventory, if needed, without delay.

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

This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the full flow test line, coolant injection into the RPV is not required during the Surveillance .

REFERENCES 1. Information Notice 84-81, "Inadvertent Reduction in Primary Coolant Inventory in Bailing Water Reactors During Shutdown and Startup," November 1984 . UFSAR, Section 6.3.3.2.

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 .

LaSalle 1 and 2 B 3 . 5 . 2-13 Revision

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS CECCS) , REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING CRCIC)

SYSTEM B 3.5.3 RCIC System BASES BACKGROUND The RCIC System is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level. Under these conditions, the High Pressure Core Spray (HPCS) 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 unit, piping and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the head spray nozzle. A l" H2 purge line is connected from the injection line to the reactor head vent to prevent hydrogen buildup (Ref . 4) . The purge line contains an orifice to minimize RCIC flow bypassing the RPV and ensures that sufficient injection flow is delivered to the RPV .

Suction pip i ng is provided from the condensate storage tank (CST) and the suppression poo l. Pump suction is norma l ly aligned to the CST to minimize injection of suppression pool water into the RPV . However, if the CST water supply is low an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System . The steam supply to the turbine is piped from main steam line B, upstream of the inboard main steam line isolation valve .

The RCIC System i s designed to provide core coo ling for a wide range of reactor pressures, 135 psig to 1185 psig.

Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed . As the RCIC flow increases, the turbine contro l valve is automatically adjusted to maintain design flow. Exhaust steam from the (continued)

La Sal le 1 and 2 B 3 .5.3 - 1 Revision

RCIC System B 3.5.3 BASES BACKGROUND RCIC turbine is discharged to the suppression pool. A full (continued) fl ow test line is provided to route water to the CST or the suppression pool to allow testing of the RCIC System during normal operation without injecting water into the RPV .

The RCIC pump is provided with a minimum flow bypass line, which discharges to the suppression pool . The valve in this li ne automatically opens to prevent pump damage due to overheating when other discharge line valves are cl osed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge line "keep fill" system is designed to maintain the pump discharge line filled with water.

APPLICABLE The function of the RCIC System is to respond to transient SAFETY ANALYSES events by providing makeup coolant to the reactor . The RCIC System is not an Engineered Safety Feature System and no credit i s taken in the safety analyses fo r RCIC System operation. Based on it s contribution to the reduction of overall plant risk, the system satisfies Criterion 4 of 10 CFR 50 . 36(c)(2)(ii) .

LCO Th e OPERAB ILIT Y of the RCIC System provides adequate core cooling such that actuation of any of the ECCS subsystems is not required in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC System has sufficient capacity to maintain RPV inventory during an isolation event . Management of gas voids i s important to RCIC System OPERABILITY.

APPLICABILITY The RCIC System is required to be OPERABLE in MODE 1, and MODES 2 and 3 with reactor steam dome pressure> 150 psig since RCIC i s the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure~ 150 psig, and in MODES 4 and 5, RCIC is not required to be OPERABLE since the ECCS injection/spray subsystems can provide sufficient flow to the ve ssel .

(continued)

LaSalle 1 and 2 B 3 . 5. 3-2 Revision

RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3. 5.3.4 (continued)

REQUIREMENTS assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the reactor steam pressure and flow are adequate to perform the test . The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for the flow tests after the required pressure and flow are reached are sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SRs . The Surveillance Frequency is controlled under the Surveillance Frequency Control Program .

SR 3.5 . 3.5 The RCIC System is required to actuate automatically to perform its design function . This Surveillance verifies tha t with a r equire d system in iti ation signa l (ac t ual or simulated) t he automatic initiation logic of RCIC wi ll cause the system to operate as designed, i . e . , actuation of the system throughout its emergency operating sequence , which includes automatic pump startup and actuation of all automatic valves to t heir required positions. Th i s Surveilla nce also ensures that t he RCIC Sys t em wi ll automatically restart on an actual or simulated RPV low water level (Level 2) signal received subsequent to an actual or simulated RPV high water level (Level 8) shutdown signal, and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5. 3~ overlaps t his Surveillance to provide complete testing of the assumed design function .

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

This SR is modified by a Note that excludes vessel injection during the Surveillance . Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance .

(continued)

LaSalle 1 and 2 B 3.5.3-7 Revision

PC IVs B 3 . 6 . 1.3 BASES (continued)

LCO PCIVs form a part of the primary containment boundary . The PCIV safety function is related to minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during a OBA.

The power operated, automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal. The valves covered by this LCO are l isted with their associated stroke times in the Technical Requirements Manual (Ref . 1) .

The normally closed manual PCIVs are considered OPERABLE when the valves are closed and blind flanges are in place, or open under administrative controls . Normal l y closed auto matic PC IVs whic h are req uired by design (e.g., to meet 10 CFR 50 Appendix R requirements) to be de-activated and closed, are considered OPERABLE when the valves are de-activated and closed . These passive isolation valves and devices are those listed in Reference 1. MSIVs and hydrostatical l y tested valves must meet addit i onal leakage rate requirements . Other PCIV leakage rate s are addressed by LCO 3 . 6. 1.1, "Primary Containment," as Type B or C testing .

Thi s LCO provides assurance that the PCIV s will perform their designed safe ty functions to min im ize t he l oss of reactor coolant inventory and establish the primary containment boundary during accidents .

APPLICABILITY In MODES 1, 2, and 3, a OBA co uld cause a release of radioactive material to primary conta inment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, most PCIVs are not required to be OPERABLE and the primary containment pur9e valves are not required to be normally closed in MOOES 4 and 5. Certain valves are required to be OPERABLE, however, to prevent inadvertent reactor vessel draindown . These valves are those whose associated instrumentation is required to be OPERABLE accordin9 to LCO 3.3 . 6. 1, "Primary Containment Isolation Instrumentation." (This does not include the valves that isolate the associated instrumentation . )

(continued)

LaSalle 1 and 2 B 3.6 .1. 3-3 Revision

PC IVs B 3 . 6.1. 3 BASES ACTIONS lL...l (continued) leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway l~akage of the two devices. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for hydrostatically tested line leakage not on a closed system is reasonable considering the time required to restore leakage by isolating the penetration and the relative importance of the hydro statica lly tested line leakage to the overall containment function . The Completion Time of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s for MSIV leakage allows a period of time to restore the MSIV leakage rate to within limit given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for hydrostatically tested line leakage on a clo sed system is acceptable based on the available water seal expected to remain as a gaseous fission product boundary dur i ng the accident, and, in many cases, t he associated closed syste m. The closed system mu st meet the requirements of Reference 5.

E.1. and E. 2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2 , or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant mu st 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 Ti me s are reasonab l e, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

~.land ~.2 If any Required Action and associated Completion Time cannot be met for PGIV(s) required OPERABLE in MODE 4 or 5, the plant must be placed in a condition in which the LGO does not apply. Action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown (continued)

LaSalle 1 and 2 B 3.6 . 1. 3-9 Revision

PC IVs B 3.6.1.3 BASES (continued)

AGTIO~IS r . l and r.2 (continyed) and sybseqyent potential for fission prodyct release.

Actions mYst continye yntil OPDRVs are sYspended. If syspendin§ the OPDRVs ldOYld resylt in closin§ the residYal heat removal ( RHR) shYtdo1m cool in§ i sol ati on valves, an alternative Reqyired Action is provided to immediately initiate action to restore the valves to OPERABLE statYs.

This alloHs RHR shYtdown coolin§ to remain in service while actions are bein§ taken to restore the valve.

SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR verifies that the 8 inch and 26 inch primary containment purge valves are closed as required or, if open, opened for an allowable reason.

The SR is modified by a Note stating that the SR is not required to be met when the purge valves are open for the stated reasons . The Note states that these valves may be opened for inerting, de -inert ing, pressure control, ALARA, or air quality considerations for personnel entry, or for Surveillances that require the valves to be open, provided the drywell purge valves and suppression chamber purge valves are not open si multaneous l y . This i s required to prevent a bypass path between the suppression chamber and the drywell, which would allow steam and gases from a LOCA to bypass the downcomers to the suppression pool. These primary containment purge val ves are capable of closing in the environment following a LOCA . Therefore, these valves are allowed to be open for limited periods of time . The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.1.3 . 2 This SR verifies that each primary containment isolation manual valve and blind flange that i s located outside primary containment and not locked, sea led, or otherwise secured and is required to be closed during accident conditions, is closed. The SR help s to ensure that post (continued)

LaSalle 1 and 2 B 3.6.1.3-10 Revision

r Suppression Pool Water Level B 3.6 . 2.2 BASES APPLICABLE Suppression pool water level satisfies Criteria 2 and 3 of SAFETY ANALYSES 10 CFR 50.36(c)(2)(ii ).

(continued)

LCO A limit that suppression pool water level be;;,: -4.5 inches and~ 3 inches (referenced to plant elevation 699 ft 11 inches) i s required to ensure that the primary containment conditions assumed for the safety analysis are met. Either the high or low water level limits were used in the safety analysis, depending upon which is conservative for a particular calculation.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant loads on the primary contai nm ent. In MOD ES 4 and 5, the probability and consequences of these events are reduced because of the pressure and temperat ure limitations in these MODES. The requirements for maintaining suppression pool water level within l imits in MODE 4 or 5 is addressed i n LCO 3.5.2,

" RPV Water Inventory Control EGGS Sh1:Jtdo1,m ."

ACTIONS Ll With suppression poo l water leve l outside the limits, the conditions assumed for the safety analysis are not met . If water level is below the min i mum level, the pressure suppression function still exists as long as the downcomers are covered, RCIC turbine exhausts are covered, and S/RV quenchers are covered . If suppression pool water level is above the maximum level, protection against overpressurization sti l l exists due to the margin in the peak containment pressure analysis and the capability of the suppression pool sprays. 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 specified limits . Also, it takes into account the low probability of an event impact i ng the suppression pool water level occurring during this interval.

(continued)

LaSalle 1 and 2 B 3 . 6 . 2.2-2 Revision

Secondary Containment B 3.6 . 4. 1 BASES APPLICABLE structure will be treated by the SGT System prior to SAFETY ANALYSES discharge to the environment.

(continued)

Secondary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(i i ) .

LCO An OPERABLE secondary containment provides a control volume into which fission products that bypass or leak from primary containmen t, or are released from the reactor coolant pressure boundary components located in secondary containment, can be diluted and processed prior to release to the environment . For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained .

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product re l ease to primary co ntainme nt th at leaks to seco nd ary 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 te mperature 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 whic h significant releases of radioactive materia l can be postulated, such as d~rin9 operations with a potential for draining the reactor vessel (OPQRVs), during COR E ALTERATIONST or during movement of irradiated fuel assemblies in the secondary containment .

(continued)

LaSalle 1 and 2 B 3 . 6 . 4 . 1-2 Revision

Secondary Containment B 3.6.4.1 BASES (continued)

ACTIONS 1L...l 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 per i od 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 OPERABI LITY) occurring during periods where secondary containment is inoperable is minimal .

If the secondary containment cannot be restored to OPERABLE status wit hin the required Completion Ti me, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 i s similar to or lower than the risk in MODE 4 (Ref . 3), because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERAB LE status wi l l be short . However, voluntary entry into MODE 4 may be made as it i s also an acceptab l e low-risk state . Th e al lowed Completion Time is reasonable, based on operating experience, to reach the required pl ant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.l , and C.2 , and G.3 Movement of irradiated fuel assemblies i n the secondary containment 7 and CORE ALTERATIONS , and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is t he only barrier to re l ease of f i ssion product s to the environment. CORE ALTERATIONS and movement of irradiated fuel asse mbl ies must be immediate l y sus pe nded if the secondary containment is inoperable .

Suspension of these activities s hall not preclude completing an action that invo l ves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

(continued)

LaSalle 1 and 2 B 3 . 6 . 4.1-3 Revision

~

Secondary Containment B 3.6 . 4 .1 BASES ACT IONS C. l , and C. 2. and C. 3 (continued)

Required Action C. l has been modified by a Note stat in g tha t LCO 3.0.3 is not appl i cable . If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3 .0. 3 would not specify any action. If mo ving irradiated fuel assemblies whil e i n MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of i rradiated fuel asse mblies wou l d not be a s uffi cient reason to require a reactor shutdown.

SURVEILL AN CE SR 3. 6 . 4.1.1 REQUIREMENTS This SR ensures that the seco ndary cont ainment boundary is su ffi ciently leak tight to preclude exfiltration. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program .

SR 3.6.4.1.2 Verifying that one secondary containment access door in each access opening is closed provides adequate assurance that exfiltration from the secondary containme nt will not occur .

An access opening co ntains at l east one inner and one outer door . In some cases a secondary containment barrier contains multiple inner or multiple outer doors. For these cases, the access openings share the inner door or the outer door, i.e., the access openings have a common inner door or outer door . The intent is to not breac h the seco ndary containment, whi ch is achieved by maintaining the inner or outer portion of the barrier closed except when the access opening is being used for entry and exit ; i.e., all inner doors closed or all outer doors closed . Thus each access opening has one door cl osed . Th e Surveil lan ce Frequency is controlled under the Surveilla nce Frequency Contro l Program.

(continued)

LaSalle 1 and 2 B 3 .6 .4 .1 -4 Revision

SC IVs B 3.6.4 . 2 BASES APPLICABLE the boundary established by SCIVs is required to ensure that SAFETY ANALYSES leakage from the primary containment is processed by the (continued) Standby Gas Treatment (SGT) System before being released to the environment.

Maintaining SCIVs OPERABLE with isolation times within limit s ensures that fission products will remain trapped in side secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO SCIVs form a part of the secondary containment boundary . The SCIV safety fun ction i s related to contro l of offsite radiation releases resulting fro m DBAs.

The power operated, automatic isolation valves are considered OPERABLE when their isolation times are within li mits and the valves actuate on an auto mati c isolation signal. The valve s covered by this LCO, along with their associated stroke times, are listed in the Technical Requirements Manual (Ref . 3).

The normally closed manual SC IVs are cons i dered OPERABLE when the va lv es are cl osed and blin d flanges are in pl ace, or open under administrative controls. These passive i solation valves or devices are listed in Reference 3 .

APPLICAB IL ITY In MODE S 1, 2, and 3, a OBA could l ead to a fission produ ct relea se to the primary containment that leaks to the secondary containment. Therefore, OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probabi l ity 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 releases of radioactive material can be post ul ated, suc h as during operations Hith a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONST or during movement of irradiated fuel assemblies in the secondary containment.

(continued)

LaSalle 1 and 2 B 3.6.4. 2-2 Revision

SC IVs B 3 . 6. 4. 2 BASES ACTIONS .8.......1 (cont i nued)

The Condition has been modified by a Note stating that Condition Bis only applicable to penetration flow paths with two isolation valves . This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations .

C.l and C.2 If any Re qui red Actio n and associated Completion Time cannot be met, t he plant must be broug ht to a MODE in whic h t he LCO does not app l y . To ac hieve th i s 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 reac h the required pl ant conditions fro m ful l power conditio ns i n an orderly man ner and without chall enging plant syste ms .

D. l , and D.2 , and D.3 If any Required Action and associated Completion Time cannot be met, the plant must be placed in a condition in wh i ch the LCO does not apply. If applicable, CORE ALTERATIONS and the movement of irradiated fuel assemblies in the secondary containme nt must be im mediately suspended . Suspension of these act i vities shall not preclude completion of move ment of a compone nt to a safe pos i t i on. Also, if applicable, action must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions ~ust continue until OPDRVs are suspended.

Required Action D.l has been modified by a Note stating that LCO 3. 0. 3 is not applicable. I f moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3 . 0. 3 would not specify any action . If mov in g irradiated fuel assemblies while in MODE 1, 2, or 3, t he fuel move ment is independent of reactor operations . Therefore, i n eit her case, i nability to suspend movement of irradia t ed fuel asse mblies would not be a s uf ficient reason to require a reactor s hutdown .

(continued)

LaSal l e 1 and 2 B 3 . 6 . 4. 2-5 Revision

SGT Syste m B 3. 6. 4.3 BASES APPLICAB I LITY other situations under which significant releases of (continued) radioactive material can be postulated, such as dYring operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS 7 or dur i ng movement of i r r adiate d fuel assemb l ies in t he secondary containment .

ACTIONS With one SGT subsys t em i nopera bl e , the i noperable s ubsystem must be restored to OPERABLE sta t us within 7 days . In this condition , the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity re l ease contro l function . However , the overa ll system rel i ability is r educed bec ause a sing le fail ur e in the OPERABLE su bsys t em could result in the radioactiv i ty release control fu nct ion not being adequately performed. The 7 day Completion Time is based on considera t ion of such factors as the availabi lity of the OPERABLE redundant SG T subsystem and the l ow probabi li ty of a OBA occurr in g duri ng t his per i od .

If the SGT subsyste m cannot be restored to OP ERABLE sta t us wit hin the required Co mpletio n Ti me in MO DE 1, 2, or 3, the plant must be brought to a MODE in which the overall pl ant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Re maining i n the App l icability of the LCO is acceptable beca use t he plant ri sk in MOD E 3 is s im ilar to or l ower t ha n the risk in MOD E 4 (Ref . 5) and because the ti me spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short . However, vo l untary entry into MODE 4 may be made as it is al so an acceptab l e low-r i sk s t ate . The al lowed Co mpletion Ti me is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems .

C. l . C.2.1 . and C. 2. 2. and C.2.3 During movement of irradiated fuel assemblies in the secondary containment 7 or during CORE ALTERATIONS , or dYring OPDRVs, wh en Require d Action A. l cannot be co mpleted wi t hin th e required Complet i on Time, the OPERABLE SG T subsys t em (continued)

LaSalle 1 and 2 B 3 . 6 . 4 . 3-3 Rev i sion

SGT System B 3.6 . 4. 3 BASES ACT IONS C. l . C. 2. 1. and C.2 . 2 . and C. 2.3 (continued) should be immediately placed in operation . This Required Action ensures that the remaining subsystem is OPERABLE, th at no f ai l ures t hat could prevent auto ma t i c actuat i on wil l occ ur, and t hat any oth er failure would be readily detected .

An alternative to Re quired Ac t ion C. l is t o immed i at ely s uspend act iv ities t hat represen t a pote nt i al for re l easing ra di oacti ve materia l t o t he seco nd ary co nta inm ent , t hus placing t he unit i n a conditio n t hat min imi zes risk. If app l icable , CORE ALTERATIONS and movement of irradiated fuel assemblies must be i mm ediately suspended . Suspens i on of t hese act i vi ties s hall not prec l ude complet i on of movement of a co mpon ent to a sa f e posi ti on. Also, i f applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindo1m and subse~uent potential for fission product release. Action must continue until OPDRVs are suspended.

Th e Requ ir ed Actions of Condit i on C have bee n modified by a Note stat i ng 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 i r radiated fuel assemblies wh ile in MOD E 1, 2, or 3, the fu el movement is i ndepende nt of reac t or operat i ons . The r ef or e, in eith er case, inab i l ity to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

If both SGT subsyste ms are inoperable in MODE 1, 2, or 3 ,

th e SGT sys t em may not be capab l e of supporting t he required ra dioacti vi ty release contro l fun ct ion . Th e r efore, the pl ant must be broug ht to a MODE in which the overa ll pl ant risk is minimized. To ac hi eve this status, t he 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 . Rema i ning in the Applicab i lity of the LCO is acceptable because t he plant r i sk in MO DE 3 is s im i l ar to or l ower t han the ris k in MO DE 4 (Ref. 5) and bec au se t he time spent in MOD E 3 to perform the necessary repairs to restore the system to OPERAB LE status wi l l be s hort . Ho wever, vol untary entry i nto MOD E 4 may be made as it is also an acce pt able lo w-r i sk s t at e. The all owed Co mpletio n Tim e is reasonable, based on operati ng experience ,

(continued)

LaSalle 1 and 2 B 3.6 .4 .3-4 Rev i sion

SGT System B 3 . 6.4 . 3 BASES ACTIONS lL...l (con tin ued) to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems .

E.l E. 2 and E. 2J When two SG T sub syste ms are inoperable, i f applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secon dary conta inment must be immediatel y suspended.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subseQuent potential for fission product release . Action must continue until OPDRVs are suspended.

Required Action E.l has been modified by a Note stating that LCO 3 . 0 . 3 is not app l icable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3 . 0.3 would not specify any action . If moving ir radiate d fuel assemblies while i n MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Ther efore , in either case, inability to suspend moveme nt of irradiated fuel assemblies would not be a sufficie nt reason to require a reactor shutdown .

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating (from the control room) each SGT subsystem for

~ 15 continuous minutes 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. The Survei llan ce Frequency is controlled under the Surveillance Frequency Control Program.

(co ntinued )

La Sal le 1 and 2 B 3.6. 4. 3-5 Revision

CRAF System B 3.7 .4 BAS ES LCO Additionally, the portions of the Control Room Area HVAC (continued) System that supply the outside air to the EMUs are required to be OPERABLE . This includes the outside air intakes ,

associated dampers and ductwork .

In order for the CRAF subsystems to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analysis for DBAs, and that CRE occupants are protected from hazardous chemicals and smo ke.

The LCO is modified by a Note allowing the CRE boundary to be opened intermittently under administrative controls .

This Note only app lie s to openi ng s in the CRE bounda ry that can be rap idl y restored to the design co nditi on, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, the se cont rol s shou ld be proceduralized and consist of sta ti oning a dedicated individual at the opening who is in continuous communication with the operators in the CRE . This individual will have a method to rapidly close the opening and to restore the CRE boundary to a cond ition equivalent to the design co nditi on when a need f or the CRAF System to be in the pressurization mode of operation is in dicated .

APPLICABILITY In MODES 1, 2, and 3, the CRAF System must be OPERAB LE to ensure that the CRE will remain habitable during and following a OBA, since the OBA could lead to a fission product release .

In MODES 4 and 5, the probability and consequences of a OBA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the CRAF System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

~~~~~~~~~d~ uring movement of ir radiated fuel assemblies in the secondary containmentT and

b. d~uring CORE ALTERATIONS-t---a-A-4.

(continued)

LaSalle 1 and 2 B 3.7.4-4 Re vi sion

CRAF Syste m B 3.7 .4 BASES (continued)

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

ACT IONS Ll With one CRAF subsystem inoperable, for reasons other than an inoperable CRE boundary, the inoperable CRAF subsystem must be restored to OPERABLE status within 7 days. With the uni t in this condition, the remaining OPERABLE CRAF subsystem is adequate to perform the CRE occupant protection fun ction . However, the overall reliability is reduced because a failure in the OPERABLE subsystem could result in los s of CRAF System function. The 7 day Completion Time is based on the l ow probabi l ity of a DBA occurring dur in g this time period, and that the remaining subsys tem can provide the required capabilities.

B.l. B.2 and B.3 If the unfiltered in l eakage 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 DBA co nsequen ces (allowed to be up to 5 rem TEDE), or inadequate protection of CRE occupants from hazardou s che mical s or smoke, the CRE boundary is ino perab le. Actions mu st be taken to restore an OPERABLE CRE boundary within 90 days.

During the period that th e CRE boundary is considered inoper ab le, action must be initiated to imp l ement mitigating actions to le ssen the effect on CRE occupants from the potential hazards of a radiological or chem i cal 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 radiologica l exposures will not exceed th e calcu lated dose of the licensing basis analyses of OBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke . These mitigating actions (i . e., act ions that are taken to offset the consequences of the inoperable CRE boundary) should be preplanned for implantation upon entry into the cond ition, regardless of whether entry is intentional or unintentional.

(co ntinued)

LaSalle 1 and 2 B 3.7.4-5 Revision

CRAF System B 3.7.4 BASES ACT IONS B.l. B.2 and B.3 (continued)

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 miti gating actions. Th e 90 day Completion Time is reasonab le 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 adverse l y affect their abi l ity to co ntrol 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 In MODE 1, 2, or 3, if the inoperable CRA F subsystem or the CRE boundary cannot be restored to OPERAB LE status within the required Completio n Time, the unit must be placed in a MODE that minimizes overall plant risk . To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because th e plant risk in MODE 3 is s imil ar to or lower than the risk in MODE 4 (Re f. 6) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The al l owed Completion Time is reasonable, based on ope r ating experience, to reach the required unit conditions fro m full power conditions in an orderly manner and without challenging unit systems.

D.l. D.2.1. and D.2 . 2. and D. 2. 3 LCO 3 . 0.3 is not applicable while in MODE 4 or 5. However, since irradiated fue l assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition Dare modified by a Note ind i cating that LCO 3. 0 . 3 does not apply. If moving irradiated fuel assemb li es while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require (continued)

LaSalle 1 and 2 B 3. 7.4 -6 Revision

CRA F System B 3.7.4 BASES ACT IONS D.1. D. 2.1. and D.2.2 , and D. 2.3 (continued) the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies . The Note to the ACTIONS, "LCO 3 . 0.3 is not applicable," ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

During movement of irradiated f uel assembl i es in t he secondary containment, or during CORE ALTERATIONS, or during OPDRVs, if the inoperable CRAF subsystem ca nnot be restored to OPERABLE status within the required Completion Time, the OPERABLE CRAF subsystem may be placed in the pressurization mode. This action ensures that the remaining subsyste m is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternat iv e to Required Action D.l i s to immediately suspend act i vities that pre se nt a potential for releasing radioactivity that might require the CRA F System to be in the pressurization mode of operat ion. This places the unit in a cond ition that minimize s the accident risk.

If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be suspended immediatel y . Suspens ion of these activities shall not preclude completion of move ment of a co mponent to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended .

Ll If both CRAF subsystems are inoperable in MODE 1, 2, or 3 ,

for reasons other than an inoperable CRE boundary Ci .e .,

Condition B), the CRA F System may not be capable of performing the intended function . Therefore, the pl ant must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in (co ntinued )

LaSalle 1 and 2 B 3.7.4- 7 Revision

CRAF System B 3.7.4 BASES ACTIONS Ll (cont inued )

MODE 4 (Ref. 6) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short . However, voluntar y entry into MODE 4 may be made as it is also an acceptable low-risk state . The allowed Completion Time is reasonable, based on operating experience, to reach the required plant cond i tions fro m full power conditions is an orderly manner and with out chall enging plant sys tem s .

F. 1. and F.2 , and r".3 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 Required Actions of Condition Fare modified by a Note indicating t hat LCO 3.0.3 does not apply. If moving irr adiated fuel assemblies while in MOD E 1, 2, or 3 , the fuel movement is independent of reactor operations.

Entering LCO 3 . 0.3 while in MO DE 1, 2, or 3 would require the unit to be shut down, but wou ld not requ i re immediate suspension of movement of irradiated fuel assemblies . The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3 . 0. 3.

During movement of irradiated fuel assembli es in t he secondary containment,- or dur i ng CORE ALTERATIONS, -G-F-dYring OPDRVs, with two CRAF s ubs ystems inoperable, or with one or more CRAF subsystems inoperable due t o an in operable CRE boundary, action must be taken immed i ately to suspend activities that present a potential for re l easing radioactivity that might require the CRAF System to be in the pressurization mode of operation. This places the unit in a condition that minimizes the accident risk.

If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary contain ment must be suspended immedia tely . Suspens i on of these activiti es shal l not preclude completion of move ment of a co mp onen t to a safe po sition . If applicable, action myst be initiated immediately to syspend OPDRVs to minimize the probability of a vessel drai ndO'.m and sybseqyent potential for fission prodyct release. Action mYst continYe Yntil the OPDRVs are syspended.

(continued)

LaSalle 1 and 2 B 3.7 .4 -8 Revision

Control Room Area Ventilation AC System B 3.7.5 BASES LCO The Control Room Area Ventilation AC Sys tem is considered (continued) OPERABLE when the individual components necessary to maintain the control room and AEERs temperatures are OPERABLE in both subsystems . These components include the supply and return air fans, direct expansion cool ing coils, an air-cooled condenser, a refrigerant compressor and receiver, ductwork, dampers, and instrumentation and controls.

APPLICABILITY In MODE 1, 2, or 3, the Control Room Area Ventilation AC System must be OPERABLE to ensure that the control room and AEERs temperatures will not exceed equipment OPERABILITY limit s during operation of the Control Room Area Filtration (CRAF) System in the pressurization mode.

In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Contro l Room Area Ventilation AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

-a-.---d~uring movement of irradiated fuel assemblies in the secondary containment and+

b. d~uring CORE ALTERATIONS .-;--a-n4 C. During operations 1,1ith a potential for draining the reactor vessel (OPDRVs).

ACT IONS With one control room area ventilation AC subsystem inoperable, the inoperable control room area ventilation AC subsystem must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE control room area ventilation AC subsystem is adequate to perform the contro l room air conditioning function. However, the overal l reliability is reduced because a single failure in the OPERABLE s ub syste m could result in loss of the control room area ventilation air conditioning function . The 30 day Completion Time is based (continued)

LaSalle 1 and 2 B 3 . 7. 5-3 Revision

Control Room Area Ventilation AC System B 3 . 7. 5 BASES ACTIONS D.l. D. 2.1. and D.2.2 , and D.2.3 (continued)

LCO 3.0.3 is not applicable while in MODE 4 or 5. Howev er ,

since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condit i on Dare modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemb li es while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations .

Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shut down, but would not require immediate suspension of movement of irradiated fuel assemblies . The Note to the ACTIONS, "LCO 3 .0. 3 is not applicable," ensures that the actions for i mmediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0 . 3.

During movement of irradiated fuel assemblies in the secondary containment, or during CORE ALTERATIONS, or durin§ OPDRVs, if Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE control room AC subsystem may be placed immediately in operation .

This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent actuation will occur, and that any active fai lure will be readily detected.

An alternative to Required Action 0.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room . This places the unit in a conditio n that minimizes ri s k.

If applicable, CORE ALTERATIONS and movement of irradiated fuel assemb l ies in the secondary contain ment must be s uspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

(cont inu ed)

La Salle 1 and 2 B 3.7.5-5 Rev i sion

Control Room Area Ventilation AC System B 3 . 7.5 BASES ACTIONS E.l , and E.2 , and E.3 (cont inued )

The Required Actions of Condition E. l are modified by a Note indicating that LCO 3 . 0 . 3 does not apply . If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the secondary containment, or during CORE ALTERATIONS, or dLirin9 OPORVs if Required Action s 8.1 and 8.2 cannot be met within the required Completion Times action must be taken to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes ri sk .

If applicable, CORE ALTERATIONS and hand l ing of irradiated fuel in the secondary containment must be suspended immediately. Suspension of these activities s hall not preclude completion of movement of a co mponent to a safe position. Also, if applicable, action mList be initiated immediately to sLispend OPQRVs to minimize the probability of a vessel draindown and sLibseqLient potential for fission prodLict release. Action mList continLie Lintil the OPQRVs are SLispended.

SURVEILLANCE SR 3.7.5 . 1 REQUIREMENTS Thi s SR monitors the control room and AEER temperatures for indication of Control Room Area Ventilation AC System performance . Trending of control room area temperature will provide a qualitative assessment of refrigeration unit OPERABILITY. Limiting the average temperature of the Control Room and AEER to less than or equal to 85°F provides a thre sho ld beyond which the operating control room area ventilation AC subsystem is no longer demonstrating capability to perfor m its function . This thre sho ld provides margin to temperature l imit s at which equipment qualification requirements cou ld be chal lenged. Subsystem operation is routinely alternated to suppo rt planned maintenance and to ensure each subsystem provides reliable service . The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

La Sal le 1 and 2 B 3.7.5-6 Rev i sion

AC Sources-Shutdown B 3.8. 2 B 3 .8 ELECTRICAL POWER SYSTEMS B 3 .8.2 AC Sources-Shutdow n BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8 .1, "AC Sources-Operating."

APPLICAB LE The OPERABIL ITY of the m1n1mum AC sources during MOD ES 4 SAFETY ANALYSES and 5, and during movement of irradiated fuel assemb l ies in the secondary containment ensures that:

a. The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capabi lity 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 afl-inadvertent draindown of the vessel or a fuel handli ng accident.

In general, when the unit is shutdown the Technical Specificat ion s (TS) requirements ensure that the unit ha s the capab ility to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or lo ss 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 specif ic analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence significantly reduced or eliminated, and minimal consequences. These deviations from OBA analysis assumptions and design requireme nts during s hutdown conditions are allowed by the LCO for required systems .

(continued)

LaSalle 1 and 2 B 3.8.2-1 Revi s ion

AC Sources-Shutd own l

I B 3.8.2 BASES LCO powered from offsite power. An OPERABLE unit DG, associated (continued) with a Division 1 or Division 2 Distribution System emergency bus required OPERABLE by LCO 3.8.8, ensures a diverse power source is available to provide electrical power support, assuming a loss of the offsite circuit .

Similarly, when the High Pressure Core Spray (HPCS) System is required to be OPERABLE, an OPERABLE Division 3 DG ensures a diver se sou rce of power for the HPCS Sys tem is available to provide el ectrical power s upport, assum in g a loss of the offsite power circ uit. Addit i onally, when the Standby Gas Treatment (SGT) Syste m, Control Room Area Filtration (CRAF) System, or Control Room Area Ventilation Air Conditioning System i s required to be OPERABLE, one qualified offsite circuit (normal or alternate) between the offsite transmission network and the opposite unit Division 2 onsite Class lE AC electrical power distribution subsy ste m or an opposite unit DG capable of suppor ting the opposite unit Division 2 ons ite Class lE AC electrical power distribution subsystem i s required to be OPERABLE.

Together, OPERABILITY of the required offsite circuit(s) and DG( s) ensure the availability of sufficient AC sourc es to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g ., fuel handling accidents , reactor vessel draindown ) .

The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective emergency bus(es), and of accepting required load s during an acc ident. Qual ifi ed offsite circuits are tho se t hat are described in the UFSAR and are part of the l icensing basis for the plant. An OPERABLE qualified normal offsite circ uit consists of the required incoming breaker(s) and disconnects from the 345 kV sw it chyard to and including the SAT or UAT (backfeed mode ),

the respective circuit path to and including the feeder breakers to the required Division 1, 2, and 3 emergency buses.

An OPERABLE qualified alternate offsite circuit consists of the required incoming breaker(s) and disconnects from the 345 kV swi tchyard to and including the SAT or UAT (backfeed mode), to and in cl uding the opposite unit 4.16 kV emergency bu s, the opposite unit circuit path to and including the unit tie breakers (breakers 1414, 1424, 2414, and 2424), and the respective circuit path to the required Division 1 and 2 emergency buses.

(continued)

LaSalle 1 and 2 B 3 .8. 2-3 Revision

AC Sources-Shutdown B 3.8.2 BASES LCO The required DG must be capable of starting, accelerating to (continued) rated speed and voltage, and connecting to its respective emergency bus on detection of bus un dervo ltage, and accepting required loads . This sequence must be accomplished within 13 seconds . Each DG must also be capable of accepting required loads within the assumed loading se quence interval s, and mu st continue to operate until offsite power can be restored to the emergency buses.

These capabilities are required to be met from a variety of in itial conditions such as: DG in stand by with the engine hot and DG in standby with the engine at ambient cond ition s .

Additional DG capabilities must be demonstrated to meet required Surveillances, e . g., capability of the Division 1 and 2 DGs to revert to standby status on an ECCS s ignal while operating in parallel test mod e.

Proper se quencing of loads, inc lu ding tripping of nonessential loads, is a required function for DG OPERABILITY . The necessary portions of the DG Cooling Water System and Ultimate Heat Sink capable of providing cooling to the required DG(s) are also required.

It is acceptable for divisions to be cross tied during s hu tdown conditions, permitting a sing le offsite power circuit to supply all required divisions .

APPLICABILITY The AC so urce s required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment provide assurance that:

a. Systems that provide core cooling to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel ;

b. Systems needed to mitigate a fuel handling accident are available; C. Systems necessary to mitigate the effects of events that can lead to core damage during s hutdown are available; and

d. Instrumentation and contro l capabi lity is available for monitoring and maintaining the unit in a co l d shutdown condition or refueling condition .

(continued)

LaSalle 1 and 2 B 3.8.2-4 Revision

AC Sources-Shu td own B 3.8.2 BASES APPLICABILITY The AC power requirements for MODE S 1, 2, and 3 are covered (continued) in LCO 3.8.1 .

ACT IONS LCO 3.0.3 i s not appl i cable while in MODE 4 or 5 . However, since irradiated fue l assembly movement can occur in MODE 1, 2, or 3, t he ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicab l e. 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 move ment is independent of reactor operations. Entering LCO 3 .0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies. The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.

An offsite circuit is considered inoperable if it is not available to one requ i red 4.16 kV emergency bus. If two or more 4.16 kV emergency buse s are required per LCO 3.8.8, di vision(s) with offs i te power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, and fuel movement~

operations 11i th a potential for drai ni n9 the reactor vessel .

By the allowance of t he option to declare required feature s in operab l e that are not capable of being powered from offsite power, approp riate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS.

Required feature s re maining capab l e of being powered from a qualified offsite circuit, even if that circuit is considered inoperable becau se it i s not capable of powering other required features, are not declared inoperab l e by this Required Action. For example, if both Divisi on 1 and 2 emergency buse s are required OPERABLE by LCO 3.8 .8 and only the Division 1 emergency buses are not ca pable of being powered from offsite power, then only the required feature s powered from Divi si on 1 emergency buses are required to be declared in operable.

(continued)

LaSalle 1 and 2 B 3.8.2-5 Revision

AC Sources-Sh utdow n B 3. 8 . 2 BASES ACTIONS A. 2. 1 , A. 2. 2 , A. 2 . 3 , A. 2. 4 B. l , B. 2 , and B. 3 , and 8 . 4 (continued)

With the offsite circuit not available to all required divisions , the option still exists to declare all required features i no perable per Requ i red Act i on A. l . Since this opt i on may inv olve un des i red ad mi nistra tiv e efforts, t he all owance for sufficie ntly conser vative act i ons is made .

Wi th the required DG i noperable , the mi ni mum required diversity of AC powe r sources is not ava il ab l e . It i s ,

th erefore , r equire d to suspe nd CORE ALT ERAT ION S, and mo vement of irradiated fuel asse mblies i n t he seconda ry containment , and activities that coYld potentially resylt in inadvertent draining of the reactor vessel .

Suspensio n of these act ivit i es s hal l not pr ec l ude co mpl etio n of actions to estab l is h a safe conservative conditio n.

These actio ns minim i ze probabi li ty of the occurrence of postulated events. I t is further required to init i ate action immed i ately t o restore t he required AC sources and to co ntinue t hi s act i on unt i l resto r ation i s accomp li s hed in or der to pr ovide the necessary AC power to th e plan t safety systems .

Th e Completion Ti me of immediately is co ns i stent with the r eq uired t i mes for ac ti ons req uir ing pro mpt attentio n. The r estoratio n of the re qui red AC el ectrica l power so ur ces s hould be co mpleted as quickly as possib l e in order to minimize the time during which t he plant safety systems may be without sufficient power .

Pursuant to LCO 3.0 . 6, t he Distr i bution System ACT IO NS are not entered even if all AC sou r ces to it are inoperab l e, resulting in de-energization. Therefore, the Required Actions of Condition A have been modified by a Note to in dicate th at when Co ndition A is entere d wi th no AC power t o any re qui red emer gency bus, AC TI ONS fo r LCO 3 . 8 . 8 must be i mmediate l y entered . Th is Note al l ows Con dition A t o provide requirements for the loss of the offsite circuit whether or not a division is de-energized . LCO 3 . 8.8 provides t he appropr i ate restrict i ons for the situat i on i nvolving a de-energize d divis i on .

(co nt i nued)

LaSal l e 1 and 2 B 3 . 8 . 2-6 Re vision

AC Sources- Shutdown B 3.8 . 2 BASES ACTIONS Ll (continued)

When the HPCS System is required to be OPERABLE , and the Division 3 DG is inoperable, the required diversity of AC power sources to the HPCS System is not available . Since t hese sources only affect the HPCS System, the HPCS System is declare d i noperable and the Required Act i ons of LCO 3 . 5 . 2 , " RPV Water Inventory Contra/Emergency Core Cool i ng Sys terns Shutdo*.m ," entere d.

In the event all so urces of po wer to Division 3 are l os t ,

Condition A will also be entered and direct that t he AC TI ONS of LCO 3 . 8 . 8 be taken . If on l y t he Divisio n 3 DG is inoperable , and power is still supplied to HPCS System, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the DG to OPERABLE . This is reasonab l e considering the HPCS System wi l l still perform its function, absent a loss of offsite power .

Wh en the SGT Syste m, CRAF Syste m, or Cont r ol Room Ar ea Ve ntilatio n Air Condit i oning Sys t em is re quired to be OP ERABLE , and the requ i red opposite unit Division 2 AC source is i noperable, the assoc i ated SGT subsystem , CRAF subsystem, and control room ven ti lation area air conditioni ng subsyste m are declared inoperable and t he Required Actions of the affec t ed LCOs are entered .

The immedia t e Comp l et i on Time i s consiste nt with t he re quired t i mes for actio ns requ i r i ng prom pt attent i on . The restorat i on of the required oppos i te unit Di vision 2 AC el ectrical power source should be completed as qu i ck l y as possible in order to minimize the time during which the aforementioned safety systems are without sufficient power .

SURV EI LL ANCE SR 3 . 8 . 2. 1 REQUIREM ENTS SR 3 . 8 . 2 .1 requires t he SRs f rom LCO 3 . 8 .1 th at are necessary for ensuri ng t he OP ERAB I LITY of t he AC so ur ces in oth er tha n MODES 1, 2, and 3 t o be appl i cab l e . SR 3.8.1 . 8 i s not req uired to be met since only one of fsite circ uit is required to be OPERAB LE . SR 3 . 8 . 1 . 17 is not require d t o be (continued)

LaSal l e 1 and 2 B 3.8.2-7 Re vision

AC Sources-Shutdown B 3.8.2 BASES SURVEILLANCE SR 3 .8 .2.1 (continued)

REQUIREMENTS met because the required OPERABLE DG(s) is not required to undergo periods of being synchron ize d to the offsite circuit. SR 3 . 8.1.20 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE. Refer to the corresponding Ba ses for LCO 3.8 .1 for a discussion of each SR.

Thi s SR is modified by two Notes. The reason for Note 1 is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inope rable during the performance of SRs , and to preclude de-energizing a required 4 . 16 kV emergency bus or disconnecting a requ ir ed offsite circuit during performance of SRs. With limited AC sources availab l e, a sing l e event could compromise both the required circu it and the DG . It i s the intent that these SRs must still be capable of being met, but actual performance is not required during period s wh en the DG and offs it e circuit are requ ired to be OPERABLE.

Note 2 sta tes that SRs 3 . 8.1.12 and 3 .8.1.19 are not required to be met '..'hen its associated EGGS subsystem(s) are not required to be OPERABLE . These SRs demonstrate the DG response to an EGGS initiation signal (either alone or in conjunction >1i th a lass of offsi te po*.1er signal). This is consistent with the EGGS instrumentation requirements that do not require the EGGS initiation signals when the associated EGGS subsystem is not required to be OPERABLE per LGO 3. 5. 2, "EGGS Shutdo*,m."

REFERENCES None.

LaSalle 1 and 2 B 3 . 8 . 2-8 Revision

DC Sources-Shutdown B 3.8 .5 B 3.8 ELECTRICA L POWER SYSTEMS B 3.8.5 DC Sources-Shutdown BASES BACKGROUND A description of the DC sources i s provided in the Bases for LCO 3 . 8 .4, "DC Sources-Operating. "

APPLICA BLE The init ia l co ndit i ons of Desig n Basis Acc i dent and SAFETY ANA LYSES transient anal yses in t he UFSAR , Chapter 6 (Ref. 1) an d Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE . The DC el ectrical power syste m provides normal and emergency DC el ectrica l power for the diesel generators , emergency auxiliaries, and contro l and switching during all MODES of operation and during movement of irradiated fuel assemblies i n the secondary containment.

Th e OPERABIL IT Y of t he DC subsys t ems is co ns i stent with the ini tial ass um ptions of th e acc i dent ana l yses and t he re quiremen t s for the supported systems' OPE RABILITY.

The OPERABI LI TY of the minimum DC electrica l 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 mai ntained in t he shutdown or refue l ing condition for extended periods ;

b. Suffi cie nt ins tr umentation and control capabi l i t y is avai l abl e for mon i toring and maintai ni ng the un i t sta tu s ; and C. Adequate DC electrical power is provided to mi t i gate events postulated during shutdown, such as -a-fl-inadvertent draindown of the vessel or a fuel handling accident .

In genera l, when t he unit is s hu t down, t he Technica l Specificat i ons requi rements ens ure that t he unit has t he capability to mit i ga t e th e co nse quences of postula t ed accidents . However, assu ming a s in gl e fa i l ure and co ncurrent l oss of all offsite or all onsi t e power is not required . The rationa l e for th i s is based on the fac t that many Design Basis Accidents (DBAs) that are analyzed in MODES 1 , 2, and 3 have no specific analyses in MODES 4 (continued)

LaSalle 1 and 2 B 3 . 8 . 5-1 Rev i s i on

DC Sources-Shutdown B 3.8.5 BASES LCO consequences of postu l ated events during shutdown (continued) (e.g., fuel handling accidents and inadvertent reactor vessel drai ndmvn ).

APPLICABILITY Th e DC electrical power sources requ ired to be OPERAB LE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment provide assurance th at:

a. Required features to provide core cooling adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel ;

b. Requi red features needed to mitigate a fu el handli ng accident are avai l able; C. Requ ir ed features necessary to mitiga te the effects of even t s th at can l ead to co re damage duri ng shu td own are avai l able; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown cond ition or refue l ing con diti on.

The DC electrical power require ments for MODES 1, 2, and 3 are covered in LCO 3 . 8 . 4 .

ACTIONS LCO 3.0 . 3 is not app li cable while in MO DE 4 or 5. However, si nce 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 wh ile in MODE 4 or 5 , LCO 3.0 . 3 would not specify any actio n. If moving irradiated fuel asse mblies whi l e in MODE l, 2 , or 3, the fuel movement is independent of reactor operations. Entering LCO 3 . 0. 3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of irradiated fuel assemblies . The Note to the AC TI ONS, "LC O 3.0.3 i s not applicable, " ensures that the act i ons for imm ediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3 .

(co ntinu ed)

LaSalle 1 and 2 B 3.8.5-3 Revision

DC Sources- Shutdown B 3.8. 5 BASES ACTIONS B.l , B.2.1 , B. 2. 2, and B. 2. 3 , and B. 2.4 (cont in ued)

By allowing 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 instance s this option may involve undesired administratiie efforts. Therefore, the allowance for sufficient l y conservative actions is made (i.e., to suspend CORE ALTERA TIONS, and movement of irradiated fuel assemblie s in the secondary conta inment , and any activities that could result in inadvertent draining of the reactor vessel ).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

The se actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accompli s hed in order to provide the necessary DC electrical power to the plant safe ty 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 s hould be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3 .8 . 5.1 requires all Surveil lance s required by SR 3.8 . 4.1 through SR 3.8.4.4 to be applicable. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of each SR .

This SR is modified by a Note. The reason for the Note is t o preclude requiring the OPERABLE DC so urce s from being di scharged below their capability to provide the required power su pp l y or otherwise rendered inoperab l e during the performance of SRs . It i s the i ntent that the se SRs must still be capable of being met, but actual performance is not required.

(continued)

La Sal le 1 and 2 B 3.8 . 5-7 Revision

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 systems is provided in the Bases for LCO 3.8.7, "Distribution Systems-Operating . "

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref . 2), assume Engineered Safety Feature CESF) systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded.

The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent drai ndo,,m of the vessel or a fuel handling accident .

The AC and DC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

(continued)

LaSalle 1 and 2 B 3.8.8-1 Revision

Distribution Sys tems-Shutdown B 3 . 8.8 BASES (continued)

LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of ne cessary support features. This LCO explicitly require s energization of the portions of the electrical distribution system ,

including the opposite unit Division 2 electrical distribution subsystem, necessary to support OPERABILITY of Techn ical Specifications' required syste ms, equip ment, and components-both specifically addressed by their own LCOs, and implicitly required by the definition of OPERABILITY.

Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents and inadvertent reactor vessel draindown ) .

APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment provide assurance that :

a. Systems that provide core cooling to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent drai ndo*,m of the reactor 11essel ;

b. Systems needed to mitigate a fuel handling accident are ava i lable; C. Systems neces sary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capabi lity is available for monitoring and maintaining the unit in a cold shutdown or refueling condition.

The AC and DC electrical power distribution subsystem requirements for MODES 1, 2, and 3 are co vered in LCO 3 . 8.7 .

(continued)

LaSalle 1 and 2 B 3.8.8-2 Revision

Distribut i on Systems-Shutdown B 3.8.8 BASES (continued)

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE l, 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 would not specify any action. If moving i rradiated fuel asse mblies wh il e in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3 .0.3 while in MODE l, 2, or 3 would req uir e the un i t to be s hu tdown, but would not require immediate suspension of movemen t of irrad i ated fuel assemblies . The Note to the AC TIONS, "LCO 3 . 0.3 is not applicable ," ensures that the actions for immediate suspension of irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3 .

A. I , A. 2.1, A. 2. 2, A. 2.3 , and A.2 . 4. and A.2 . 5 Altho ugh redundant required features may require redundant divi sio ns of electrica l power distribution subsystems to be OPERABLE, one OPERABLE di stribut i on subsyste m divis i on may be capable of supporting suffic ient required features to allow cont inuati on of CORE ALTERATIONS, and fuel movement, and operations with a potential for draining the reactor vessel . By all owing the option to declare required feature s associated wi th an inop erable di stributio n subsyste m inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances, this option may involve undesired administrative efforts . Th erefore, the allowance for suffic i ently conservative act i ons is made (i . e., to suspend CORE ALTERATIONS, and movement of irradiated fuel assemblies in the secondary containment---a--R4 any activities that could result in inadvertent draining of the reactor vessel ).

Suspension of these activities sha l l not pre clude co mpletion of actions to establish a safe conservative conditio n.

These actions minimize the probability of the occurrence of postulated events . It is further required to immediately initiate act i on to restore the required AC and DC ele ctrical power distr i bution sub systems and to cont inu e this ac ti on until restoration is accomplishe d in order to provide the necessary power to the plant safety syste ms .

(co ntinued)

LaSalle 1 and 2 B 3.8 . 8-3 Revision

Distribution Systems-Shutdown B 3.8.8 BASES ACTIONS A. l. A.2.1 , A.2 . 2. A.2 . 3 . and A. 2.4 . and A. 2. 5 (continued)

Notwithstanding performance of the above conservative Required Actions, a required residual heat removal-shutdown cooling (RHR -SD C) subsystem may be inoperable. In this case, Required Actions A. 2. 1 through A.2 . 34 do not adequately address the concerns relating to coolan t circulation and heat removal. Pursuant to LCO 3. 0.6, the RHR-SDC ACT IONS would not be entered . Therefore, Required Action A. 2 . 4~ is provided to direct declaring RHR-SDC inoperable, which results in taking the appropriate RHR-SDC ACTIONS.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention . The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power .

SURVEILLANCE SR 3,8.8.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized . The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequen cy Control Program.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 15 .

LaSalle 1 and 2 B 3.8.8 -4 Revision

Inservice Leak and Hydrostatic Testing Operation B 3.10.8 BASES APPLICABL E In the unlikely event of any primary system leak that could SAFETY ANALYSES result in draining of the RPV , the reactor vessel would (continued ) rapidly depressurize , all011ing the 1011 pressure core cooling systems to operate . The make-up capability of the low pressure coolant injection and 101,1 pressure core spray subsystems, as required in MODE 4 by LCO 3 . 5.2, "RPV Water Inventory Contra/EGGS Shutdo*Hn ," would be more than adequate to keep the RPV water level above the TAF---B-F-e-flooded under this low decay heat load condition. Small system leaks would be detected by leakage in spect i ons before s igni fica nt inventory l oss occurred.

For the purposes of these tests, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and during postulated accident conditions.

As described in LCO 3.0. 7, comp lian ce with Specia l Operations LCO s is optional, and therefore, no criteria of 10 CFR 50.36(c)(2)(i i ) apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is opt i onal . Operation at reactor coo l ant temperatures> 200 °F, can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures

> 200 °F, performance of inservice leak and hydrostatic testing would also necessitate the inoperability of some subsystems normally required to be OPERABLE when> 200 ° F.

Additionally, even with required minimum reactor coolant temperatures~ 200 ° F, RCS temperatures may drift above 200°F during the performance of inservice leak and hydrostatic testing or during subsequent control rod scram time testing, which is typically performed in conjunction with inservice leak and hydrostatic testing. While this Special Operations LCO is prov id ed for inservice le ak and hydrostatic testing, and for scram time testing initiated in conjunction with an inservice leak or hydrostatic test, parallel performance of other tests and inspections is not precluded.

(continued)

LaSalle 1 and 2 B 3.10.8-3 Revision

v t e Letter Sequence Request
EPID:L-2017-LLA-0415, TSTF-542 (Approved, Closed)
Initiation Request Acceptance... Supplement Administration Questions Answers Results Approval Other: ML19016A171
MONTHYEAR2018-10-15 [Table View]

RS-17-127, Submittal of Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.

Text

Subject:

Subject:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

Navigation menu

RS-17-127, Submittal of Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.

Text

Subject:

Subject:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

Navigation menu

Search