Text

Application to Revise Technical Specifications to Adopt TSTF-542. Reactor Pressure Vessel Water Inventory Control.
	ML17347A788
Person / Time
Site:	Perry
Issue date:	12/06/2017
From:	Hamilton D; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-17-045
	Download: ML17347A788 (277)
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{{#Wiki_filter:FENOC IU Perry Nuclear Power Plant P.O. Box 97 l0 Genter Road Frctftrergy Nuclear @rating funpany Parry, Ohio44OBl David B.llilrtlffi, ffi-280-5382 Wce Presidant December 6, 2017 L-17-045 10 cFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Application to Revlge Technical SpecificatioFs to Adopt TSTF-542. "Reactor Pressure Vessel Water lnventorv Control' Pursuant to t0 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) is submifting a request for an amendment to the Technical Specifications for the Perry Nuclear Power Plant. The proposed change replaces existing Technical Specifications (TS) requirements related to 'operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Gontrol (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. Attachment 1 provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachrnent 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only. Approval of the proposed amendment is requested by December 19,2018. Once approved, the amendrnent shall be implemented within g0 days. ln accordancewith 10 CFR 50.91, a copyof this application, with attachments, is being provided to the designated State of Ohio fficial. There are no regulatory commitments contained in this submittal. !f there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.

Perry Nuclear Power Plant L-17-045 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on December 6 ,2017. Sincerely, David B. Hamilton Attachments:

1. Description and Assessment

2. Proposed Technica! Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

4. Proposed Technical Specification Bases Changes (Mark-Up) cc: NRC Region lll Administrator NRC Resident lnspector NRC Project Manager Branch Chiel Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board

Attachment 1 L-17-045 Description and Assessment Page 1 of I

1.0 DESCRIPTION

The proposed change replaces existing Technical Specifications (TS) requirements related to "operations which have the potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water lnventory 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 FirstEnergy Nuclear Operating Company (FENOC) has reviewed the safety evaluation provided to the Technical Specifications Task Force on December 20,2016, as well as the information provided in TSTF-542. FENOC has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the NRC staff are applicable to the Perry Nuclear Power Plant (PNPP) and justify this amendment for the incorporation of the changes to the PNPP TS. The following PNPP 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 lsolation Cooling (RCIC) System Instrumentation 3.3.6.1, Primary Containment and Drywell lsolation lnstrumentation 3.3.7.1, Control Room Emergency Recirculation (CRER) System lnstrumentation 3.5.2, ECCS - Shutdown 3.6.1.2, Primary Containment Air Locks 3.6.1,3, Primary Containment lsolation Valves (PClvs) 3.6.1.10, Primary Containment - Shutdown 3.6.1.1 1, Containment Vacuum Breakers 3.6.1 .1 2, Containment Humidity Control 3.6.4.1, Secondary Containment 3.6.4.2, Secondary Containment lsolation Valves (SClvs) 3.6.4.3, Annulus Exhaust Gas Treatment (AEGT) System 3.7.3, Control Room Emergency Recirculation (CRER) System 3.7.4, Control Room Heating, Ventilating, and Air Conditioning (HVAC) System 3.8.2, AC Sources - Shutdown 3.8.5, DC Sources - Shutdown 3.8.8, Distribution Systems - Shutdown L-17-045 Description and Assessment Page 2 of I 2.2 Variations FENOC is proposing the following variations from the TS changes described in the TSTF-542 or the applicable parts of the NRC staffs safety evaluation. These variations do not affect the applicability of TSTF-542 or the staffs safety evaluation to the proposed license amendment. 2.2.1 The PNPP TS utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-542 was based. Specifically, the titles and numbers for the following PNPP TS differ from the STS discussed in TSTF-542: r PNPP TS 3.3.6.1, "Primary Containment and Drywell lsolation Instrumentation," corresponds to STS 3.3.6.1, "Primary Containment lsolation lnstrumentation,"

      . PNPP TS 3.3.7.1, "Control Room Emergency Recirculation (CRER) System lnstrumentation," corresponds to STS 3.3.7.1, "Control Room Fresh Air (CRFA)

System I nstrumentatio n," o PNPP TS 3.6.1.10, "Primary Containment - Shutdown," does not have a corresponding TS in the STS,

      . PNPP TS 3.6.1.11, Containment Vacuum Breakers," does not have a corresponding TS in the STS,
      . PNPP TS 3.6.1.12, "Containment Humidity Control," does not have a corresponding TS in the STS, I    PNPP TS 3.6.4.3, "Annulus Exhaust Gas Treatment (AEGT) System,"

corresponds to STS 3.6.4.3, "Standby Gas Treatment (SGT) System,"

      . PNPP TS 3.7.3, "Control Room Emergency Recirculation (CRER) System,"

corresponds to STS 3.7.3, "Control Room Fresh Air (CRFA) System,"

      . PNPP TS 3.7.4, "Control Room Heating, Ventilating, and Air Conditioning (HVAC)

System," corresponds to STS 3.7.4, "Control Room Air Conditioning (AC) System,"

      . PNPP TS do not include the inverter requirements found in STS 3.8.8, "lnverters - Shutdown," which is included in TSTF-542, r    PNPP TS 3.8.8, "Distribution Systems - Shutdown," corresponds to STS 3.8.10, "Distribution Systems - Shutdowl'r."

These differences are administrative and do not affect the applicability of TSTF-542 to the PNPP TS. Additionally, TS title updates have been made to align with those in the TSTF. These title updates are administrative and do not affect the applicability of TSTF-5 42 to the PN PP TS. 2.2.2 The PNPP TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement (SR) Frequencies for T$ 3.3.5.2 and TS 3.5.2 are "ln accordance with the Surveillance Frequency Control Program." Specifically, the initial frequencies for the new SRs in the licensee-controlled Surveillance Frequency Control Program will align with the STS frequencies discussed in TSTF-542, except as noted below: L-17-045 Description and Assessment Page 3 of I t sR 3.3.5.2.3 frequency will be "24 months" [PNPP has a 24-month operating cycle] a sR 3.5.2.7 frequency will be "24 months" [PNPP has a 24-month operating cycleJ a sR 3.5.2.8 frequency will be "24 months" IPNPP has a 24-month operating cycle] 2.2.3 The PNPP TS contains the following instrumentation information that is a variation from the STS on which TSTF-542 was based.

     . PNPP TS Table 3.3.5.1-1 , Functions 1.d, 1.e, and 2.d are identified as low pressure coolant injection (LPCI) and Iow pressure core spray (LPCS) subsystem injection valve permissives on reactor vessel pressure Iow versus STS identified LPCI and LPCS functions for reactor steam dome pressure low. The function of these PNPP instruments is identical to those described in the STS and differ in name only. The PNPP nomenclature will be retained in the proposed TS Table 3.3.5.2-1 for Functions 1.a, 1.c, and 2.a.
     . PNPP TS Table 3.3.5.1-1 , Functions 1.d and 1.e address reactor vessel pressure low (injection valve permissives) for LPCS and LPCI, respectfully, versus STS Function 1.d that addresses both LPCS and LPCI injection (pressure) permissives as one function. The PNPP format with two separate functions will be retained in the proposed TS Table 3.3.5.2-1 for Functions 1.a and 1.c.
     . PNPP TS Table 3.3.5.1-1 , Function 2.d addresses reactorvessel pressure low (LPCI injection valve permissive), which is applicable in Modes 1 through 5, versus STS Function 2.d that, as written, is only applicable in lVlodes 1 through 3.

With the proposed changes, PNPP TS Table 3.3.5.1-1 , Function 2.d, will only be applicable in Modes 1 through 3. 2.2.4 PNPP TS Table 3.3.5.1-1 , Functions 1.a and 2.a currently reference Note (f), which is associated with PNPP TS 3.6.4.3, "Annulus Exhaust Gas Treatment (AEGT) System." Due to PNPP's design as a primary containment plant, the AEGT system is not required to be operable during Modes 4 and 5 and will no longer be applicable during OPDRVS. As such, Note (f) will be deleted consistentwith other TSTF-542 OPDRv-related changes. 2.2.5 TS 3.3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control lnstrumentation," Table 3.3.5.2-1 , is revised to reflect the PNPP design. 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 TSTF-S42 are not included in the proposed Technical Specifications. This variation corrects an error in TSTF-542 that affects the BWR/5 and BWR/6 ECGS instrumentation req uirements. The purpose of the manual initiation function is to allow manual actuation of the ECCS subsystem required by TS 3.5.2 to mitigate a draining event. The "Reactor Vessel Water Level - High, Level S" signal prevents overfilling of the reactor vessel into the main steam lines by closing the HPCS injection valves when the water level L-17-045 Description and Assessment Page 4 of I is above the Level I setpoint. Therefore, if HPCS is the required ECCS subsystem and the water leve! is above Level 8, manually actuating Function 3.e will not inject inventory into the reactor vessel. This is not the desired response. lf the Level I function is retained in Table 3.3.5.2-1 , the function would need to be rendered inoperable in order to inject water when above the Level 8 water level. This would not be consistent with including the function in Table 3.3.5.2-1 . The PNPP has the capability to manually start the HPCS pump and to open the HPCS injection valve if needed, not utilizing Functions 3.a and 3.e. lf desired to inject water into the reactor pressure vessel using the HPCS, the reactor operator can follow procedural steps to take manual control of the pump and injection valve to add inventory. !f the water level is above Level 8, then manual override of the Level I function can be performed to allow the HPCS injection valve to be opened. These actions can be performed from the control room and can be accomplished well within the 1-hour minimum drain time limit specified in TS 3.5.2, Condition E. Consequently, the Function 3.a and 3.e instrumentation functions are not needed to actuate the HPCS subsystem components to mitigate a draining event. The ability to override the HPGS Level I isolation is already part of the PNPP emergency operating procedures and is practiced during operator training. SR 3.5.2.8 is revised to assure that the HPCS manual start capability (including the HPCS Level I isolation override feature) is tested. As part of this correction, TS 3.3.5.2, Condition E has also been deleted. Additionally, the associated HPCS functions have been relabeled to account for the deletions of Functions 3.a and 3.e. 2.2.6 PNPP TS 3.3.6.1, "Primary Containment and Drywell Isolation lnstrumentation," Required Action (RA) J.2, which states to "lnitiate action to isolate the Residual Heat Removal (RHR) Shutdown Cooling System suction from the reactor vessel," will be deleted. The direction to initiate action to close the RHR shutdown cooling (SDC) isolation valves in Mode 3 is in direct conflict with TS 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown," which requires two RHR SDC subsystems to be operable, and if not, to take immediate action to restore an RHR SDC subsystem to operable status (RA A.1). Therefore, RA J.2 should be deleted. Removing RA J.2 is also appropriate to protect plant safety. As discussed in the Bases to Function 5.b, the Reactor Vessel Water Level - Low, Level 3 Function associated with the RHR SDC subsystem is not directly assumed in the safety analyses because a break of the RHR SDC subsystem is bounded by breaks of the reactor recirculation system and main steam lines. Specifically, for the RHR SDC isolation valves to be open in ft/ode 3, reactor steam dome pressure would need to be below the RHR cut-in permissive pressure. Should a loss of coolant accident (LOCA) occur inside primary containment, TS 3.5.1, "ECCS - Operating," explicitly credits the manual closing of the RHR SDC isolation valves and alignment of RHR in the LPCI mode. Similarly, if the break is on the RHR SDC subsystem outside primary containment, credit can still be given for manual closing of the RHR SDC isolation valves and alignment of an intact LPCI loop. ln either case, core uncovery would not result and radiological consequences are bounded by the LOCA and main steam L-17-045 Description and Assessment Page 5 of 9 line break accidents. For these reasons, it is not critical to immediately initiate action to close the RHR SDC isolation valves (RA J.2) if Function 5.b is inoperable. 2.2.7 PNPP TS 3.3.6.1, "Primary Containment and Drywell lsolation lnstrumentation," Required Action (RA) J.3.1 through J.3.3 will be deleted. RA J.3.1, "lnitiate action to restore primary containment to OPERABLE status;" RA J .3.2, "lnitiate action to restore isolation capability in each required primary containment penetration flow path not isolated;" and RA J.3.3, "lnitiate action to close one door in each primary containment air lock," are no longer required. Currently, primary containment is required to be operable in Modes 1, 2, and 3 and during OPDRVS. For primary containment to be operable, the primary containment penetration flow paths are required to be isolated, and at least one primary containment airlock door is required closed. With the deletion of OPDRVs and the Mode 4 and 5 requirements from Table 3.3.6.1-1 , Function 5.b, this required action no longer applies. Related requirements are included in the proposed TS 3.5.2. 2.2.8 The PNPP TS do not contain a Note on LCO 3.5.2 regarding realignment to the LPCI mode. The Note that is provided on LCO 3.5.2 in STS is limited to SR 3.5.2.4 in the PNPP TS. The proposed PNPP LCO 3.5.2 will include this Note, which is relocated from SR 3.5.2.4 to align with the STS. This is a minor variation, as the purpose of the Note is the same as the one described in the STS and the Note is applicable to the PNPP. 2.2.9 Optional Required Actions C.3 and D.4 from the TSTF-542 proposed changes to TS 3.5.2, will not be included. By design, the PNPP is a primary containment plant. As such, the action to "verifit one standby gas treatment subsystem is capable of being placed in operation," which would be required to support secondary containment operability, is not required. 2.2.10 PNPP TS 3.6.1.2, "Primary Containment Air Locks," currently includes in its Applicability, "During operations with a potential for draining the reactor vessel (OPDRVS)," which will be deleted consistent with other TSTF-542 changes. This PNPP TS also includes Required Action E.2,'*lnitiate action to suspend OPDRVs," which will also be deleted consistent with other TSTF-542 OPDRV-related changes. Related requirements will be included in the proposed TS 3.5.2. 2-2.11 The Applicability to TS 3.6.1.3, "Primary Containment lsolation Valves (PClVs)," currently states: MODES 1 , 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment and Drywell lsolation lnstrumentation." Though TSTF-542 relocated most of the non-Mode 1,2, or 3 instrumentation requirements to LCO 3.3.5.2 [Note (d), "During movement of recently irradiated fuel assemblies in primary containment," remains in Table 3.3.6.1-1], the Applicability statement remains applicable for Condition F when PCIVs are required to be L-17-045 Description and Assessment Page 6 of I OPERABLE during movement of recently irradiated fuel assemblies in primary containment. Condition G, which only applies to conditions in Mode 4 and 5 or during OPDRVS, is deleted in its entirety. This is considered an administrative variation and is consistent with other TSTF-542 OPDRV-related changes. 2.2.12 STS and TSTF-542 do not have a corresponding TS 3.6.1.10, "Primary Containment - Shutdown." PNPP TS 3.6.1.10 currently includes in its Applicability, "During operations with a potential for draining the reactor vessel (OPDRVs)," which will be deleted consistent with other TSTF-542 changes. This PNPP TS also includes Required Action A.2, "lnitiate action to suspend OPDRVs," which will also be deleted consistent with other TSTF-542 OPDRV-related changes. By design, the PNPP is a primary containment plant. During shutdown, the primary containment performs a similar function to the secondary containment in other boiling water reactor designs. 2.2.13 STS and TSTF-542 do not have a corresponding TS 3.6.1.11, "Containment Vacuum Breakers." PNPP TS 3.6.1.11 currently includes in its Applicability, "During operations with a potential for draining the reactor vessel (OPDRVs)," which will be deleted consistent with other TSTF-542 changes. This PNPP TS also includes Required Action B.2.2, "lnitiate action to suspend OPDRVs," which will also be deleted consistent with other TSTF-542 OPDRV-related changes. 2.2.14 STS and TSTF-542 do not have a corresponding TS 3.6.1 .12, "Containment Humidity Control.' PNPP TS 3.6.1.12 currently includes in its Applicability, "During operations with a potential for draining the reactor vessel (OPDRVS)," which will be deleted consistent with other TSTF-542 changes. This PNPP TS also includes Required Action C.2, "lnitiate action to suspend OPDRVs," which will also be deleted consistent with other TSTF-542 OPDRV-related changes.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis FENOC requests adoption of TSTF-542, "Reactor Pressure Vessel Water lnventory Control," which is an approved change to the Standard Technical Specifications (STS), into the Perry Nuclear Power Plant (PNPP) 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 lnventory 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. FirstEnergy Nuclear Operating Company (FENOC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "lssuance of amendment," as discussed below: L-17-045 Description and Assessment Page 7 of I 1 . Does the proposed amendment involve a significant increase in the probability or consequences sf 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 ttflode 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 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 tVlodes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or othenruise, 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 capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed. The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. L-17-045 Description and Assessment Page 8 of I

2. Does the proposed amendment create the possibility of a new or different kind of accident from any 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. The proposed change will not alter the design function of the equipment involved. Underthe 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 ditferent than if those systems were unable to perform their function under the current TS req u irements. 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 license bases. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident 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 WC. The current requirements do not have a stated safety basis and no margin of safety is established in the license 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 RPVwater level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. Based on the above, FENOC 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. L-17-045 Description and Assessment Page I of I

4.0 ENVIRONMENTAL CONSIDERATION

The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 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(cXg). Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Attachment 2 L-17-045 pRoposED TEcHNtcAL SPECIFICATTON GHANGES (MARK-UP) (68 pages follow)

TECHNICAL SPECI FICATION MARK.UP TABLE OF CONTENTS 0 USE AND APPLICATION 1 Definitions..... 1 .0-1 2 Logical Connectors.. 1 .0-B 3 Completion Times 1 .0-1 1 4 Frequency... 1.0-24 2.0 SAFETY LllvllTS (SLs) 2.1 SLs. 2.0-1 2.2 SL Violations. 2.0-1 3.0 LrtMrTrNG CONDtTtON FOR OPERATTON (LCO) AppLlCABtLrTy 3.0-1 3.0 SURVEILLANCE REQUIREIUENT (SR) APPLICABILITY. 3.0-4 3.1 R EACTIVITY CONTRO L SYSTE tUS 3.1 .1 SHUTDOWN MARGIN (SD]\/)... 3.1 -1 3.1 .2 Reactivity Anomalies. . 3.1 -5 3.1.3 Control Rod OPERABILITY.... 3.1 -7 3.1 .4 Control Rod Scrarn Times 3.1 -12 3.1 .5 Control Rod Scram Accumulators. 3.1 -15 3.1 .6 Control Rod Pattern 3.1 -18 3.1.7 Standby Liquid Control (SLC) System 3.1 -20 3.1 .B Scram Discharge Volume (SDV) Vent and Drain Valves.. 3.1 -24 3.2 POWER DISTHIBUTION LIIilITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATTON RATE (APLHGH)........3.2-1 3.2.2 MlNttvlutu cHlTrcAL powER RATIO (MCPR) .....3.2-2 3.2.3 LINEAR HEAT GENERATTON RATE (LHGH) ......3.2-3 3.3 INSTRUMENTATION 3.3. 1 .1 Reactor Protection System (RPS) lnstrumentation 3.3- 1 3.3.1 .2 Source Range Monitor (SH[fl) lnstrumentation 3.3-10 3.3.2.1 Control Flod Block lnstrumentation ..... 3.3-15 3.3.3.1 Post Accident Monitoring (PANI) Instrumentation . 3.3-20 3.3.3.2 Remote Shutdown System . .. 3.3-24 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) lnstrumentation..... .3.3-26 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) lnstrumentation 3.3-29 3.3.5.1 Emergency Core Cooling System (ECCS) lnstrumentation. . 3.3-32 3.3.5.2 neactor Pressur Instrumentation. 3*3*L3 Reactor Core lsolation Cooling (RCIC) System lnstrumentation. . 3.3-44 3.3.6.1 Primary Containment lsolation lnstrumentation ... 3.3-48 3.3.6.2 Residual Heat Removal (RHR) Containment Spray System lnstrumentation. 3.3-60 (continued) PEFIRY _ UNIT 1 Amendment No. 69

TECHNICAL SPECIFICATION MARK-UP TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6.3 Suppression Pool lvlakeup (SPMU) System lnstrumentation 3.3-64 3.3.6.4 Relief and Low-Low Set (LLS) lnstrumentation 3.3-68 3.3.7.1 Control Room Emergency Recirculation (CHER) System Instrumentation. ...... 3.3-70 3.3.8.1 Loss of Power (LOP) lnstrumentation 3.3-74 3.3.B.2 Reactor Protection System (RPS) Electric Power l\Ionitoring..... .3.3-77 3.4 REACTOR COOLANT SYSTEIVI (RCS) 3.4.1 Recirculation Loops Operating 3.4-1 3.4.2 Flow Control Valves (FCVs) 3.4-6 3.4.3 Jet Pumps. . . 3.4-8 3.4.4 Safetyi Relief Valves (S/RVs).... 3.4-1 0 3.4.5 RCS Operational LEAKAGE 3.4-12 3.4.6 RCS Pressure Isolation Valve (PlV) Leakage 3.4-14 3.4.7 RCS Leakage Detection lnstrumentation. 3.4-16 3.4.8 RCS Specific Activity. 3.4-19 3.4.9 Residual Heat Removal (HHR) Shutdown Cooling System - Hot Shutdown 3.4-21 3.4.1 0 Hesidual Heat Flemoval (RHR) Shutdown Cooling System - Cold Shutdown..... 3.4-24 3.4.11 RCS Pressure and Temperature (Pfl-) Limits. 3.4-26 3.4.12 Heactor Steam Dome Pressure 3.4-32 3.5 EMERGENCY CORE COOLTNG SYSTEh/S (ECCS). RpV WATER TNVENTOW CNIEOL AND REACTOR CORE TSOLATTON COOLTNG (RClC) SYSTEM 3.5.1 ECCS - Operating. .3.5-1 3.5.2 Reactor Pressure Vessel (RPV) Water lnventorv Controlffi 3.5-6 3.5.3 RCIC System 3.5-10 3.6 CONTAINMENT SYSTEIVIS 3.6 1 .1 Primary Containment - Operating..... 3.6- 1 3.6 1 .2 Primary Containment Air Locks 3.6-3 3.6 't .3 Primary Containment lsolation Valves (PCIVs).... 3.6-9 3.6 1 .4 Primary Containment Pressure 3.6-20 3.6 1 tr Primary Containment Air Temperature. 3.6-21 3.6 'l .6 Low-Low Set (LLS) Valves 3.6-22 3.6 .7

    '1 Flesidual Heat Hemoval (HHR) Containment Spray System                       3.6-24 3.6 1 .8      Feedwater Leakage Control System (FWLCS). ...                               3.6-26 J.O 1 I       [\Iain Steam lsolation Valve (lMSlV) Leakage Control System (LCS)           3.6-27 3.6 1 10      Primary Containment - Shutdown........                                      3.6-29 3.6.1.1't     Containment Vacuum Breakers.                                            .. 3.6-31 3.6.'l .12    ContainmentHumidityControl                                           ....3.6-34 3.6.2.1       Suppression Pool Average Temperature.                                 .... 3.6-36 (continued)

PERRY _ UNIT 1 tl Amendment No. 6,9

TEGH NIGAL SPECIFICATION MARK-UP ["NEW" PAGEI elmiilons Lt t-t Oetinitlons (c DBAIN-TIME rne nnnlru rtn4f is th inventorv in anO an to drain to tne too o BPV-,assuming al fne water invent limitin0 drain rarc; b) fhe limiting dra through a single oe tlow rate, or tne sum multiole penetrat common mode taitur normal power, sing tlow patns nenw tn t . Penetration tlo closed svstem, or ic vatves tnat are loc secureO in tne Oos other devices that tnrouptr tne pene Z. Penetration ttow vatves tnat witt cl power prior to tne n tne fnf wnen actuat isolation instrumentation : or S. Penetration flo can be ctoseO prio ing equal to the TAF bv a the task, who is in c the co.[trol room, i capable of closin0 the oenetration flow oath isolation devices c) The penetration flow paths reouired to be evaluated per paragraph b) ar instantaneoustv a and no water is assu tne nPV water lnve , PERRY - UNIT 1 1 .0-2a Amendment No

TECH NIGAL SPECIFICATION MARK-UP ["NEW" PAGE] Definilions Lt L]-Delurlbns DMIUIIME Ot No additional dr rcsnlinued) e) Realistic cross uEed= A boundins DHAIN TI calculated value. {continued) PEFIRY _ UNIT 1 1.0-2b Amendment No.

ECCS Instrumentation TECHNICAL SPECIFICATION 3.3. t.l PROVIDED FOR CONTEXT 3.3 II{STRUI{EHTAT IOH 3.3 .5.1 Emergency Core Cooling System (ECCSI Instrumentation tco 3.3.5.1 The ICCS lnstrumentetion for erch Function in Tabl e 3.3.5. I-I shal I be 0PERABLI. APPLICA$ILITY: According ts Trble 3.3.5.1-1, NITIOHS Separate f,ondition entry is allowed for each channel. COH[}ITIOH REQUIREo ACTIoH TOHPLETIOf{ TII-IE A One or mCIr'E channel s A.t Enter the Condition Immedi ately i noperabl e . referenced in TabIe 3.3.5.1*l for the channel. (conti nued) PERRY UHIT I 3.3-32 Amendment I'lo " 69

TECHNICAL S PECIFICATION MARK.UP ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. As required by Required 8.1 Action A,1 and @in referenced in Table 3.3.5.1-1. Z. Only applicable for Functions 1.a, 1.b, 2.a and 2.b. Declare supported t hour from discovery feature(s) inoperable when of loss of initiation its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions AND 8.2

                                  @in
                           ------L     Only applicable for Functions 3.a and 3.b.

Declare High Pressure t hour from discovery Core Spray (HPCS) of loss of HPCS System inoperable. initiation capability AND 8.3 Place channel in trip. 24 hours (continued) PERRY - UNIT 1 3.3-33 Amendment No. 147

TECH NICAL SPECIFICATION MARK.UP ECCS lnstrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TI]VIE C. As required by Required Action A.1 and referenced in Table 3.3.5. 1- 1 .

1. Only applicable for Functions 1.c, 1.d, 1.e, 2.c, and 2.d.

Declare supported t hour from discovery feature(s) inoperable when of loss of initiation its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions AND c.2 Restore channel to 24 hours OPERABLE status. (continued) PERRY - UNIT 1 3.3-34 Amendment No. 147

TECHNICAL SPECIFICATION MARK.UP ECCS lnstrumentation 3.3.5.1 ACTIONS continu CONDITION HEQUIRED ACTION COMPLETION TI]\4E E. As required by Flequired E.1 --NOTES: Action A.1 and @in referenced in Table 3.3,5.1-1 . 1 Only applicable for Functions 1.f, 1.9, and Z.e. t hour from discovery Declare supported of loss of initiation feature(s) inoperable when capability for its redundant feature feature(s) in both ECCS initiation capability divisions is inoperable. AND 7 days E.2 Restore channel to OPEHABLE status F. As required by Required F.1 Declare Automatic t hour from discovery Action A.1 and Depressu rization System of loss of ADS referenced in (ADS) valves inoperable. initiation capability in Table 3.3.5.1-1. both trip systems AND F.2 Place channel in trip. 96 hours from discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable AND I days (continued) PERRY _ UNIT 1 3.3-36 AmendmentNo. 111

TEC HNICAL SPECI FICATION MARK.UP ECCS lnstrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES 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 (LPCI) and Low Pressure Core Spray (LPCS) Subsystems

a. Reactor VesselWater 1,2, + 2@J B sR 3.3.5.1.1 > 14.3 inches Level - Low Low Low, 4+elffi_ sR 3.3.5.1.2 Level 1 5taHs sR 3.3.5.1.3 sR 3.3.s.1.s sR 3.3.5.1.6

b. Drywell Pressure - High 1,2, 3 2(b) B sR 3.3.5.1.1 < 1.88 psig sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

c. LPCI Pump A 1,2,+ C sR 3.3.s.1.2 < 5.25 Start - Time Delay sR 3.3.5.1.4 seconds Relay 4Fi$ft) sR 3.3.s.1.6

d. Reactor Vessel 1,2,3 C sR 3.3.s.1.1 > 482.7 psig Pressure - Low (LPCS sR 3.3.5.1.2 and lnjection Valve sR 3.3.5.1.3 < 607.7 psig Permissive) sR 3.3.5.1.5 sR 3.3.5.1.6 4t')#{+) + B sR 3,3,5,1,1 >-q#-Z+s,tg sR-3'# end sE+3s+3 <+g#+sie

                                                                                       #-ffi sR#

e. Reactor Vessel 1,2.3 c sR 3.3.5.1.1 > 490.0 psig Pressure - Low (LPCI sR 3.3.s.1.2 and lnjection Valve sR 3.3.5.1.3 s 537.1 psig Permissive) sR 3.3.5.1.5 sR 3.3.5.1.6 4$r+4 + B sR 3,3,5,1 ,1 >+e0+sig sE# effd ffi <+S#-psrg sR#rs SR-#

f. LPCS Pump Discharge 1,2, 3; E sR 3.3.5.1.1 > 1200 gpm Flow - Low (Bypass) 4e)r5+4 sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6 (continued)

(b) Also required to initiate the associated diesel generator and AEGT subsystem. (f) When asseeiated AEGT subsystems are required te be OPERABTE per te0 3,6,4,3, Annulus trxhust Gas PERRY - UNIT 1 3.3-39 Amendment No. 164

TEC H NIGAL SPECIFICATION MARK-UP ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REOUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION 4.1 REQUIREMENTS VALUE

1. Low Pressure Coolant lnjection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems (continued)

g. LPCI Pump A Discharge 1,2, + E sR 3.3.s.1.1 > 1450 gpm Flow - Low (Bypass) 4)#$ sR 3.3.s.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

h. Manual lnitiation 1,2,+ C sR 3.3.5.1.6 NA 4(e)St")

2, LPCI B and LPCI C Subsystems

a. Reactor VesselWater 1,2,+ 2(b) B sR 3.3.5.1.1 > 14.3 inches Level - Low Low Low, ,r{al(& sR 3.3.5.1.2 Level 1 cl{S sR 3.3.5.1.3 E

sR 3.3.5.1.5 sR 3.3.5.1.6

b. Drywell Pressure - High 1,2, 3 2(b) B sR 3.3.5.1.1 s 1.88 psig sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3,5.1.5 sR 3.3.5.1.6

c. LPCI Pump B Start - Time 1, 2, 3.i C sR 3.3,5.1.2 s 5.25 Delay Relay 4(4-5+4 sR 3.3.5.1.4 seconds sR 3.3.5.1.6

d. Reactor Vessel 1,2,3 1 per C sR 3.3.5.1.1 > 490.0 psig Pressure - Low (LPCI subsystem sR 3.3.5.1.2 and Injection Valve sR 3.3.5.1.3 < 537.1 psig Permissive) sR 3.3.5.1.s for LPCI B; sR 3.3.5.1.6 and

                                                                                                        > 490.0 psig and
                                                                                                        < 537.1 psig for LPCI C
                                         +),S@             +p+               g     sR  9,3.5,1=1      ><e0$+sg sugsystem                   sE#5+2             end sE-=e5=13         <+e+l+slg w

SR--3.3S=H fer tPe I B;

                                                                                                        +HC
                                                                                                        >+e0+Bs+g end
                                                                                                        <+g+=+psrg fer tPCl e (continued)

(a) When asseeiated ECCS subsystem(s) are reqsired te be OPER ETE Ber tCO 3,5,2, ECCS Shutdewn= (b) Also required to initiate the associated diesel generator and AEGT subsystem. (0 When asseeiated AEGT subsysteme are required te be OPERABTE Ber te O 3,6,4,3, Annulus Exhaust Gas PERRY - UNIT 1 3.3-40 Amendment No.164

TECH N ICAL SPECIFICATION MARK.UP ECCS lnstrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI B and LPCI C Subsystems (continued)

e. LPCI Pump B and LPCI 1,2. 3, 1 per pump E sR 3.3.5.1.1 > 1450 gpm Pump C Discharge +t4+(a) sR 3.3.5.'1.2 FIow - Low (Bypass) sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

f. Manual lnitiation 1,2, + c sR 3.3.5.1.6 NA

                                       +t")S@

3. High Pressure Core Spray (HPCS) System

a. Reactor VesselWater 1,2, + 4(e) B sR 3.3.5.1.'t > 127.6 inches Level - Low Low, +(a)-rs@ sR 3.3.5.1.2 Level 2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

b. Drywell Pressure - High 1,2,3 4(e) B sR 3.3.5.1.1 < 1.88 psig sR 3.3.5.1.2 sR 3.3.5.1,3 sR 3.3.5.1.5 sR 3.3.5.1.6

c. Reactor VesselWater 1,2, 4 4 B sR 3.3.5.1.1 <221.7 inches Level- High, Level I a{4,4(4 sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.s.1.6

d. Condensate Storage Tank 1,2, + 2 D sR 3.3.s.1.1 > 90,300 Level - Low +t")r+S sR 3.3.5.1.2 gallons sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

e. Suppression PoolWater 1, 2,3 2 D sR 3.3.s.1.1 <18ft6inches Level - High sR 3.3"s,1.2 sR 3.3.5.1.3 sR 3.3.5.1.7 sR 3.3.5.1.6 (continued)

(a) When asseeiated treCS subsystem(s) are required te be OPERABTE per t6O 3,5,2, ECCS Shutdewn, eendensate sterage tank while tank water level is net within the Iimits ef SR 3,5,2,2, (e) Also required to initiate the associated diesel generator. PERRY _ UNIT 1 3,3-41 Amendment No. 164

TEC H NICAL SPECIFICATION MARK.UP ECCS lnstrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION 4.1 REQUIREMENTS VALUE

3. High Pressure Core Spray (HPCS) System (continued)

f. HPCS Pump Discharge 1,2,3t 1 E sR 3.3.5.1.1 > 120 psig Pressure - High (Bypass) +(4+@ sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

g. HPCS System Flow 1,2, + 1 E sR 3.3.5.1.1 > 600 gpm Rate - Low (Bypass) +{4+S sR 3.3.5.1.2 sR 3.3.s.1.3 sR 3.3.s.1.5 sR 3.3.5.1.6

h. Manual lnitiation 1,2,+ 1 C sR 3.3.5.1.6 NA

                                   +)rs@

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water 1, 2(d), 3(d) 2 F sR 3.3.5.1.1 >'14.3 inches Level - Low Low Low, sR 3.3.5.1.2 Level 1 sR 3.3.5,1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

b. ADS Initiation Timer 1, 2(d), 3(d) 1 l.J sR 3.3.5.1.2 u 100.5 seconds sR 3.3.5.1.4 and sR 3.3.5,1.6 < 109.5 seconds

c. Reactor VesselWater 1, 2(d), 3(d) 1 F sR 3.3.5.1.1 >- 177.1 inches Level- Low, Level 3 sR 3.3.5.1.2 (Confirmatory) sR 3.3.s,1.3 sR 3.3.5,1.5 sR 3.3.s.1.6

d. LPCS Pump Discharge 1, 2(d), 3(d) 2 L, sR 3.3.5.1.1 > 125 psig Pressure - High sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

e. LPCI Pump A Discharge 1, 2(d), 3(d) 2 tJ sR 3.3.5.1.1 > 115 psig Pressure - High sR 3.3.s.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

f. Manual lnitiation 1, 2(d), 3(d) 2 tJ sR 3.3.5.1.6 NA (continued)

(a) When asseeiated ECCS subsystem(s) are required te be OPERABTE per te0 3,5,2, ECCS Shutdewn, (d) With reactor steam dome pressure > 150 psig. PERRY - UNIT 1 3.3-42 Amendment No. 16,1

TECHNIGAL SPECIFICATION RPV Water lnventorv MARK-UP ["NEW" PAGEI 3.3.5.2 3.3 lNSTRUIVENTATIffi 3.3.5.2 neactor Pr 1CO s.3.5.2 The HPV faUe S.3.5.2-1 sh APPI lCARll ITY' Aennrdinn tn Tahla i 3 5 2-1 ACTIONS Separate ConOitio CONDITION REAUIRED ACTION COIVIPLETION TIIVIE A. One or more chann# A.1 Enter the Condi Irunediately inooerable. referencedin Table S.3.5.2-1 fo channel. B. As reouired bv Reouired 9.1 Declare assocl lmmediatelv Aelion A.lsel penetration flow p referenced in lncapabte of automffi Table 3.S.5.2{ isolail-m= AND e.2 Calculate DRAIlrLTIl\/E. lmmediatelv C. ns requireO nv ne C.t Ptace cnannel ip. t hour AclionAJ-anel rclertrced-ul Table 3.3.5.2-t . (continrred) PERRY - UNIT 1 3.3-43a Amendment No.

TECH NICAL SPECIFICATION HPV Water lnventor MARK-UP ["NEW" PAGEI 3.3-52 CffDII]ON REOUIRED ACTION COIVIPLETION TIIVE

n. ns requireO Uv neq D.t Declare HPCS sv l-h!.rJr AclionAt ancl inape.rahie-rcIercncedn Table 3.3.5.e-t. G n.Z nlign the HPCS p t hour suction to the sup ion PoaL E. As reouired bv Reouired E.t Restore channeh 24 hours Astton A-Land OPEHABLE status relerencedun Table 3.3.5.2-1 .

F. Reouired Action and F. t Declare assocl fxnqcdialelv associated Comple iniectionlsprav s fime of Condition C.D. lxoBerable-E, or F not met. PERRY - UNIT 1 3.3-43b Amendment No.

TECH NICAL SPEGIFICATION RPV Water lnventorv MARK-UP ["NEW" PAGEI 3t=52 SUHVEILLANCE HEOU OTE neter to faUle S.3. SUFIVEILLANCE FREOUENCY SR S.3.5.2.1 Perfo lrccco4gnce yilttrthe Suryei.llance Freouencv Control Prooram SR 3.3.5.e.2 Perfo lnaccffdance lurlh-Ihe

                                                                  $uveillanee Ercquency Control Prooram SR   3.3.5.2.3     Perform LOGIC SYSTEM FUNCTIONAL TEST.          lnaccofdaff.e wlh-the Sweillance Ercqlency Contro! Prooram PERRY _ UNIT     1                      3.3-43c                   Amendment No

TECH NICAL SPECIFICATION MARK-UP ["NEW" PAGE] RPV Water lnventor e3-52 faUb S.3.5.2-1 (p RPV Water lnvento ion APPLICAELE CONDTIIOMS MODES_OE EEOTJIBED EEEEBENCED A]HEB CHANNELS EBOM SEECIEIED EEB BEOIIIBED STJBVEILLANCE ALLOIVAELE FUNCTION CONDITIONS ACTIONA.l BEOUIREMENTS VALUE

1. Low Pressure Coo lniection-A ll PCI\ and I nw Pressrrre Core Snrav

{LPCSI Srrhsvstems a Fleactor Vessel Pressrrre - Low E l_(r) a sE_3*ffi2J sB__3-ti-z-e =_482J-psis and (t-PCS lniection Vam <_60zJ-.osio __---Pemisstve)

b. LPCS Pumo Discharoe Flow- low fBvnass) 4J te E sE_3-ffi-21 E-1200spm sE-?-?.522 c Reactor Vessel 4J I (r) c sE_3^?.5-2J f+gU0-psig Pressure - Low sR 3.3.5.2.2 and (t-PCl tniection Vam <j3il_.osis Permissive)

d. LPCI Pump A Discharoe Low (Bvnass'l 4t le E sB_3-3521 I-1450-opm Flow - sE_3-?^522

e. IVanual lnitiation dE

                                           -l.{J,         1_(")

E sB-?-?.52-? NA

2. LPCI Band LPCrc Subsvstems a Reaetor Vessel Pressrrre - Low (LPCI U l-oer subsystem c sB-?-?-tet sE_3.?S22 8490J-psis and tniection Valve @ affiZJ*o,tig Permissive) forLECtE

_____nxd 4490-0-psis and g 53zl-"osio IoTLPXJ-C

n. LPCl Pump A and LPO 4J ]-peruump E SE_.3-?S-2J al450spm Pump C Oiscnarge @ sR 3.3.5.2.2 Flow - Low (Bvoass)

c. lrilanual lnitiatiol 4J la E sB_3-35.2.s NA fcontinrredl (ai nssociateO w (RPVI Water lnventorv Control."

PERRY - UNIT 1 3.3-43d Amendment No.

TECHNICAL SPECIFICATION nPV Water lnventorv MARK-UP ["NEW" PAGEI 3-3:il ranP g.S.S.Z-t (p RPV Water lnvento ion AEELICAELE CANDII]AItE MODESIE EEOIJIEED BEEEBENCED AIHEB GTIANNELS EEOId SPECIEIED EEE BECIUIBED SIJBVEILLANCE ALLO]AIAELE FUNCTION CONDITIONS ACTION A.1 REGUIREMENTS VALUE S. Hioh Pressure Co {HPCS) Svstem

a. ConOensate Stor a(b) q(b) ,(a) D sE-L15.e1 Level - Low sE33-522 =r0J00 gailotrs h HPCS Prrmn Discharoe Prcssr rro Hioh {Bvnass) 4 1_t")

E sB_3-?-52-1 sE_3r5-22 a l20+srs

c. HPCS Svstem Flow nate - Low (gvpa 4J a E sE-3-?l2;t sFt 3.3.5.2.2 =-000-0.0m

+. nHH Svstem tsol*ion

a. Reactor VesselW @ zircne-tnp B sE_3l.5.?.J 2 177.1 inches Level-1ow level3 syslem sB_3.?5-22 S- neactor Water Clry (RWCU) Svstem tsomion a Rcantnr Vessel Water @ Zin-one-Idp B sE_3-ffi2-r Z-127$inr.he.r lcvpl-l Low Level 2 svslem sR 3.3.5.2.2 (at nssociatea wit (nPU Water tnvent (n) Wnen UPCS is OPe Controt," anA all (c) Wnen automatic PERRY _ UNIT 1 3.3-43e Amendment No

TECHNIGAL SPECI FICATION MARK.UP RCIC System lnstrumentation 3.3.5.42 3.3 INSTRUMENTATION 3.3.5.32 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.92 The RCIC System instrumentation for each Function in Table 3.3.5.3+1 shall be OPERABLE. APPLICABILITY: MODE 1, TVIODES 2 and 3 with reactor steam dome pressure > 150 psig. ACTIONS E Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels 4.1 Enter the Condition lmmediately inoperable. referenced in Table 3.3.5.9+1 for the channel. B. As required by Required 8,1 Declare RCIC System t hour from discovery Action A.1 and inoperable. of loss of RCIC referenced in initiation capability Table 3.3.5.3+1. AND 8.2 Place channel in trip. 24 hours C. As required by Required c.1 Restore channel to 24 hours Action A.1 and OPERABLE status. referenced in Table 3.3.5.3+1. (continued) PERRY - UNIT 1 3.3-44 Amendment No. 69

TEC HNICAL SPECIFIGATION MARK.UP RCIC System Instrumentation 3.3.5.32 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TlME D. As required by Required D.1 ...NOTE Action A.1 and Only applicable if RCIC referenced in pump suction is not aligned Table 3.3.5.3+1. to the suppression pool. Declare RCIC System t hour from discovery inoperable. of loss of RCIC initiation capability AND D.2.1 Place channel in trip. 24 hours OR D.2.2 Align RCIC pump suction 24 hours to the suppression pool. E. Required Action and E.1 Declare RCIC System lmmediately associated Completion inoperable. Time of Condition B, C, or D not met. PERRY _ UNIT 1 3.3-45 Amendment No. 6S

TECHNICAL SPECIFICATION MARK.UP RCIC System lnstrumentation 3.3.5.32 SURVEILLANCE REQUIREM ENTS 1 Refer to Table 3.3.5.9+1 to determine which SRs apply for each RCIC Function. 2 When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 5; and (b) for up to 6 hours for Functions 1, 2, 3, and 4 provided the associated Function maintains RCIC initiation capability. SURVEILLANCE FREQUENCY SR 3.3.5.32.1 Perform CHANNEL CHECK. ln accordance with the Surveillance Frequency Control P ram SR 3.3.5.32.2 Perform CHANNEL FUNCTIONAL TEST. ln accordance with the Surveillance Frequency Control ram SR 3.3.5.32.3 Calibrate the trip unit. ln accordance with the Surveillance Frequency Control SR 3.3.5.32.4 Perform CHANNEL CALIBRATION. ln accordance with the Surveillance Frequency Control ram SR 3.3.5.32.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. ln accordance with the Surveillance Frequency Control ram SR 3.3.5.32.6 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program PERRY - UNIT 1 3.3-46 Amendment No. 171

TEC H NICAL SPECIFICATION MARK.UP RCIC System lnstrumentation 3.3.5.32 Table 3.3.5.82-1 (page 1 of 1) Reactor Core Isolation Cooling System Instrumentation CONDITIONS REQUIRED REFERENCED CHANNELS FROM PER REQUIRED SURVEILLANCE FUNCTION FUNCTION ACTION 4.1 REQUIREMENTS ALLOWABLE VALUE

1. Reactor Vessel Water Leve! - 4 B sR 3.3.5.i?.1 > 127.6 inches Low Low, Level 2 sR 3.3.5.32.2 sR 3.3.5.33.3 sR 3.3.5.8a.4 sR 3.s.5.3e.s

2. Reactor Vessel Water Level - 4 C sR 3.3.5.32.1 s 221.7 inches High, Level I sR 3.3.5.32.2 sR 3.3.5.32.3 sR 3.3.5.32.4 sR 3.3.5.33.s

3. Condensate Storage Tank 2 D sR 3.3.5.32.1 > 90,300 gallons Level - Low sR 3.3.5.32.2 sR 3.3.s.93.3 sR 3.3.5.32.4 sR 3.3.5.92.5

4. Suppression Pool Water 2 D sR 3.3.5.32.1 S 18 ft 6 inches Level - High sR 3.3.5.92.2 sR 3.3.s.E2.3 sR 3.3.5.32.6 sR 3.3.5.32.5

5. Manual lnitiation C sR 3.3.5.32.5 NA PERRY - UNIT 1 3.3-47 Amendment No. 1-15

TECHNIGAL SPECIFICATION Primary Contalnment and Dryrell Isolation Instrumentation PROVIDED FOR CONTEXT J,3.6,I 3 .3 IHSTRUIIET{TATIOT{ 3.3. E. I Primary Contai nnent and Dryrrel l IsoI ation Instrumentation LCo 3.3.6. I The primary contai nment and dryuel I i sol at i on instrumentetion for each Function in TabIe 3.3.8.1*l shell be 0PERIIBLE. APPLICABILITY: Accordtng to rrble 3.3.6.I-I" ACTI0T{S COHDITIOI.I HEQUIREO ACTIOH TOHPLTTtrOI{ TIHI A. One 0r more required A.l Place channel in l2 hours for channel s inoperable. tri p. Functions e.b" 5.b, and 5.d

                                                                               &flp I{   hours for Functions sther than Funct,ions [.b, 5,b, and 5.d B    One or  more aut,omatic      B,I     Hest ore i so'lat i on           I  hour Functions with                       capahi J i ty.

isolatton capabil ity not maintained. (conti nuedI PERRY UI{IT I 3.3-{g Amendment t{o " 69

TECHNICAL SPECIFICATION MARK-UP Primary Containment and Drywell lsolation Instrumentation 3.3.6.1 ACTIONS continued CONDITION HEQUIFTED ACTION COTVIPLETION TIME J. As required by Required J.1 Initiate action to restore lmmediately Action C.1 and channel to OPERABLE referenced in status. Table 3.3.6.1-1 . OR J:2 Initiate a6tio mmeAra+ety Residual Heat Remeval

                                    /DLlD\      Qltr rlr{n'^,n
                                    \r rr rr rI ien R

mmedia*ety OPEFIABtE status, AND hqneediately ifl AND e ___+er_r+issible+Hd+ adm inistrative ee ntrel, lnitiate aetien te elese ene l#ffidia++ (continued) PEFIRY _ UNIT 1 3.3-51 Amendment No. 69

TECHNICAL SPECI FICATION MARK.UP Primary Containment and Drywell lsolation Instrumentation 3.3.6.1 ACTIONS continu CONDITION REQUIRED ACTION CO[/PLETION TI]UE K. As required by Required K.1 lsolate the affected lmmediately Action C.1 and penetration flow path(s) referenced in Table 3.3.6.1-1. OR K.2+ Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment. ANEI lmnqedie*ely lnffieC+a+dy Atin4;{-fld Table 3,3=6,1 1 , Ygs,s,g{= PERRY _ UNIT 1 3.3-52 Amendment No. {2

Primary Contai nment and Drywel I Isolation Instrumentation TECH NICAL SPECI FICATION 3.3.6. I PROVIDED FOR CONTEXT Table 3.3.6.1-1 (pgse 1 of 6) Prinrary tantaiffr*nt ard Sryxel't- Isotsti rn Instrurentatiqr APPI I EIBLE coilDtIIots I'lilES OR REOUIRES REFERE}ICED OTHER clnliltELs FEfl{ SPEII FIEO PER IRIP REOUI RED SURVEITLAIIEE ALLCI.IABLE Ful{cilofi EOTIDI TIO}IS SYSIEH fiETIOI{ E,I REOUINEHEIITS VALUE t, l,lsin Steam Line Isolation a Rcactor Vessel LJater 1r4,3 e D SR 3.5.6. 1.I E 14,3 inches Levet - Lou La* lox. SR 3,5.6. 1.? Level t SR r,5.6. 1.5 SR 3.!.6. 1.4 SR 3.r.6, 1.5 SR I.3.6. 1,6 b- l,lain Steam Line 2 E sR 5.5.6, .1 e ru5.? psis Pressure' LoH sn 5.3.6, .e sR 5.5.6. .5 sR 3.I.6, .4 sR 5,3.6. .5 sE 3.3.6. .6 c l,lain Stesm Line 1.e.5 E per tlst sn 5.5-6.1. I s 256,5 psfd s[ 5.3.6.I ,E D F tots - tligh sR 3.5.6,1 .3 s* 3,3.6.1.4 sR r.1.6.1.5 sR 3.5.6.1.6

d. Condenser vsctrun - LoH 1,e(al, z 0 SR r.5.6.Lr a 7.6 inches SR 3.5.6.I,a Hg vacun SR 3.5,6.1 .3 5(a) SR 3.I.6.I .4 sl 5.5.6. r .5 e Xain St,eun Line Fipe I rerI z D sR 3.3.6. ,l.1 < l5E,90t Tunnel Ternerature - sR LI.6. t.e H igh sR 3,3.6, r.4 sR 3,5.6. I.5 sR 3.3.6. 1.?

f . l,tain Steam Line 1,2,3 ?, D $R 3.3.6. .t s 149-6"F Iurbine Bui lding SR 3.5.6. ,? Teilperature-H igh SR 3.5.6, .4 SR 3.3.6. .5

g. l,larurat Initiation z 5 z E sR 3.5.6.t.5 llA z Primary Contairment and Dryrell IsoIat ion a Reactgr Vessel lJater I, e,3 z(b) il SR 5.3 6. .1 > 1e?.6 inches Level - Lor l-on, Level ? SR 3.3 6. -a

                                                                                            $[   5,5 6. .5 SR   T.T 6- .4 SR   3.5 6. R (continued)

(a) tJf th any turbim stop valve not closed. (b) Eequired to initiate thr sssociated dryt{etI isoletfon function. PTRRY - UNIT 1 3.3-54 Amendrnent No " 130

TECHNICAL S PECI FICATION Primary Containment and Drywell lsolation Instrumentation MARK.UP 3.3.6.1 Table 3.3.6.1-1 (page 2 of 6) Primary Containment and Drywell lsolation lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROhT SPECIFIED PER TFIIP REOUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 HEQUIREMENTS VALUE

2. Primary Containment and Drywell lsolation a--Ea+e#+esse+ {} .r{b} t sR 3,3,6,1,1 >-{+7$inhs Water f"

sE#r+ tveHv+-l+u6 sE=+*#L3

      @                                                                                sR 3,3,6=1,4 SE#

b. Drywell Pressure - High 1 ,2,3 2F) H sR 3.3.6.1.1 s 1.88 psig sR 3.3.6.1.2 sR 3,3"6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

c. Reactor VesselWater 1,2,3 2(b) F sR 3.3.6.1.1 > 14.3 inches Level - Low Low Low, sR 3.3.6.1.2 Level 1 (ECCS Divisions sR 3.3.6.1.3 1 and 2) sR 3.3.6.1.4 sR 3.3.6.1.5

{i 2tbi t sR 3,3,6,1,1 >-+4iflhs sR 3=3.6=1* sR 3,3,6,1,3 sR 3=3=6=1=4 sR 3,3,6,1,5

d. Drywell Pressure - High 1,2,3 2 F sR 3.3.6.1.1 < 1.88 psig (ECCS Divisions 1 and 2) sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

e. Reactor VesselWater 1,2,3 4 F sR 3-3.6.1.1 z 127.6 inches Level - Low Low, Level 2 sR 3.3.6.1.2 (HPCS) sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5 t) 4 t sR 3,3,6.1,1 >-++7$inhs sR 33.&1,2 SE-SS;l=3 sR 3,3,6,1,4 sR 3,3,6,1,5 f . Drywell Pressure - High 1,2,3 4 F sR 3_3.6.1.1 < 1.88 psig (HPCS) sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

g. Containment and Drywell 1,2,3 2(b) F sR 3.3.6.1.1 < 4.0 mR/hr Purge Exhaust Plenum sR 3.3.6.1.2 above Radiation - High sR 3.3.6.1.4 background sR 3.3.6.1.5 (continued)

(b) Required to initiate the drywell isolation function. (e) During eperatiens with a petential fer draining the reaeter vessel, PERRY _ UNIT 1 3.3-55 Amendment No. {42

TEGHNICAL SPECI FICATION MARK.UP Primary Containment and Drywell lsolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 6) Primary Containment and Drywell lsolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

2. Primary Containment and Drywell lsolation

g. Containment and Drywell (d) 2 K sR 3.3.6.1.1 s 4.0 mR/hr Purge Exhaust Plenum sR 3.3.6.1.2 above Radiation - High sR 3.3.6.1.4 background (continued) sR 3.3.6.1.5

h. Manual Initiation 1,2,3 2(b) G sR 3.3.6.1.5 NA (d) 2 K sR 3.3.6.1.s NA

3. Reactor Core lsolation Cooling (RCIC) System Isolation

a. RCIC Steam Line Flow - 1, 2,3 1 F sR 3.3.6.1.1 < 298.5 inches High sR 3.3.6.1.2 water sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

b. RCIC Steam Line Flow 1,2,3 I F sR 3.3.6.1.2 l 3 seconds and Time Delay sR 3.3.6.1.4 < 13 seconds sR 3.3.6.1.5

c. RCIC Steam Supply Line 1,2,3 1 F sR 3.3.6.1.1 > 55 psig Pressure - Low sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.s

d. RCIC Turbine Exhaust 1, 2,3 2 F sR 3.3.6.1.1 s 20 psig Diaphragm Pressure - SR 3.3.6.1.2 High SR 3.3.6.1.3 SR 3.3.6.1.4 sR 3.3.6.1.5

e. RCIC Equipment Area 1,2, 3 1 F sR 3.3.6.1.1 s 145.9'F Ambient Temperature - sR 3.3.6.1.4 High sR 3.3.6.1.5 sR 3.3.6.1.7

f. Main Steam Line Pipe 1,2, 3 1 F sR 3.3.6.1.1 < 158.9'F Tunnel Temperature - sR 3.3.6.1.4 High sR 3.3.6.1.s sR 3.3.6.1.7 (continued)

(b) Required to initiate the drywell isolation function (d)Duringmovementofrecentlyirradiatedfuel assemblies in primary containment. PERRY _ UNIT 1 3.3-56 Amendment No. {2

TECHNICAL SPECIFICATION MARK.UP Primary Containment and Drywell lsolation lnstrumentation 3.3.6.1 Table 3.3.6.1-1 (page 6 of 6) Primary Containment and Drywell lsolation lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 HEQUIREMENTS VALUE

5. RHR System lsolation

a. FIHR Equipment Area 2(e), 3(e) 1 per area F sR 3.3.6.1.1 s 15g.g"F Ambient sR 3.3.6.1.4 Temperature - High sR 3.3.6.1.5 sR 3.3.6.1.7

b. Fleactor Vessel Water 1, 2(o) , 3(9) 2 F sR 3.3.6.1.1 > 177.1 inches Level - Low, Level 3 sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5 2(e), 3(e) ,4, E 2#) J sR 3.3.6.1.1 > 177.1 inches sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

c. Reactor Vessel Steam 1,2,3 2 F sR 3.3.6.1.1 < 150 psig Dome Pressure - High sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sH 3.3.6.1.5

d. Drywell Pressure - High 1, 2, 3 2 F sH 3.3.6.1.1 s 1.88 psig sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

e. Manual lnitiation 1,2,3 2 tl sR 3.3.6.1.5 NA (e) With reactor vessel steam dome pressure less than the RHH cut in permissive pressure.

(f) Only ene trip system required in MODES 1 and 5 with RHR Shutdewn eeeling System integrity maintained, (g) With reactor vessel steam dome pressure greater than or equal to the RHB cut in permissive pressure. PERRY _ UNIT 1 3.3-59 Amendment No 85

CRER System Instrumentatlon TECHNICAL SPECI FICATION 3,3 - 7.1 PROVIDED FOR GONTEXT 3.3 I HSTRUHEHTAT IOI{ 3.3 "T.l Control Room Emergency Recirculatlon (CRER) System Instrumentrtioa LCo 3.3.7,I The Cf,ER System instrumentation for each Functlon ln Tahle 3.3.7.1-l sha'lI be 0PERABLE. APPLICABILIil: Accordi ng to Tabl e 3.3.7.I-I . ACTI0T{S COI,IDI T trOH ffiqUIRED ACTI0t'l TOI.IPLETIOH TIHE A One or more channel s A.I Enter the Condttion Immed i ately i noperabl e. referenced in Table 3.3.7.I-l for the channel. 8" As r*quired by B. r Declare as$osiated I hour from Hequired Action A.l CRER subsystem discovery of and referenced in t noperabl e" loss of CRER Tabl* 3.3.7.I-l. initiation capabil ity in huth trip systems ffiu B.E Place channel ln 24 hours trip" {conti nued} pERRy UilIT I 3 .3-70 Amendment Ho. 69

TEC H NICAL SPECI FICATION MARK.UP CRER System lnstrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1) Control Room Emergency Recirculation System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PEH TRIP HEQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIHEMENTS VALUE

1. Reactor VesselWater 1, 2, 3; 2 B sR 3.3.7.1.1 > 14.3 inches Level - Low Low Low, {} sR 3.3.7.1.2 Level 1 sR 3.3.7.1.3 sR 3.3.7.1.4 sR 3.3.7.1.5

2. Drywell Pressure - High 1,2,3 2 B sR 3.3.7.1.1 < 1.88 psig sR 3.3.7.1.2 sR 3_3.7.1.3 sFl 3.3.7.1.4 sR 3.3.7,1.5

3. Control Room Ventilation 1,2,3, C sR 3.3.7.1.1 s 800 cpm Radiation Monitor (b) sR 3.3.7.1.2 sR 3.3.7.1.4 sFr 3.3.7.1.5 (a) During eperatiens with a petentialfer draining the reaeter vessel, (b)Duringmovementofrecentlyirradiatedfuel assemblies in the primary containment or fuel handling building.

PERRY _ UNIT 1 3.3-73 Amendment No. {42 I

TECH NICAL SPECIFICATION MARK-UP ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CflIEOL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of eight safety/relief valves shall be OPERABLE" APPLICABILITY: [/oDE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure s 150 psig. ACTIONS NOTE--- LCO 3.0.4.b is not applicable to HPCS. CONDITION REQUIRED ACTION COTVIPLETION TITUE A. One Iow pressure ECCS A.1 Restore low pressure 7 days injection/spray su bsystem ECCS injection/spray inoperable. subsystem to OPERABLE status. B. High Pressure Core 8.1 Verify by administrative t hour Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC is required to be OPERABLE. AND 8.2 Restore HPCS System to 14 days OPERABLE status. (continued) PERRY - UNIT 1 3.5-1 Amendment No. {$t I

TECH NICAL SPECIFICATION MARK.UP 3.5.2 3.5 ETUERGENCY CORE COOLTNG SYSTET\4S (ECCS), RPV WATER TNVENTOW cflIEOL AND REACTOR CORE TSOLATTON COOLTNG (HCIC) SYSTETVT 3.5.2 LCO 3.5.2 Onnlru fln/f ot nPV wat he:Sg_heus-AND TreOrc ECCS injection/spray subsystems shall be OPERABLE. n low Pressure Cool OPf nnelE during al capable ot being ma APPLICABILITY: MODES 4-,_and5 ACTIONS CONDITION REQUIRED ACTION CO]\/PLETION TIME A. On+rEequired ECCS 4.1 Restore required ECCS 4 hours i njection/spray su bsystem i njection/spray su bsystem inoperable. to OPERABLE status. B. Required Action and 8.1 lnitiate action to suspenC lmmediately associated Completion Time of Condition A not @ met. a method of water l ion capable of operating without offsite elffi po_\ /et. c.@ c.1 i+je+in/spra# subsys+effis containment boun is inserab{+DRAIN TIIUE capable.ltbeing

     < 36 hours and > I hours.               established in les DMIUIIME PERRY      - UNIT   1                              3.5-6                         Amendment No. 69

TEC HNICAL SPECI FICATION MARK.UP 3.5.2 CONDITION REQUIRED ACTION COMPLETION TIIUE AND c.2 Restere ene ECCS 4 hours in te OPEFIAEtE statu+ Verifv each primarv containment pene ion ttow patn is capaUte U being isolated in lessthan the DRAIN TllVIE. (continued) PEFIRY _ UNIT 1 3.5-6 Amendment No. 69

TECHNICAL SPECIFICATION MARK.UP 3.5.2 ACTIONS continued CONDITION REQUIRED ACTION CO[/PLETION TIME D. D.1 NOTE---

   @                        neouireO ECCS
   @                        iniection/sprav s TIIME < I hours.        additional method iniection sndt be operating without electrical oower, lnitiate aetien te restere  Immediately action to establis adOitional metnoO of iniection with wat capabte of maintai water level > TAF for E 36 hours.

AND D,2 lnitiate action to rester+ lmmediately i requi+edggtaEluhprimary co ntai n m e nt penetra*ien

                           +teu+pa*n+e+

isdte+broEdarv. AND D.3 lmmediately action to isolate e primarv containmeft penetration flow o verifv it can be manuallv isolated from the o room. PERRY _ UNIT 1 3.5-7 Amendment No. 171

TECH NICAL S PECI FICATION MARK.UP 3.5.2 CONDITION REQUIRED ACTION COMPLETION TITUE E. nequired Action aM E;t lnitiate action to lnmedrateiv associale.dGompletion DRAIN TIME to > S0 ho Time of Condition 0 or DTIIeL OE DRAINTII\tlE<t hour. SU RVE ILLANCE REQU I REIVIENTS SURVEILLANCE FREQUENCY SE-312J Veritv ORAIN TIIUE > Il-accordarce lvrlh-the Survcillanee Freouencv Control Prooram sR 3.5.2.?+ Verify, for aeaeh required low pressure ECCS In accordance injection/spray subsystem, the suppression pool with the water level is > 16 ft 6 in. Surveillance Frequency Control Program (continued) PEHHY _ UNIT 1 3.5-7 Amendment No. 171

TEGHNIGAL SPECI FICATION MARK.UP 3.5.2 SU RVEI LLANCE REQU I RE]VIENTS continued SUHVEILLANCE FREQUENCY sR 3.5.2.32 Verify, for athe required High Pressure Core Spray ln accordance (HPCS) System, the: with the Surveillance

a. Suppression pool water level is > 16 ft 6 in; or Frequency Control Program

b. Condensate storage tank water volume is

                     > 249,700 gal.

sR 3.5.2.43 Verify, for theeash required ECCS injection/spray ln accordance subsystem, the piping is filled with water from the with the pump discharge valve to the injection valve. Surveillance Frequency Control Program sR 3.5.2.54 ing Verify=-lor lheeash req u red EC CS i nj ectio n/spray i ln accordance subsystem* eaeh manual, power operated, and with the automatic valve in the flow path, that is not locked, Surveillance sealed, or otherwise secured in position, is in the Frequency correct position. Control Program (continued) PERRY _ UNIT 1 3.5-8 Amendment No. 171

TECH NICAL S PECIFIGATION MARK.UP 3.5.2 SU RVEI LLANCE REQU I FIEN/ENTS continued SURVEILLANCE FREQUENCY sR 3.5.2.65 Operate tne requir Itriccordance sunsvstem through udthlhe

                   > t O minutes.                                      Surveillance Ercguency Control Prooram whieh ineludes the speeified reaeter te eentainment  {n+e+dene INSERVICE REAGTER TE          TESTING CONTAINMENT          PRGRAM WETWEtt DIFFERtrNTIN  t SYSTEM FtOW RATE                PRESSUHE tPGS        >+l-lftpm        ++ee+sid tPGI        >++egpm          --eq+sid HPCS         >+l-l+gpm       +g+sie SEI-5ZJ           Veritv eacn vaUe c                                   lfficcor4grce a penetration flo                            im      wtlh-lhe positlon on an actual                                Sursillanc.e ftequency Control Program sR 3.5.2.96 Verify tne requireO l-PCI o                          ln accordance actuates on a manua                                  with the reouired HPCS subsvstem can be manuallv              Surveillance ooerateO.                                            Frequency Control Program PERRY _ UNIT   1                             3.5-9                     Amendment No. 175

TECHNICAL S PECIFICATION MARK.UP RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTOW CWIEOT AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE. APPLICABILITYT MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig. ACTIONS TE LCO 3.0.4.b is not applicable to RCIC. CONDITION REQUIRED ACTION COMPLETION TITVIE A. RCIC System inoperable. 4.1 Verify by administrative t hour means High Pressure Core Spray System is OPERABLE. AND 4.2 Restore RCIC System to 14 days OPERABLE status. B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time not met. AND 8.2 Reduce reactor steam 36 hours dome pressure to < 150 psig. PERRY - UNIT 1 3,5-10 Amendment No. {#t I

TECH NICAL SPECIFICATION MARK.UP Primary Containment Air Locks 3.6.1 .2 3.6 CONTAINMENT SYSTEIUS 3.6.1 ,2 Primary Containment Air Locks LCO 3.6.1.2 Two primary containment air locks shal! be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment=; {ePgEV+)=

1. Entry and exit is permissible to perform repairs of the affected air lock cornponents.

2. Separate Condition entry is allowed for each air lock.

e Enter applicable Conditions and Required Actions of LCO 3.6.1."1, "Primary Containment-Operating," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3. CONDITION HEQUIRED ACTION COMPLETION TITUE A. One or more primary -------NOTES------ containment air locks with 1. Required Actions A.1, A.2, one primary containment and A.3 are not applicable if both air lock door inoperable. doors in the same air lock are inoperable and Condition C is entered.

2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.

continued PERRY _ UNIT 1 3.6-3 Amendment No. 6gr-g

TECHNICAL SPECIFICATION MARK.UP Primary Containment Air Locks 3.6.1 .2 ACTIONS CONDITION REQUIRED ACTION COTVIPLETION TII\/E C. (continued) c.3 Restore air lock to 24 hours OPERABLE status D. Required Action and D.1 Be in IUODE 3. 12 hours associated Completion Time of Condition A, B, or AND C not met in [\4ODE 1 , 2, or 3. D.2 Be in IVIODE 4. 36 hours E. Required Action and E.1 Suspend movement of Immediately associated Completion recently irradiated fuel Time of Condition A, B, or assemblies in the primary C not met during containment. movement of recently irradiated fuel assemblies AAIEI in the primary containment.fr4u+ing l##diately e)PEhEl\fu= PBEI/tr PERRY _ UNIT 1 3.6-6 Amendment No. 85 I

TEGHNICAL SPECI FICATION F[IUs PROVIDED FOR CONTEXT 3 6.1.3

3. 6 C0I{TAI[{1'IEHT SYSTEI{s

3. 6. I . 3 Primary fionta i rrment Isol at i on Ual ves (PCIUS]

LCo 3"6.1.3 Each P[It,, except containment Yacuum breakers, sha'l] be OPERABLE. APPLICABILITY: I{0DES l, 2, and 3, tlhen associated instrumentation ts requtred to be 0FIRABLE per LCO 3. 3,6.1, nPrlmary Containment and Drywell Isol ati on Instrumentati on.

ACTIoilS

                                      -;;;il
          ; ; ; ;; ; ;;ilil ; ; il ;

l;penetration ;;;ntffi ' il; ; ;; l;; ; il [ ; ; ;;;

  -                                                                -42         -    -

I: flor paths may be unisolated intermittently under admini strati ve controls.

2. Separate Condition entry ls allowed for each penetration flow path.

3. tnter npplicab'le Conditions and Required Actions for system$ rade inoperable hy P[IUs.

4. Enter applicable Conditions and Eequired Actions of LCO 3.6.1.1, nPrimary

[ontainmgnt{lperatif,gr' when PCIV leahrge results in exceeding overall containment leakage rrte acceptance criteria in il0DES l, ?, and 3. COHOIT IOt{ TEOUIRTil ATTIOH COHPLETIOH T I[,IE A. One or more A.I Isolate the affected 4 hours except penetration flow paths p+netration flow path for main steam rtith one FCIU by use of at least line' ino p erable exc*Bt due DnE closed and de-to 1 eakage not uli th i n acttvated automatic AHD limit. ualve, closed manual rual ve, bI ind f I ange, I hours for rnain or check val ve xi th steam 1 ine tlow through the vatr ve secured. [uD { cont t nued } PTRRY UHIT I 3.6-9 Amendment t{0. 69

TECHNIGAL S PECIFICATION MARK.UP PCIVs 3.6.1.3 ACTIONS CONDITION HEQUIRED ACTION COMPLETION TIME D. (continued) D.3 Perform SR 3.6.1.3.6 for Once per 92 days the resilient seal purge valves closed to comply with Hequired Action D.1. E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, B, AND C, or D not met in IVODE 1, 2, or 3. E.2 Be in MODE 4. 36 hours F. Flequired Action and F.1 Suspend movement of Immediately associated Completion recently irradiated fuel Time of Condition A, B, assemblies in primary C, or D not met for containment. PCIV(s) required to be OPERABLE during movement of recently irradiated fuel assemblies in the primary containment. {en+inued} PERRY _ UNIT 1 3.6- 1 3 Amendm.nl N1s. {49

TECHNICAL SPECIFICATION MARK.UP PCIVs 3.6.1 .3 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME l#ffid+I{y

   @                            )PDEVs" gR req+,tim
   @                                          l#ffiedilely s+a*+s Intentionallv blank PERRY _ UNIT   1                  3.6-14          Amendment No. {2   I

TEC H N ICAL SPECIFICATION MARK-UP Primary Contai nment-Shutdown 3.6.1 .10 3.6 CONTAINIVIENT SYSTEMS 3.6.1 .1 0 Primary Containment-Shutdown LCO 3.6.1.10 Primary containment shall be OPERABLE. APPLICABILITY: During movement of recently irradiated fuel assemblies in the primary containment*; Buring eperatiens with a petential fer draining the reaeter vessel PS*V$, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment 4.1 Suspend movement of lmmediately inoperable. recently irradiated fuel assemblies in the primary containment. A,$IEI l##eCje+ely

                                          )PDRffu=

PERRY - UNIT 1 3.6-29 Amendment No. 42 I

TECHNIGAL SPECIFICATION Containment Vacuum Breakers MARK-UP 3.6.1 .1 1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 1 Containment Vacuum Breakers LCO 3.6.1 .1 1 Three containment vacuum breakers shall be OPERABLE and four containment vacuum breakers shall be closed. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment=; During eperatiens with a petential fer draining the reaeter vesse! (gpsR\tuI ACTIONS Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating" when the containment vacuum relief subsystem leakage results in exceeding overall containment leakage acceptance criteria. CONDITION REQUIRED ACTION COMPLETION TIME A. .NOTE-. A.1 Close the associated motor 4 hours Separate Condition entry operated isolation valve. is allowed for each containment vacuum AND breaker. 4.2 Restore required 72 hours containment vacuum One or two containment breaker to OPERABLE vacuum breakers not status. closed. OR One required containment vacuum breaker inoperable for other reasons. (continued) PERRY _ UNIT 1 3.6-31 Amendment No. 111 I

TECHNICAL S PECIFICATION MARK.UP Containment Vacuum Breakers 3.6.1 .1 1 ACTIONS continu CONDITION REQUIHED ACTION COTVIPLETION TI]VIE B. Required Action and associated Completion ;ilililH"'J5,i ;-, ; ;;; Time of Condition A not ";,, met. 8.1.1 Be in TUODE 3. 12 hours OH AUD Three or more containment vacuum 8.1.2 Be in MODE 4. 36 hours breakers not closed. AND OR Two or more required containment vacuum

                            ;;rilffi-[?Iil;;;;;;i of recently irradiated fuel breakers inoperable for  assemblies in the primary other reasons.           containment=@

B.2J Suspend movement of lmmediately recently irradiated f uel assemblies in the primary containment. {.mffiedi+ely GIPDHVS: PERRY _ UNIT 1 3.6-32 Amendment No. {2

TECH NICAL SPECI FICATION MARK.UP Containment H um idity Control 3.6.1.12 3.6 CONTAINMENT SYSTETVIS 3.6.1 .12 Containment Humidity Control LCO 3.6.1.12 Containment average temperature-to-relative humidity shall be maintained within limits. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment=; (gPgRVs)= ACTIONS CONDIT!ON REQUIRED ACTION COMPLETION TIME A. Requirements of LCO not 4.1 Restore containment 8 hours met. average tem perature-to-relative humidity to within limits. (continued) PERRY. UNIT 1 3.6-34 Amendment No. 111

TECH NICAL SPEGIFICATION MARK.UP Containment Humidity Control 3.6.1 .1 2 ACTIONS continu CONDITION REQUIRED ACTION CONTPLETION TITUE B. Required Action and 8.1 Be in [\4ODE 3. 12 hours associated Completion Time of Condition A not AND met in N4ODE 'l , 2, or 3. 8.2 Be in TUODE 4, 36 hours C. Required Action and c.1 Suspend movement of Immediately associated Completion recently irradiated fuel Time of Condition A not assemblies in the primary met during movement of containment. recently irradiated fuel assemblies in the primary ANEI co nta n m e nt.;++4u+ing i

    )PDEVS;                                                                  l#meC+a*ely
                                              )pDEl/s; SURVEI LLANCE REQU I REMENT SURVEILLANCE                                         FREQUENCY sFl 3.6.1 .12.'r     Verify contai n m ent average tem peratu re-to-relative     In accordance humidity to be within limits.                              with the Surveillance Frequency Control Program PERRY _ UNIT       1                             3.6-35                           Amendment No. '171      I

TEGHNICAL SPECIFICATION MARK.UP Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE. APPLICABILITY: IVIODES 1, 2, and 3, During movement of recently irradiated fue! assemblies in the primary containment=; Dsring eperatiens with a petentiel fer draining the reaeter vessel psRV+ ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondarycontainment 4.1 Restore secondary 4 hours inoperable in MODE 1, 2, containment to or 3. OPERABLE status. B. Required Action and 8.1 Be in MODE 3 12 hours associated Completion Time of Condition A not AND met. 8.2 Be in MODE 4. 36 hours (continued) PERRY _ UNIT 1 3.6-51 Amendment No. {2 I

TECHNIGAL SPECIFICATION MARK.UP Secondary Containment 3.6.4.1 ACTIONS contin CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment c.1 Suspend movement of Immediately inoperable during recently irradiated fuel movement of recently assemblies in the primary irradiated fuel assemblies containment. in the primary containment.pr4u+ing ANE E)PDRVs. finmeeikly lPDRIfu" SURVEI LLANCE REQUI REMENTS SURVEILLANCE FREQUENCY sR 3.6.4.1.1 Verify secondary containment vacuum is > 0.66 inch In accordance of vacuum water gauge. with the Surveillance Frequency Control Program sR 3.6.4,1 ,2 Verify the primary containment equipment hatch is ln accordance closed and sealed and the shield blocks are with the installed adjacent to the shield building. Surveillance Frequency Control Program sR 3.6.4.1.3 Verify each secondary containment access door is ln accordance closed, except when the access opening is being with the used for entry and exit. Survei!lance Frequency Control Program PERRY _ UNIT 1 3.6-52 Amendment No. 171

TECHNICAL SPECIFICATION MARK.UP SCIVs 3.6.4.2 3.6 CONTAINTVIENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) LCO 3.6.4.2 Each SCIV shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment*; {sPsRVs} ACTIONS 1 Penetration flow paths may be unisolated intermittently under administrative controls. 2 Separate Condition entry is allowed for each penetration flow path. 3 Enter applicable Conditions and Required Actions for systems made inoperable by SCIVS. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration 4.1 lsolate the affected 8 hours flow paths with one SCIV penetration flow path by inoperable. use of at least one closed manual valve or blind flange. AND continued PERRY - UNIT 1 3.6-53 Amendment No. 42

TECHNICAL SPECI FICATION MARK.UP SCIVs 3.6.4.2 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Suspend movement of lmmediately associated Completion recently irradiated fuel Time of Condition A or B assemblies in the primary not met during movement containment. of recently irradiated fuel assemblies in the primary ANEI containment.pr4u+ing CPDRVS" l#ffiediately tl2BRlls" SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY sR 3.6.4.2.1 NOT

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

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

Verify each secondary containment isolation manual In accordance valve and blind flange that is not Iocked, sealed, or with the othennrise secured and is required to be closed Surveillance during accident conditions is closed. Frequency Control Program PERRY - UNIT 1 3.6-55 Amendment No. 171

TEGHNICAL SPECIFICATION MARK-UP AEGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System LCO 3.6.4.3 Two AEGT subsystems shall be OPERABLE APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment=; During eBeratiens with a Betential fer draining the reaeter vessel PDR\tuI ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One AEGT subsystem A.1 Restore AEGT subsystem 7 days inoperable. to OPERABLE status. B. Required Action and 8.1 Be in MODE 3 12 hours associated Completion Time of Condition A not AND met in MODE 1, 2, or 3. 8.2 Be in MODE 4. 36 hours C. Required Action and c.1 Place OPERABLE AEGT lmmediately associated Completion subsystem in operation. Time of Condition A not met during movement of OR recently irradiated fuel assemblies in the primary co nta i n m e nt r-er-d+*ring PDRlls-continu PERRY - UNIT 1 3.6-56 Amendment No. 42

TEC HNIGAL SPECIFICATION MARK-UP AEGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COTVIPLETION TIME C, (continued) c.21 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment. NND mmedia+ety PBRllsr D. Two AEGT subsystems D.1 Enter LCO 3.0.3. lmmediately inoperable in MODE 1,2, or 3. E. Two AEGT subsystems E.1 Suspend movement of lmmediately inoperable during recently irradiated fuel movement of recently assembles in the primary irradiated fuel assemblies containment. in the primary contai n ment r++du+ingt ANE OPDRVS-mmeAie*ety CPEhRVS-PERRY - UNIT 1 3.6-57 Amendment No, 42

TECHNICAL S PECI FICATION MARK-UP CRER System 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Control Room Emergency Recirculation (CRER) System LCO 3.7.3 Two CRER subsystems shall be OPERABLE. NOT tr The Control Room Envelope (CRE) boundary may be opened intermittently under administrative control. APPLICABILITY: IUODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment or fuel handling buildifig=, pPDRVS+" ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRER subsystem A.1 Restore CRER subsystem 7 days inoperable for reasons to OPERABLE status. other than Condition B. B. One or more CRER 8.1 Initiate action to implement lmmediately subsystems inoperable mitigating actions. due to inoperable CRE boundary in Mode 1, 2, AND or 3. 8.2 Verify mitigating actions 24 hours ensure CRE occupant rad iolog ical exposures will not exceed limits, and CRE occupants are protected from chemical and smoke hazards. AND 8.3 Restore CRE boundary to 90 days OPERABLE status. (continued) PERRY - UNIT 1 3.7-4 Amendm*n1 51s. {4.8 t

TECH NICAL SPECI FICATION MARK-UP CRER System 3.7.3 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B AND not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours D. Required Action and associated Completion LCO 3.0.3 is not applicable. Time of Condition A not met during movement of recently irradiated fuel D.1 Place OPERABLE CRER Immediately assemblies in the primary subsystem in emergency containment or fuel recirculation mode. handling building.p+

    @                         OR D.2+   Suspend movement of       lmmediately recently irradiated fuel assemblies in the primary containment and fuel handling building.

AND l#ffiediately CIPDRVS; E. Two CRER subsystems E.1 Enter LCO 3.0.3. lmmediately inoperable in MODE 1, 2, or 3 for reasons other than Condition B. (continued) PERRY - UNIT 1 3.7-5 Amendment No. 148

TEGHNICAL SPECIFICATION MARK.UP CRER System 3.7.3 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME F. Two CRER subsystems F.1 Suspend movement of lmmediately inoperable during recently irradiated fuel movement of recently assemblies in the primary irradiated fuel assemblies containment and fuel in the primary handling building. containment or fuel handling building.r++ ANB l#ffieC+ate{y OR lPDR\ls" One or more CRER subsystems inoperable due to inoperable CRE boundary during movement of recently irradiated fuel assemblies in the primary containment or fuel handling building=pr SU RVEI LLANCE REQUI REMENTS SURVEILLANCE FREQUENCY sR 3.7.3.1 Operate each CRER subsystem for > 10 continuous ln accordance hours with the heaters operating. with the Surveillance Frequency Control Program sR 3.7.3.2 Perform required CRER filter testing in accordance ln accordance with the Ventilation Filter Testing Program (VFTP). with the VFTP (continued) PERRY _ UNIT 1 3.7-6 Amendment No. 171

TECH NICAL SPECI FIGATION MARK.UP Control Room HVAC System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Heating, Ventilating, and Air Conditioning (HVAC) System LCO 3.7.4 Two control room HVAC subsystems shall be OPERABLE. APPLICABILITY: I/ODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building=; (gPgRV+)= ACTIONS CONDITION REQUIRED ACTION COt\4PLETION TIIVIE A. One control room HVAC 4.1 Restore control room 30 days subsystem inoperable. HVAC subsystem to OPERABLE status. B. Two control room HVAC 8.1 Verify control room air Once per 4 hours su bsystems inoperable. temperature is < 90"F. AND 8.2 Restore one control room 7 days HVAC subsystem to OPERABLE status. C. Required Action and c.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B AND not met in MODE 1,2, or C.2 Be in MODE 4. 36 hours (continued) PERRY - UNIT 1 3.7-8 Amendment No. 442

TECH NICAL SPECI FICATION MARK.UP Control Room HVAC System 3.7.4 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion LCO 3.0.3 is not applicable. Time of Condition A not met during movement of recently irradiated fuel D.1 Place OPERABLE control lmmediately assemblies in the primary room HVAC subsystem in containment or fuel operation, handling building=p+

    @                         OR D.21   Suspend movement of           Immediately recently irradiated fuel assemblies in the primary containment and fuel handling building.

AND lffiffieCi+ely

                                     'PDRVS-(continued)

PERRY - UNIT 1 3.7-9 Amendment No. 4S2 I

TEC H NICAL SPECIFICATION MARK.UP Control Room HVAC System 3.7.4 ACTIONS contin CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and associated Completion LCO 3.0.3 is not applicable. Time of Condition B not met during movement of recently irradiated fuel E.1 Suspend movement of Immediately assemblies in the primary recently irradiated fuel containment or fuel assemblies in the primary handling building.pr containment and fuel

    @                                       handling building.

AND mmeaie+ety CIPDRVS; SURVEI LLANCE REQU I RETVIENTS SURVEILLANCE FREQUENCY sR 3.7.4.1 Verify each control room HVAC subsystem has the ln accordance capability to remove the assumed heat load. with the Surveillance Frequency Control Program PERRY _ UNIT 1 3.7-10 Amendment No, 171

TECH NICAL SPECI FICATION AC Sources - Shutdurn 3.8.2 PROVIDED FOR GONTEXT 3.8 ELECTRICAL POI.IER SYSTETfi 3.8,2 AC Sources

                    -Shutdnnr LCo 3.8.2          The following AC       electrical power sources shall be 0PERABLET a  . One qua'li fied ci rcuit between the sffsite transmi ssion network and the onsite Class lE AC electrical porver distribution subsystem(s) required by LCO 3.8.8.
                           "Di stri buti on Systems - Shutdmnn'  :

b. &re diesel generator (DG) capable of supplying one division of the Division I or 2 onsite C]abs IE AC electr! ca1 poder di stri but'i on subsystem( s ) requi red hy LCO 3,8.8: and c One gualified circuit, othen than the cjrcuit in LCO 3.8.2.a, betueen the offsite transmission netrlork and the Division 3 onsite Class lE electrical Dfiiler distribution subsystem, or the Division 3 DG capable of supplying the Division 3 onsite Class lt AC elehtrical por,rer di stri buti on subsystem, when the Di vl s t on 3 onsi te Cl ass lE el eclfi gal pu,rbr di stri buti on subsystem i s requi red by LCO 3.8. B.

APPL IIABIL ITY : tffiDtS 4 and 5 , During nrcvement of recently irradiated fuel assemblies in the pri mary contai nment 0r fuel hand'l i ng bui I di ng . PERRY - I}IIT 1 3.8-17 TfiffiTENT NO. 102

TECHNICAL SPECI FICATION MARK.UP AC Sources - Shutdown 3.8.2 ACTIONS NOTE------ LCO 3.0.3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIIUE A. LCO ltem a not met. --------NOTE-------- Enter applicable Condition and Flequired Actions of LCO 3.8.8, when any required division is de-energized as a result of Condition A A.1 Declare required feature(s) lmmediately with no offsite power available from a required circuit inoperable. OR A.2.1 Suspend CORE lmmediately ALTERATIONS AND 4.2.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment and fuel handling building. AND

                                                                 +mmegie{ely AND (continued)

PEFTRY _ UNIT 1 3.8-18 Amendment No. {42

TECH NICAL SPECIFICATION MARK-UP AC Sources - Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION CO[/PLETION TIME A. (Continued) A.2.94 lnitiate action to restore lmmediately required offsite power circuit to OPERABLE status. B. LCO Item b not met. 8.1 Suspend CORE Immediately ALTERATIONS. AND 8.2 Suspend movement of lmmediately recently irradiated fuel assemblies in primary containment and fuel handling building. AND l#ffiedia+ety IPDEVS; AAIS B.g4 lnitiate action to restore lmmediately required DG to OPERABLE status. C. LCO ltem c not met. c.1 Declare High Pressure 72 hours Core Spray System inoperable. PERRY _ UNIT 1 3.8-19 Amendment No. 12

TEC H N ICAL SPECI FICATION MARK.UP AC Sources - Shutdown 3.8.2 SUFIVEI LLANCE REQU I REMENTS SURVEILLANCE FREQUENCY sR 3.8.2.1

                     ; i;;il;;,ffi;;t3l5i;;;;i; performed:

SR 3.8.'1.3, SR 3.8.1.8 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 the associated ECCS subsystem(s) are not required to be OPEHABLE per LCO 3.5,2, "

Pressure Vessel (R Control." ln accordance with applicable For AC sources required to be OPEFIABLE, the SRs following SRs are applicable: sR 3.8.1.1 sR 3,8.1.7 sR 3.8.1.'r4 sR 3.8.1.2 sR 3.8.1 .9 sR 3.8.1.15 sFl 3.8.1 .3 sR 3.8.1.10 sH 3.8.1.16 sFl 3.8.1 .4 sR 3.8.1.11 sH 3.8.1 .18 sR 3.8.1 .5 sR 3.8,1 .12 sR 3.8.1 .19 sR 3.8.1 .6 sR 3.8.1.13 PERRY _ UNIT 1 3.8-20 Amendment No. 164

t TECHNIGAL S PECI FICATION PROVIDED FOR CONTEXT 0C Sources

                                                                              -Shutdown 3.8.5

3. B 'ELECTRICAL POHER SYSTEHS 3.8.5 DC Sources *Shutdown LCo 3 .8. 5 The following DC electrica1 por.ler subsystems shalI be OPERABLT:

a. 0ne Class lE DC electrical power subsystem catable of supplyjng one d'ivision of the Divisio-n I 0r 2'ons'ite Cl ass lE el ectri cal powef d'istributi on subsystem(s )

required by LCO 3.8.8. "0istribution Systenrl Shutdown";

b. One Class 1E_battery or battery charger. other than the DC electrical power subsystem in LCO 3.8.5.a, capable of supplying the i.emajning Division 1 or Div'ision 2'onsite Class 1E DC electrical power distribution subsystem when required by LCO 3.8.8; and

c. The D'ivision 3 DC electrical Dolver subsystem caDable of supplying the Division 3 onsr'te Class l-E DC eleLtrical t

poi.ler distribution subsystem, when the Division 3 onsite Class IE DC electrical 'power distributjon subsystem 'is requ'i red by LCO 3 . I . B. APPLICABILITY: IfiDES 4 and 5, During movement of recently irradiated fuel assemblies in the primary contai nment or fuel handl i ng bui I di ng. I I PIRRY - UNIT 1 3. B-28 Amendment No. toz,

TECHNICAL SPECIFICATION MARK.UP DC Sources - Shutdown 3.8.5 ACTIONS LCO 3.0.3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TITVIE A. One or more required DC 4.1 Declare affected required Immediately electrical power feature(s) inoperable. subsystems inoperable. OR 4.2.1 Suspend CORE lmmediately ALTERATIONS. AND 4.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the primary containment and fuel handling building. AND tmme+iatety vessel=

                            --+AlD A.2.94 lnitiate action to restore  lmmediately required DC electrical power subsystems to OPERABLE status.

PERRY _ UNIT 1 3.8-29 Amendment No. 442

TECH NICAL SPECI FICATION Di stri buti on Systems - Shutdoutn 3.8.8 PROVIDED FOR CONTEXT 3.8 ELECTRICAL Pfl.IER SYSTEFIS 3.I . I Di stri buti on Systems - Shutdown LCo 3,8.8 The necessary oorti ons of the 0l vi si ott I , Di vi s'i on 2 , and Di vi s i on 3 Af hnd tlC el ectri cal po{er di strl buti on subsystems shall be OPERABLE to bupport equi pmnt requ'ired to be OPERABLE-APPLICABILITY; ffiIDES 4 and 5. lluri ng moyement of recently i rradi ated fuel assemhl i es in the primary containment or fuel handling building. AITIO}IS HOTE LCO 3.0.3 is not applicable. COilDITIOI,I RIffJIRED ACTION CI}I"IPLETION TIHE A One or more requi red A.t Decl are associ ated Inmedr ately AC or DC electrical supported requi red polner di stri buti on feature(s ) subsystems i noperabl e. i noperabl e, E& A, A,l Suspend CffiE ilunedi atel y ALTERATIONS. AilD h.z .? Suspend movemnt of Irrunedi atel y recent'ly i rradi ated fuel assembl i es i n the primary containrnent and fuel handl i ng bui I di ng. AHE (conti nued) PEHRY - IT.Iil 1 3.8-38 Frerffi hh" 10?

TEC HNICAL SPECIFICATION MARK.UP Distribution Systems - Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIt\4E A. (continued) l#rcCiate{y

                                ----+AtD 4.2.?4 lnitiate actions to restore        Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.

AND 4.2.# Declare associated lmmediately required shutdown cooling su bsystem (s) i noperable and not in operation. SU RVEILLANCE HEQU I REMENTS SUHVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and voltage to ln accordance required AC and DC electrical power distribution with the subsystems. Surveillance Frequency Contro! Program PERRY _ UNIT 1 3.8-39 Amendment No. 171

Attachment 3 L-17-04s REVISED TEGHNICAL SPECIFICATION PAGES (FOR INFORMATION ONLY) (68 pages follow)

TECH NICAL SPECI FICATION INFORMATION ONLY CLEAN, RETYPED PAGE TABLE OF CONTENTS 1 0 USE AND APPLICATION 1 1 Def initions. .. .. 1.0-1 1 2 Logical Connectors 1.0-8 1 3 Completion Times..... 1.0-1 1 1 4 Frequency..... 1.0-24 2.0 SAFETY LItvllTS (SLs) 2.1 SLs. 2.0-1 2.2 SL Violations.... 2.0-1 3.0 LrtulTlNG coNDrrroN FoR opERATtoN (LCo) AppLtcABtLlTy. ...3.0-1 3.0 SURVEILLANCE HEQUIRETUENT (SR) AppLrCABtLtTy. 3.0-4 3.1 REACT IVITY CONTRO L SYSTEN/S 3.1 .1 SHUTDOWN rvrARGrN (SDM)... 3.1-1 3.1 .2 Reactivity Anomalies 3.1 -5 3.1 .3 Control Rod OPERABILITY.... 3.1 -7 3.1 .4 Control Rod Scram Times 3.1-12 3.1 .5 Control Rod Scram Accumulators. 3.1 -1 5 3.1 .6 Control Flod Pattern.. 3.1-18 3.1 .7 Standby Liquid Control (SLC) System 3.1 -20 3.1 .8 Scram Discharge Volume (SDV) Vent and Drain Valves 3.1-24 3.2 POWER DISTHIBUTION LITVIITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATTON RATE (APLHGR) . ....3.2-1 3.2.2 rurNr[/uru cHrTrcAL PowER RATro (McpR) ....3.2-2 3.2.3 LTNEAR HEAT GENERATTON RATE (LHGR) .....3.2-3 3.3 INSTFIU]\4ENTATION 3.3.1 .1 Reactor Protection System (RPS) lnstrumentation 3.3- 1 3.3.1 .2 Source Range Monitor (SRlvl) lnstrumentation 3.3-10 3.3.2.1 Control Rod Block lnstrumentation 3.3-15 3.3.3.1 Post Accident Monitoring (PAtVl) Instrumentation 3.3-20 3.3.3.2 Remote Shutdown System 3.3-24 3.3.4.'1 End of Cycle Hecirculation Pump Trip (EOC-RPT) lnstrumentation..... .3.3-26 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) lnstrumentation .. 3.3-29 3.3.5.1 Emergency Core Cooling System (ECCS) lnstrumentation. . 3.3-32 3.3.5.2 Fleactor Pressure Vessel (RPV) Water Inventory Control lnstru mentation 3.3-43a 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System lnstrumentation . 3.3-44 3.3.6.1 Primary Containment lsolation Instrumentation... . 3.3-48 Residual Heat Removal (RHR) Containment Spray AAAA J.J.O.l System lnstrumentation... . 3.3-60 (continued) PERRY _ UNIT 1 Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY CLEAN, RETYPED PAGE TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6.3 Suppression Pool Makeup (SPMU) System lnstrumentation 3.3-64 3.3.6.4 Relief and Low-Low Set (LLS) lnstrumentation s.3-68 3.3.7.1 Control Room Emergency Recirculation (CRER) System Instrumentation. 3.3-70 3.3.8.1 Loss of Power (LOP) lnstrumentation. 3.3-74 3.3.8.2 Reactor Protection System (RPS) Electric Power ltflonitoring.". .. ..3.3-77 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating. . 3.4-1 3.4.2 Flow Control Valves (FCVs). 3.4-6 3.4.3 Jet Pumps..... 3.4-8 3.4.4 Safety/Relief Valves (S/RVs) 3.4-10 3.4.5 RCS Operational LEAKAGE, . .....3.4-12 3.4.6 RCS Pressure Isolation Valve (PlV) Leakage 3.4-14 3.4.7 RCS Leakage Detection lnstrumentation 3.4-16 3.4.8 RCS Specific Activity ..3.4-19 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown. 3.4-21 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown..... 3.4-24 3.4.11 RCS Pressure and Temperature (Pff) Limits 3.4-26 3,4.12 Reactor Steam Dome Pressure 3.4-32 3.5 EMERGENCY CORE COOLING SYSTEI/S (ECCS), RPVWATER INVENTORY CONTROL, AND REACTOR CORE ISOLATlON COOLING (RCIC) SYSTEM 3.5.1 ECCS-Operating.... 3.5-1 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control 3.5-6 3.5.3 RCIC System 3.5-10 3.6 CONTAINMENT SYSTEMS 3.6.1 .1 Primary Containment - Operating. ... 3.6-1 3.6.1.2 Primary Containment Air Locks 3.6-3 3.6.1.3 Primary Containment lsolation Valves (PCIVs).. 3.6-9 3.6.1 .4 Primary Containment Pressure ..... 3.6-20 3.6.1.5 Primary Containment Air Temperature. 3.6-21 3.6.1.6 Low-Low Set (LLS) Valves. 3.6-22 3.6.1.7 Residual Heat Removal (RHR) Containment Spray System 3.6-24 3.6.1.8 Feedwater Leakage Control System (FWLCS) 3.6-26 3.6.1.9 Main Steam lsolation Valve (lVlSlV) Leakage Control System (LCS) 3.6-27 3.6.1.10 Primary Containment - Shutdown... .. 3.6-29 3.6.1.1 1 Containment Vacuum Breakers. 3.6-31 3.6.1.12 Containment Humidity Control 3.6-34 3.6.2.1 Suppression Pool Average Temperature. 3.6-36 (continued) PERRY - UNIT 1 ii Amendment No

TECHNICAL SPECIFICATION INFORMATION ONLY Def initions

                                                                                          't CLEAN, RETYPED PAGE                                                                           .1 1.1 Definitions (continued)

DFIAIN TITUE The DRAIN TITUE 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 HPV 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 f low rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e,g., seismic event, loss of normal power, single human error), for all penetration flow paths below the TAF except; 1 . Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;

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

3. Penetration flow paths with isolation devices that can be closed prior to the HPV water leve! 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 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; (continued) PERRY _ UNIT 1 1.0-2a Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY Definitions GLEAN, RETYPED PAGE 1.1 1.1 Definitions DRAIN TIME d) No additional draining events occur; and (continued) e) Realistic cross-sectional areas and drain rates are used. A bounding DRAIN TIN/E may be used in lieu of a calculated value. (continued) PERRY _ UNIT 1 1.0-2b Amendrnent No

ECCS Instrumntat i on TECHNICAL SPEGIFICATION 3.3.5.1 PROVIDED FOR CONTEXT 3- 3 IHSTRUI{EI{TATIOH 3.3.5.1 fnergency Core [ooling System (ECCS] Instrumentation LCo 3.3.5. I The ECCS instrumentation for erch Function in Tehle 3.3,5.I-l shall be 0PERABLE. APPLICAEILITY: According to Table 3,3.5.1-1. RITIOHS TOI{DIT IOH REQTJIRtD ACTIoH COHPLETIOI{ TIHE A. One or more channels A.t fnter the [ondition Immedi ately i noperabl e. l-eferenced in Tabtre 3.3.5.1-l for t,he channel . {conti nued} PERRY UI{IT I 3.3-32 Amendment tlo. 69

TECH NICAL SPECI FIGATION INFORMATION ONLY ECCS lnstrumentation CLEAN, RETYPED PAGE 3.3.5.1 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME B. As required by Required 8.1 Action A.1 and Only applicable for referenced in Functions 1.a, 1.b, 2.a Table 3.3.5. 1-1 . and 2.b. Declare supported t hour from discovery feature(s) inoperable when of loss of initiation its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions AND 8.2 Only applicable for Functions 3.a and 3.b. Declare High Pressure t hour from discovery Core Spray (HPCS) of Ioss of HPCS System inoperable. initiation capability AND 8.3 PIace channel in trip. 24 hours (continued) PERRY - UNIT 1 3.3-33 Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY ECCS Instrumentation CLEAN, RETYPED PAGE 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by Required c.1 Action A.1 and Only applicable for referenced in Functions 1.c, 1.d, 1.e, 2.c, Table 3.3.5.1-1. and 2.d. Declare supported t hour from discovery feature(s) inoperable when of loss of initiation its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions AND c.2 Restore channel to 24 hours OPERABLE status. (continued) PERRY - UNIT 1 3.3-34 Amendment No

TECH NICAL SPECI FICATION INFORMATION ONLY ECCS Instrumentation CLEAN, RETYPED PAGE 3.3.5.1 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 --NOTE. Action A.1 and Only applicable for referenced in Functions 1.f, 1.9, and 2.e. Table 3.3.5.1-1. Declare supported t hour from discovery feature(s) inoperable when of loss of initiation its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions AND E.2 Restore channel to 7 days OPERABLE status. F. As required by Required F.1 Declare Automatic t hour from discovery Action A.1 and Depressurization System of Ioss of ADS referenced in (ADS) valves inoperable. initiation capability in Table 3.3.5. 1-1 . both trip systems AND F.2 PIace channel in trip 96 hours from discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable AND 8 days (continued) PERRY - UNIT 1 3.3-36 Amendment No.

TECHNIGAL SPEGIFICATION INFORMATION ONLY ECCS Instrumentation GLEAN, RETYPED PAGE 3.3.5.1 Table 3.3.5.1-1 (page 1 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIFIED REFERENCED OTHEH CHANNELS FROM SPECIFlED PER REQUIRED SUFIVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Low Pressure Coolant lnjection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems

a. Reactor Vessel Water 1,2,3 2(b) B sR 3.s.5.1.1 > 14.3 inches Level - Low Low Low, sR 3.3.5.1.2 Level 1 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

b. Drywell Pressure - High 1 ,2,3 2(b) B sH 3.3.s.1.1 < 1.88 psig sR 3.3.5.1.2 sR 3.3.5.1.3 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 <6rtr Start - Time Delay sFr 3.3.5.1.4 seconds Relay sR 3.3.5.1.6

d. Reactor Vessel 1,2,3 C sR 3.3.5.1.1 z 482.7 psig Pressure - Low (LPCS sFl 3.3.5.1.2 and lnjection Valve sR 3.3.5.1.3 < 607.7 psig Permissive) sR 3.3.5.1.5 sH 3.3.5.1.6

e. Reactor Vessel 1,2,3 C sR 3.3.5.1.1 > 490.0 psig Pressure - Low (LPCI sR 3.3.5.1.2 and lnjection Valve sR 3.3.5.1.3 < 537.1 psig Permissive) sFt 3.3.5.1.5 sR 3.3.5.1.6

f. LPCS Pump Discharge 1,2,3 E sR 3.3.5.1.1 > 1 200 gpm Flow - Low (Bypass) sH 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6 contin (b) Also required to initiate the associated diesel generator and AEGT subsystem PERRY _ UNIT 1 3.3-39 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY ECCS lnstrumentation CLEAN, RETYPED PAGE 3.3.5.1 Table 3.3.5.1-1 (page 2 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION COND!TIONS FUNCTION ACTION 4.1 REQUIREMENTS VALUE 1, Low Pressure Coolant lnjection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems (continued)

g. LPCI Pump A Discharge 1,2, 3 1 E sR 3.3.s.1.1 > 1450 gpm Flow - Low (Bypass) sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.s sR 3.3.5.1.6 h, Manual lnitiation 1,2, 3 C sR 3.3.5.1.6 NA

2. LPCI B and LPCI C Subsystems

a. Reactor VesselWater 1,2,3 2(b) B SR 3.3.5.1 .1 > 14.3 inches Level - Low Low Low, SR 3.3.5.1 ,2 Level 1 SR 3.3.5.1 .3 SR 3.3.5.'l .5 sR 3.3.s.1.6

b. Drywell Pressure - High 1,2,3 2(b) B sR 3.3.5.1.1 < 1.88 psig sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

c. LPCI Pump B Start - Time 1,2,3 C sR 3.3.5.1.2 < 5.25 Delay Relay sR 3.3.5.1.4 seconds sR 3.3.5.1.6

d. Reactor Vessel 1,2,3 '1 per C sR 3.3.5.1.1 > 490.0 psig Pressure - Low (LPCI subsystem sR 3.3.5.1.2 and lnjection Valve sR 3.3.5.1.3 s 537.1 psig Permissive) sR 3.3.5.1.5 for LPCI B; sR 3.3.5.1.6 and

                                                                                                          > 490.0 psig and
                                                                                                          ='537.1 psig for LPCI C (continued)

(b) Also required to initiate the associated diesel generator and AEGT subsystem PERRY - UNIT 1 3.3-40 Amendment No

TECHNIGAL SPEGIFICATION INFORMATION ONLY ECCS lnstrumentation CLEAN, RETYPED PAGE 3.3.5.1 Table 3.3.5.1-1 (page 3 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTlON CONDITIONS FUNCTION ACTION 4.1 REQUIREMENTS VALUE

2. LPCI B and LPCI C S ubsystems (continued)

e. LPCI Pump B and LPCI 1,2,3 l perpump E sR 3.3.5.1.1 > 1450 gpm Pump C Discharge sR 3.3.s.1.2 Flow - Low (Bypass) sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

f. Manual Initiation 1,2,3 1 C sR 3.3,5,1.6 NA

3. High Pressure Core Spray (HPCS) System

a. Reactor Vessel Water 1,2, 3 4(e) B sR 3.3.s.1.1 > 127.6 inches Level - Low Low, sR 3.3.5.1.2 Level 2 sR 3.3.5.1.3 sR 3.3.5.'1,5 sR 3.3.5.1.6

b. Drywell Pressure - High 1,2, 3 4(e) B sR 3.3.5.1.1 s 1.88 psig sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

c. Reactor VesselWater 1, 2,3 4 B sR 3.3.5.1.1 s221.7 inches Level- High, Level I sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

d. Condensate Storage Tank 1,2,3 2 D sR 3.3.5.1.1 > 90,300 Level - Low sR 3.3.5.1.2 gallons sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6

e. Suppression PoolWater 1,2,3 2 D sR 3.3.5.1.1 <18ft6inches Level- High sR 3.3.s.1.2 sR 3.3.5.1.3 sR 3.3.5.1.7 sR 3.3.5.1.6 (continued)

(e) Also required to initiate the associated diesel generator. PERRY _ UNIT 1 3.3-41 Amendment No

TECHNICAL SPECIFICATION INFORMATION ONLY ECCS lnstrumentation CLEAN, RETYPED PAGE 3.3.5.1 Table 3.3.5.1-1 (page 4 of 5) Emergency Core Cooling System lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFEHENCED OTHER CHANNELS FROM SPECIFIED PER REQUIHED SUFIVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

3. High Pressure Core Spray (HPCS) System (continued)

f. HPCS Pump Discharge 1,2,3 E sR 3.3.5.1.1 > 120 psig Pressure - High (Bypass) sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.3.5.1.5 sR 3.3.5.1.6 g" HPCS System Flow 1,2,3 E SR 3.3.5.1.1 > 600 gpm Rate - Low (Bypass) SR 3.3.5.1.2 SR 3.3.5.1.3 SH 3.3.5.1.5 SR 3.3.5.1.6

h. Itlanual lnitiation 1,2,3 C SR 3.3.5.1.6 NA

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor VesselWater rItL c(d),J r(d) 2 F sFr 3.3.5.1.1 > 14.3 inches Level - Low Low Low, sR 3.3.5.1.2 Level 1 sH 3.3.5.1.3 sH 3.3.5.1.5 sR 3.3.5.1.6

b. ADS lnitiation Timer 1, 2(d), 3(d) G sR 3.3.5.1.2 > 100.5 seconds sFl 3.3.5.1.4 and sR 3.3.5.1.6 < 109.5 seconds

c. Reactor VesselWater 1, 2(d), 3(d) F sR 3.3.5.1.1 > 177.1 inches Level - Low, Level 3 sR 3.3.5.1"2 (Confirmatory) sR 3.3.5.1.3 sR 3.3.5.1.5 sH 3.3.5.1.6

d. LPCS Pump Discharge 1, 2(d), 3(d) 2 tl sR 3.3.5.1.1 > 125 psig Pressure - High sR 3.3.5.1.2 sH 3.3.5.1.3 sB 3.3.5.1.5 sH 3.3.5.1.6

e. LPCI Pump A Discharge 1, 2(d), 3(d) 2 G sR 3.3.5.1.1 > 115 psig Pressure - High sR 3.3.5.1.2 sR 3.3.5.1.3 sR 3.s.5.1.5 sR 3.3.5.'t.6

f. Manual lnitiation 1, 2(d), 3(d) 2 G sFr 3.3.5_1.6 NA (continued)

(d) With reactor steam dome pressure > 150 psig PERHY _ UNIT 1 3.3-42 Amendment No.

TECHNIGAL S PECI FICATION INFORMATION ONLY RPV Water lnventory Control lnstrumentation CLEAN, RETYPED PAGE 3.3.5.2 3.3 INSTRU[/ENTATION 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control lnstrumentation LCO 3.3.5.2 The RPV Water lnventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.5.2-1 . ACTIONS

                                             ----NOTE Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TINTE A. One or more channels A.1 Enter the Condition !mmediately inoperable. referenced in Table 3.3.5.2-1 for the channel. B. As required by Required 8.1 Declare associated lmmediately Action A.1 and penetration f low path(s) referenced in incapable of automatic Table 3.3.5.2-1 . isolation. AND 8.2 Calculate DRAIN TltVlE. lmmediately C. As required by Required C.1 Place channel in trip. t hour Action A.1 and referenced in Table 3.3.5.2-1 . (continued) PERRY _ UNIT 1 3.3-43a Amendment No.

TECHNIGAL SPEGIFICATION INFORMATION ONLY RPV Water Inventory Control lnstrumentation CLEAN, RETYPED PAGE 3.3.5.2 ACTIONS continued CONDITION HEQUIRED ACTION COIUPLETION TIIUE D. As required by Required D.1 Declare HPCS system t hour Action A.'l and inoperable. referenced in Table 3.3.5.2-1. OR D.2 Align the HPCS pump t hour suction to the suppression pool. E. As required by Required E.1 Restore channel to 24 hours Action A.1 and OPERABLE status. referenced in Table 3.3.5.2-1. F. Required Action and F.1 Declare associated ECCS lmmediately associated Completion i nj ectionispray subsystem Time of Condition C, D, inoperable. E, or F not met. PERRY _ UNIT 1 3.3-43b Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY RPV Water lnventory Control lnstrumentation CLEAN, RETYPED PAGE 3.3.5.2 SU RVEI LLANCE REQU IREMENTS

                                            ---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 Prograrn SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. ln accordance with the Surveillance Frequency Control Program SR 3.3.5.2.3 Perform LOGIC SYSTEIU FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program PERRY _ UNIT 1 3.3-43c Amendment No.

TECHNIGAL S PECIFIGATION INFORMATION ONLY RPV Water lnventory Control Instrumentation CLEAN, RETYPED PAGE 3.3.5.2 Table 3.3.5.2-1 (page 1 of 2) RPV Water lnventory Control lnstrumentation APPLlCABLE CONDIT!ONS IVIODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Low Pressure Coolant lnjection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems

a. Reactor Vessel 4,5 1(a) C sR 3.3.5.2.1 > 482.7 psig Pressure - Low sR 3.3.5.2.2 and (LPCS lnjection Valve < 607.7 psig Permissive)

b. LPCS Pump Discharge 4,5 1 (a) E sR 3.3.5.2.1 2 1200 gpm Flow - Low (Bypass) sR 3.3.5.2.2

c. Fleactor Vessel 4,5 1(a) C sR 3.3.5.2.1 2 490.0 psig Pressure - Low sR 3.s.5.2.2 and (LPCI lnjection Valve < 537.1 psig Permissive)

d. LPCI Pump A Discharge 4,5 1 (a) E sR 3.3.5.2.1 2 1450 gpm Flow - Low (Bypass) sR 3.3.5.2.2

e. Manual Initiation 4,5 1(a) E sR 3.3.5.2.3 NA

2. LPCI B and LPCI C Subsystems

a. Reactor Vessel 4,5 1 per C sR 3.3.5.2.1 > 490.0 psig Pressure - Low (LPCI subsystem sR 3.3.5.2.2 and lnjection Valve (a) 5 537.1 psig Permissive) for LPCI B; and

                                                                                                    > 490.0 psig and
                                                                                                    < 537.1 psig for LPCI C

b. LPCI Pump B and LPCI 4,5 1 per pump E sR 3.3.5.2.1 l 1450 gpm Pump C Discharge (a) sR 3.3.5.2.2 Flow - Low (Bypass)

c. [Vlanual lnitiation 4,5 1 (a)

E sR 3.3.5.2.3 NA (continued) (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "Reactor Pressure Vessel (RPV) Water lnventory Control." PERRY _ UNIT 1 3.3-43d Amendment No.

TECHNICAL SPEGIFIGATION INFORMATION ONLY RPV Water lnventory Control lnstrumentation CLEAN, RETYPED PAGE 3.3.5.2 Table 3.3.5.2-1 (page 2 of 2) RPV Water lnventory Control lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

3. High Pressure Core Spray (HPCS) System

a. Condensate Storage Tank a(b) c(b) 2@\ D sR 3.3.5.2.1 > 90,300 Level - Low sR 3.3.5.2.2 gallons

b. HPCS Pump Discharge 4,5 1 (a)

E sR 3.3.5.2.1 > 120 psig Pressure - High (Bypass) sH 3.3.5.2.2

c. HPCS System Flow 4,5 1 (a)

E sR 3.3.5.2.1 2 600 gpm Rate - Low (Bypass) sR 3.3.5.2.2

4. RHR System lsolation

a. Reactor Vessel Water (c) 2 in one trip B sR 3.3.5.2.1 > 177.1 inches Level - Low, Level 3 system sR 3.3.5.2.2

5. Reactor Water Cleanup (RWCU) System lsolation

a. Reactor Vessel Water (c) 2 in one trip B sR 3.3.5.2.1 > 127.6 inches Level - Low Low, Level 2 system sR 3.3.5.2.2 (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "Reactor Pressure Vessel (RPV) Water lnventory Control."

(b) When HPCS is OPERABLE for compliance with LCO 3.5.2, "Reactor Pressure Vessel (RPV) Water lnventory Control," and aligned to the condensate storage tank. (c) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME. PERRY _ UNIT 1 3.3-43e Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY RCIC System lnstrumentation CLEAN, RETYPED PAGE 3.3.5.3 3.3 INSTRUMENTATION 3.3.5.3 Reactor Core lsolation 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, IVIODES 2 and 3 with reactor steam dome pressure > 150 psig. ACTIONS Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels 4.1 Enter the Condition lmmediately inoperable. referenced in Table 3.3.5.3-1 for the channel. B. As required by Required 8.1 Declare RCIC System t hour from discovery Action A.1 and inoperable. of loss of RCIC referenced in initiation capability Table 3.3.5.3-1. AND 8.2 Place channel in trip. 24 hours C. As required by Required c.1 Restore channel to 24 hours Action A.1 and OPERABLE status. referenced in Table 3.3.5.3-1 . (continued) PERRY - UNIT 1 3.3-44 Amendment No.

TEC H NICAL SPECI FICATION INFORMATION ONLY CLEAN, RETYPED PAGE RCIC System lnstrumentation 3.3.5.3 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 ...NOTE Action A.1 and Only applicable if RCIC referenced in pump suction is not aligned Table 3.3.5.3-1 . to the suppression pool. Declare RCIC System t hour from discovery inoperable. of loss of RCIC initiation capability AND D.2.1 PIace channel in trip. 24 hours OR D.2.2 AIign RCIC pump suction 24 hours to the suppression pool. E. Required Action and E.1 Declare RCIC System lmmediately associated Completion inoperable. Time of Condition B, C, or D not met. PERRY - UNIT 1 3.3-45 Amendment No

TECHNICAL SPECIFICATION INFORMATION ONLY RCIC System Instrumentation CLEAN, RETYPED PAGE 3.3.5.3 SU RVEILLANCE REQUIREMENTS

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

2 When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 5; and (b) for up to 6 hours for Functions 1, 2, 3, and 4 provided the associated Function maintains RCIC initiation capability. SURVEILLANCE FREQUENCY SR 3.3.5.3.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control P ram SR 3.3.5.3.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control P ram SR 3.3.5.3.3 Calibrate the trip unit. In accordance with the Surveillance Frequency Control P ram SR 3.3.5.3.4 Perform CHANNEL CALIBRATION. ln accordance with the Surveillance Frequency Control P ram SR 3.3.5.3.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control P ram SR 3.3.5.3.6 Perform CHANNEL CALIBRATION. ln accordance with the Surveillance Frequency Control Program PERRY - UNIT 1 3.3-46 Amendment No

TECH NICAL SPECI FICATION INFORMATION ONLY RCIC System Instrumentation CLEAN, RETYPED PAGE 3.3.5.3 Table 3.3.5.3-1 (page 1 of 1) Reactor Core lsolation Cooling System lnstrumentation CONDITIONS REQUIRED REFERENCED CHANNELS FROM PER REQUIRED SURVEILLANCE FUNCTION FUNCTION ACTION A.1 REQUIREMENTS ALLOWABLE VALUE

1. Reactor Vessel Water Level - 4 B sR 3.3.5.3.1 Z 127.6 inches Low Low, Level 2 sR 3.3.s.3.2 sR 3.3.5.3.3 sR 3.3.5.3.4 sR 3.3.s.3.s

2. Reactor Vessel Water Level - 4 C sR 3.3.5.3.1 <221.7 inches High, Leve! I sR 3.3.5.3.2 sR 3.3.5.3.3 sR 3.3.5.3.4 sR 3.3.s.3.5

3. Condensate Storage Tank 2 D sR 3.3.5.3.1 > 90,300 gallons Level - Low sR 3.3.5.3.2 sR 3.3.5.3.3 sR 3.3.5.3.4 sR 3.3.5.3.5

4. Suppression Pool Water 2 D sR 3.3.5.3.1 S18ft6inches Level- High sR 3.3.s.3.2 sR 3.3.5.3,3 sR 3.3.5,3.6 sR 3.3.s.3.5

5. Manual lnitiation 1 C sR 3.3.s.3.5 NA PERRY - UNIT 1 3.3-47 Amendment No.

TECH NICAL SPECI FICATION Primary Contalnment and 0ryuel I Iso'l rtt on InstrumT:l:l:l PROVIDED FOR CONTEXT 3,3 II{5TRUI{ET{TAIIOI{ 3,3 .E.l Primary Contrinnent and Dryrrell Isolation Instrurrentatlon LCo 3.3.6.1 The primary contrinment rnd dryrell lsolation instrumentrtion for each Function in Tabls 3.3"S.1*l shall be 0PIBIIBLE. APPLICfiBILITY: According to lable 3.3.6.1-1. ACTI0I{S EOHI}ITION nEqulEtD AcTIor{ IOHPLTTTOI{ T IHE A. 0ne or more required A.I PIace channel in te hsurs fot" channels inoperrble. t,ri p. Functions 2.b, 5.b, and 5-d AEE e4 hours for Functtons other than Functions A.b, 5.b, and 5.d E 0ne or morg eutomatic B.l Restore isolation I hour' Functions nith capabi 1 i ty. isolatisn capabil ity not maintatned. (conti nued) PERRY UI{IT I 3.3-{8 Amendment [tlo. 69

TECHNIGAL SPEGI FIGATION INFORMATION ONLY Primary Containment and Drywell lsolation Instrumentation CLEAN, RETYPED PAGE 3.3.6.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME J. As required by Required J.1 lnitiate action to restore Immediately Action C.1 and channel to OPERABLE referenced in status. Table 3.3.6.1-1. (continued) PERRY _ UNIT 1 3.3-51 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY Primary Containment and Drywell Isolation lnstrumentation CLEAN, RETYPED PAGE 3.3.6.1 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME K. As required by Required K.1 Isolate the affected lmmediately Action C.1 and penetration flow path(s). referenced in Table 3.3.6.1-1. OR K.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment. PERRY _ UNIT 1 3.3-52 Amendment No.

Pri mary Contai nment and Dryrnrel I IsoI at i on I nstr umentati on 3.3.6.1 TEC HNICAL S PECI FICATION PROVIDED FOR CONTEXT Tabl,e 5.5.6.1-1 (page 1 of 6) Prinnry Containrnnt ard llrptet t- lsotati m In$trunentrti on APPLItABLE c${Dr TIo}ls ilmEs of, NEqUIRED REFEREIICED OTHER CHAII}IELS FROI{ SPECI FIED PER IRIP EEOUIREO SURVEI I.IAIICE ALLOIJABLE FUHETIOI{ EOIIDITIOI{S SYSTEH ACTTO!{ C.l RE([JINTIG}ITS UNLUE

t. l{ain Steam Line Isolation a Reactor Vessel Hster 1,2,5 e D sR 3.5.6 ,I z 14,3 inches Level.-Lor Lor Loxn sR 5.5.6 ,z Level I sR 5.5.6 .5 sf, 5.I 6 .4

                                                                                            $R 5.L6 .5 sR 5.5.6 .6 b- l{ain Stearn Line                         1               z            E          sR I.3.6.1       I z 7T5.? psf g Pressure - LoH                                                                  sR 3.3,6.1       2 sR 313.6.1       3 sE 3.3.6,1       4 sR 3.5,6,1       5 sR 3.3.6.1       6

c. l,lain Steam Line 1,2rI 2 per ttSL D sR 3-3.6.1 .t s 256.5 psfd Ftou - ilfsh sR 5,3.6. t .2, sR 5-5,6.1 .5 sR 3,3.6.1 .4E s* 3.3.6.1 sR 5.5.6.1 .6

d. gondenser V8cut$r - Lor{

I .z(al, ? D SR 5.r.6.1 . t a 7.6 inchee SR 3.5.6.1.a Hg vacurm SR 3.3,6.I .3 5(a) SR r.L6.I.4 SR 3.3.6. r .5 e Hain tffln ttne Fipe ? 5 ? D sR 3.3.6.1 . t g 158,90F TunneI Tenperature-- sR 3.5.6. t .a Hish sR 3.5.6.1.4

                                                                                            $R 3.5.6.1 .5 sR 3.3.6.t.7
      'f . l,{ain Steam    L ine                t r2.3            2             D          SF   3.3.6.1 . I   <  149,6"F Iurbine     Bui I ding                                                          $R   5.5.6. t.E Ietlprature- Hish                                                              SR   3,3.6. r.4 SR   3.3.6. r.5

g. l,lanual Ini tf atf on I ,Zrt ?. G sR 3.I.6.1.5 HA

2. Primary ContairuDent and DrprelI Isol,at ion

a. Reactor Vessel tJater 'l,ErJ z

(b) I SR 5 .5 6 .1 > 1a7.6 inches Level - Lon Lon, l-evel ? SR 3 .3 6 .z sI 5 .5 6 .3 sft 3 "3 6 .4 SR 5 ,3 6 .5 (cont i nued) (a) lJith any turbine stop vatve not ctosed. (b) Required to initiate the associated dryilelI isotatfon function. PIRRY - UNIT i 3.3-s4 Amendment No. 130

TECHNICAL SPECIFICATION INFORMATION ONLY Primary Containment and Drywell lsolation Instrumentation CLEAN, RETYPED PAGE 3.3.6.1 Table 3.3.6.1-1 (page 2 of 6) Primary Containment and Drywell lsolation lnstrumentation APPLlCABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVElLLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

2. Primary Containment and Drywell lsolation

b. Drywell Pressure - High 1,2,3 2(b) H sR 3.3_6.1.1 s 1.88 psig sR 3_3_6.1_2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

c. Reactor VesselWater '1, 2, 3 2(b) F sR 3.3.6.1.1 > 14.3 inches Level - Low Low Low, sR 3.3_6.1.2 Level 1 (ECCS Divisions sR 3.3.6.1.3 1 and 2) sR 3.3.6.1.4 sFl 3.3.6.1.s

d. Drywell Pressure - High 1, 2,3 2 F sR 3.3.6.1.1 < 1.88 psig (ECCS Divisions 1 and 2) sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

e. Reactor VesselWater 1,2,3 4 F sB 3.3.6.1.1 >- 127.6 inches Level - Low Low, Level 2 sH 3.3.6.1.2 (HPCS) sH 3.3.6.1.3 sH 3.3.6.1.4 sH 3.3.6.1.5

f. Drywell Pressure - High 1 ,2,3 4 F sR 3.3.6.1.1 s 1.88 psig (HPCS) sH 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.s

g. Containment and Drywelt 1,2,3 ,.,(b) 4.',

F sR 3.3.6.1.1 < 4.0 mR/hr Purge Exhaust Plenum sR 3.3.6.1.2 above Radiation - High sR 3.3.6.1.4 background sR 3.3.6.1.5 (continued) (b) Required to initiate the drywell isolation function PERRY _ UNIT 1 3.3-55 Amendment No.

TECHNIGAL SPECIFICATION INFORMATION ONLY Primary Containment and Drywell lsolation lnstrumentation CLEAN, RETYPED PAGE 3.3.6.1 Table 3.3.6.1-1 (page 3 of 6) Primary Containment and Drywell lsolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

2. Primary Containment and Drywell lsolation

g. Containment and Drywell (d) 2 K sR 3.3.6.'1.1 < 4.0 mR/hr Purge Exhaust Plenum sR 3.3.6.1.2 above Radiation - High sR 3.3.6.'1.4 background (continued) sR 3.3.6.1.5

h. Manual Initiation 1,2, 3 2(b) t sR 3.3.6.1.5 NA (d) 2 K sR 3.3.6.1.5 NA

3. Reactor Core lsolation Cooling (RCIC) System lsolation

a. RCIC Steam Line Flow - 1,2,3 F sR 3.3.6.1.1 s 298.5 inches High sR 3.3.6.1.2 water sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

b. RCIC Steam Line Flow 1,2,3 F sR 3.3.6.1.2 > 3 seconds and Time Delay sR 3.3.6.1.4 < 13 seconds sR 3.3.6.1.5

c. RCIC Steam Supply Line 1,2,3 F sR 3,3.6.1.1 > 55 psig Pressure - Low sR 3.3.6.1.2 sR 3.3,6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

d. RCIC Turbine Exhaust 1,2,3 2 F sR 3.3.6.1.1 s 20 psig Diaphragm Pressure - sR 3.3.6.1.2 High sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5

e. RCIC Equipment Area 1,2,3 F sR 3.3.6.1.1 < 145.9'F Ambient Temperature - sR 3.3.6.1.4 High sR 3.3.6.1.s sR 3.3.6.1.7

f. Main Steam Line Pipe 1,2,3 F sR 3.3.6.1.1 s 158.9"F TunnelTemperature - sR 3.3.6.1.4 High sR 3.3.6.1.5 sR 3.3.6.1.7 (continued)

(b) Required to initiate the drywell isolation function. (d) During movement of recently irradiated fuel assemblies in primary containment. PERRY - UNIT 1 3.3-56 Amendment No.

TEGHNICAL SPECIFICATION INFORMATION ONLY Primary Containment and Drywell lsolation lnstrumentation CLEAN, RETYPED PAGE 3.3.6.1 Table 3.3.6.1-1 (page 6 of 6) Primary Containment and Drywell lsolation lnstrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REOUIREMENTS VALUE

5. RHR System lsolation

a. RHR Equipment Area 2(e), 3(e) 1 per area F sR 3.3.6.1.1 < 159.9"F Ambient sR 3.3.6.1.4 Temperature - High sR 3.3.6.1.5 sR 3.3.6.1.7

b. Reactor VesselWater r r(g) t(g)

It , V 2 F sR 3.3.6.1.1 > 177.1 inches Level- Low, Level3 sR 3.3.6.1.2 sR s.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.5 r(e) r(e) 2 J sR 3.3.6.1.1 > 177.1 inches sR 3.3.6.1.2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.s

c. Reactor Vessel Steam 1,2, 3 2 F sR 3.3.6.1.1 < '150 psig Dome Pressure - High sR 3.3.6.1"2 sR 3.3.6.1.3 sR 3.3.6.1.4 sR 3.3.6.1.s

d. Drywell Pressure - High 1,2,3 2 F sR 3.3.6.1.1 < 1.88 psig sR 3.3"6.1.2 sR 3.3.6.1.3 sR 3.3.6.1,4 sR 3.3.6.1.5

e. Manual lnitiation 1,2,3 2 G sR 3.3.6.1.5 NA (e) With reactor vessel steam dome pressure less than the RHR cut in permissive pressure.

(g) With reactor vessel steam dome pressure greater than or equal to the RHR cut in permissive pressure PERRY - UNIT 1 3.3-59 Amendment No

CREil System Instrumentation TEC HNIGAL SPECIFICATION 3.3"?.1 PROVIDED FOR CONTEXT 3.3 I HSTRUHEI{TAT IOl{ 3.3 ,I.l Control Room Emergency Eecirculatlon (CRER) System Instrumentation LCo 3.3.7,1 The CRIR System inctrunentation for each Functlon tn I Tab'le 3.3.7.1-l shal be OPEBABLE. APPLICABILITY: According to Table 3.3.7,1-1. ACT IOHS TOI{OI T ION REQUIRED ACTIOH TDHPLETI0H TIHE A One or more channel s A.l Enter the Condition I mmed i atel y i noperabl e. refarenqed in Tahle 3.3.7.1-l for the channel. B As required by B" I Declare as oEiated I hour from Bequired Action n.l CHEil subsystem discovery of and referenced in inoperabl e. lsss of CRER Table 3"3.7.1*1. init.iation capabil ity in bsth trip systems AUE B.? PIace ehannel in ?4 hours tri p. (conti nued) PERRY UHIT I 3.3-70 Amendment l{0. 69

TECH NICAL SPECI FIGATION INFORMATION ONLY CRER System lnstrumentation CLEAN, RETYPED PAGE 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1) Control Floom Emergency Recirculation System lnstrumentation APPLICABLE CONDITIONS MODES OR REOUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PEH THIP HEQUlRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A-1 REQUIFIEMENTS VALUE

1. Reactor Vessel Water 1,2,3 2 B sR 3.3.7.1.1 > 14.3 inches Level - Low Low Low, sR 3.3.7.1.2 Level 1 sR 3.3.7.1.3 sR 3.3.7.1.4 sR 3.3.7_1.5

2. Drywell Pressure - High 1 ,2,3 2 B sR 3,3.7.1.1 < 1.88 psig sR 3.3.7.1.2

                                                                                      $H 3.3.7.1.3 sR 3.3.7.1.4 sR 3.3.7.1.5

3. Control Room Ventilation 1,2,3, c sR 3.3.7.1.1 < 800 cpm Radiation Monitor (b) sH 3.3.7.1.2 sR 3.3.7.1.4 sR 3.3.7.1.5 (b) During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building PERRY _ UNIT 1 3.3-73 Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY CLEAN, RETYPED PAGE ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTETVTS (ECCS), RPV WATER INVENTORY CoNTROL, AND REACTOR CORE TSOLATION COOLTNG (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of eight safety/relief valves shall be OPERABLE. APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure < 150 psig. ACTIONS NOT LCO 3.0.4.b is not applicable to HPCS. CONDITION REQUIRED ACTION COMIPLETION TIME A. One low pressure ECCS A,1 Restore low pressure 7 days i njection/spray subsystem ECCS injection/spray inoperable. subsystem to OPERABLE status. B. High Pressure Core B.1 Verify by administrative t hour Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC is required to be OPERABLE. AND 8.2 Restore HPCS System to 14 days OPERABLE status. (continued) PERRY - UNIT 1 3.5-1 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY RPV Water lnventory Control CLEAN, RETYPED PAGE 3.5.2 3.5 EMERGENCY CORE COOLING SYSTETVIS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE TSOLATTON COOLTNG (RClC) SYSTEM 3.5.2 Reactor Pressure Vessel (RPV) Water lnventory Control LCO 3.5.2 DRAIN TIIIE of RPV water inventory to the top of active fuel (TAF) shall be 2 36 hours. AND One ECCS injection/spray subsystem shal! be OPERABLE. A il;;;il;;ffi il;ffi i.?l=D il;;* OPERABLE during alignment and operation for decay

                                                                              ;;,u-  ;" ;;, J" *d
                                                                           ", heat removal, if capable of being manually realigned and not otherwise inoperable.

APPLICABILITY: MODES 4 and 5. ACTIONS CONDITION REQUIRED ACTION COTUPLETION TI[ilE A. Required ECCS 4.1 Restore required ECCS 4 hours i njection/spray subsystem i njection/spray su bsystem inoperable. to OPERABLE status. B. Required Action and 8.1 lnitiate action to establish a lmmediately associated Completion method of water injection Time of Condition A not capable of operating met. without offsite electrical power. C. DRAIN TIIVIE < 36 hours c.1 Verify primary containment 4 hours and > B hours. boundary is capable of being established in less than the DRAIN TltvlE. AND (continued) PERRY _ UNIT 1 3.5-6 Amendment No.

TECHNICAL SPECI FICATION INFORMATION ONLY RPV Water Inventory Control CLEAN, RETYPED PAGE 3.5.2 ACTIONS continu CONDITION REQUIRED ACTION CO]UPLETION TINIE C.2 Verify each primary 4 hours contai nment penetration flow path is capable of being isolated in less than thE DRAIN TIME. (continued) PEHRY - UNIT 1 3.5-6a Amendment No.

TECH NICAL SPECI FICATION INFORMATION ONLY RPV Water lnventory Control CLEAN, RETYPED PAGE 3.5.2 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME D. DRAIN TltVlE < I hours. D.1 .NOTE-- Required ECCS i njection/spray subsystem or additional method of water injection shall be capable of operating without offsite eleclrical power. Initiate action to establish an lmmediately additional method of water injection with water sources capable of maintaining RPV water level > TAF for

                                 > 36 hours.

AND D.2 lnitiate action to establish Immediately primary containment boundary. AND D.3 lnitiate action to isolate each lmmediately primary containment penetration flow path or verify it can be manually isolated lrorn the control room. E. Required Action and E.1 lnitiate action to restore Immediately associated Completion DRAIN TINilE to > 36 hours. Time of Condition C or D not met. OR DRAINTITUE<1hour. PERHY _ UNIT 1 3.5-7 Amendment No

TEGHNIGAL SPECI FICATION INFORMATION ONLY RPV Water lnventory Contro! CLEAN, RETYPED PAGE 3.5.2 SU RVE I LLANCE REQU I RETUENTS SURVEILLANCE FREQUENCY sR 3.5.2.1 Verify DRAIN TIME > 36 hours. ln accordance with the Surveillance Frequency Control Program sR 3.5.2.2 Verify, for a required low pressure ECCS ln accordance injectionispray subsystem, the suppression pool with the water level is > 16 ft 6 in. Surveillance Frequency Control Program (continued) PERRY _ UNIT 1 3.5-7a Amendment No

TECHNICAL SPECI FICATION INFORMATION ONLY RPV Water Inventory Control CLEAN, RETYPED PAGE 3.5.2 SU HVEILLANCE REQU I REMENTS continu SURVEILLANCE FREQUENCY sR 3.5.2.3 Verify, for a required High Pressure Core Spray ln accordance (HPCS) System, the: with the Surveillance

a. Suppression pool water level is > 16 ft 6 in; or Frequency Control Program

b. Condensate storage tank water volume is

                      > 249,700 gal.

sR 3.5.2.4 Verify, for the required ECCS injection/spray ln accordance subsystem, the piping is f illed with water from the with the pump discharge valve to the injection valve. Surveillance Frequency Control Program sH 3.5.2.5 Verify, for the required ECCS injectionispray ln accordance subsystem, each manual, power operated, and with the automatic valve in the flow path, that is not Iocked, Surveillance sealed, or otherwise secured in position, is in the Frequency correct position. Control Program (continued) PERRY _ UNIT 1 3.5-8 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY RPV Water Inventory Control GLEAN, RETYPED PAGE 3.5.2 SURVEI LLANCE REQU I REMENTS continu SURVEILLANCE FREQUENCY sR 3.5.2.6 Operate the required ECCS injection/spray ln accordance subsystem through the recirculation line for with the

                > 1 0 minutes.                                          Surveillance Frequency Control Program sR 3.5.2.7     Verify each valve credited for automatically isolating  ln accordance a penetration flow path actuates to the isolation       with the position on an actual or simulated isolation signal. Surveillance Frequency Control Program sR 3.5.2.8 Verify the required LPCI or LPCS subsystem              ln accordance actuates on a manual injection signal, or the          with the required HPCS subsystem can be manually                Surveillance operated.                                              Frequency Control Program PERRY  - UNIT 1                            3.5-9                       Amendment No.

TECH NIGAL SPECIFICATION INFORMATION ONLY RCIC System CLEAN, RETYPED PAGE 3.5.3 3.5 ETUERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CoNTROL, AND REACTOR CORE lSOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE. APPLICABILITY: IVIODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig ACTIONS NOTE--- LCO 3.0.4.b is not applicable to RCIC. CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable. 4.1 Verify by administrative t hour means High Pressure Core Spray System is OPERABLE. AND 4.2 Restore RCIC System to 14 days OPERABLE status. B. Required Action and 8.1 Be in IVIODE 3. 12 hours associated Completion Time not met. AND 8.2 Reduce reactor steam 36 hours dome pressure to < 150 psig. PERRY - UNIT 1 3.s-10 Amendment No.

TECHNICAL SPEGIFICATION INFORMATION ONLY Primary Containment Air Locks CLEAN, RETYPED PAGE 3.6.1.2 3.6 CONTAINTVIENT SYSTEMS 3.6.1.2 Primary Containment Air Locks LCO 3.6.1.2 Two primary containment air locks shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment. ACTIONS

1. Entry and exit is permissible to perform repairs of the affected air lock components.

2 Separate Condition entry is allowed for each air Iock. 3 Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more primary NOT Ee-_ containment air Iocks with 1. Required Actions A.1, A.2, one primary containment and A.3 are not applicable if both air lock door inoperable. doors in the same air lock are inoperable and Condition C is entered.

2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.

continued PERRY - UNIT 1 3,6-3 Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY Primary Containment Air Locks CLEAN, RETYPED PAGE 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) c.3 Restore air lock to 24 hours OPERABLE status D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, B, or AND C not met in MODE 1, 2, or 3. D.2 Be in MODE 4. 36 hours E. Required Action and E.1 Suspend movement of lmmediately associated Completion recently irradiated fuel Time of Condition A, B, or assemblies in the primary C not rnet during containment. movement of recently irradiated fuel assemblies in the primary containment. PERRY _ UNIT 1 3.6-6 Amendment No,

TECH NIGAL SPECI FICATION PtIUs PROVIDED FOR CONTEXT 3 6.I,3 3.6 COHTAIHI-IEHT SYSTEFIS 3.6 I .3 Primary f,onta innrent Isol at ion UaI ves (PCIUS) Lm 3"6.1.3 Each PCIV, except containment Yacuum breakers, shall be OPERABIE. APPLIIABILITY: ].l0DES l, ?, and 3, tlhen associ a ted lnstrumentation is required to ht 0PERABLE per LCO 3.3 .6,1, 'Primary Containment and Drywell I sol ati ont nstrumentati on. ACTIOHS

                                            ----H0TES--*--

I Penetration flou paths exce pt for the inboard 12 inch purge valve penetration flow Faths may be unisol ated lntennittently under admini strative control s, ?, Separate Condition entry is alloned for each penetration flow patft. 3 Enter applicable [onditisns and Required f,ctions for systems made tnopereble hy P[IUs. 4 Enter applicable [ondltions rnd Requtred flctions of L[0 3.8.I.1, 'Primar] Containment-Operatif,g,' rhen PCIV leakage resul ts in exceeding overal J containment leakage rate acceptance criteria tn tl0DE$ I, 2, and 3. IOHDI TIOI{ ffiOU I RTD AET I OH tot-tPlETI0l{ I I},tt A. One or more A I I sol ate the affected 4 hours except penetration flon paths penetration flotv path for main steam with one FilU by use of at least line' inoperable except due one closed and de* t o 'l eakage not ui thi n activated automatic AHD limit. Yalve, closed manual val ue, b'l i nd f I ange, I hours f or ma i rr or check valve with steanr 1 ine flow through the val ue seeured. AUE ( cont i nued) PTRflY U]{IT I 3"5-9 Amendment Ho. 59

TEC H NIGAL SPECIFICATION INFORMATION ONLY PCIVs CLEAN, RETYPED PAGE 3.6.1.3 ACTIONS CONDITION REQUIRED ACT]ON COMPLETION TITVIE D. (continued) D.3 Perform SR 3.6.1.3.6 for Once per 92 days the resilient seal purge valves closed to comply with Required Action D.1. E. Required Action and E.'l Be in MODE 3. 12 hours associated Completion Time of Condition A, B, AND C, or D not met in IVIODE 1 , 2, or 3. E.2 Be in tvlODE 4. 36 hours F. Required Action and F.1 Suspend movement of lmmediately associated Completion recently irradiated fuel Time of Condition A, B, assemblies in primary C, or D not met for containment. PCIV(s) required to be OPERABLE during movement of recently irradiated fuel assemblies in the primary containment. PERRY _ UNIT 1 3.6-13 Amendment No

TEC H NICAL SPECI FICATION INFORMATION ONLY PCIVs CLEAN, RETYPED PAGE 3.6.1.3 lntentionally blank PERRY _ UNIT 1 3.6-14 Amendment No.

TEGHNICAL SPECIFICATION INFORMATION ONLY Primary Containment-Shutdown CLEAN, RETYPED PAGE 3.6.1 .10 3.6 CONTAINTVIENT SYSTEMS 3.6.1 .1 0 Primary Containment-Shutdown LCO 3.6.1 .10 Primary containment shall be OPERABLE APPLICABILITY: During movement of recently irradiated fuel assemblies in the primary containment. ACTIONS CONDITION REQUIRED ACTION COI\4PLETION TIME A. Primary containment 4.1 Suspend movement of Immediately inoperable, recently irradiated fuel assemblies in the primary containment. PERRY - UNIT 1 3.6-29 Amendment No.

TECH NICAL SPECI FICATION INFORMATION ONLY Containment Vacuum Breakers CLEAN, RETYPED PAGE 3.6.1.1 1 3,6 CONTAINMENT SYSTETMS 3.6.1 .1 1 Containment Vacuum Breakers LCO 3,6.1 .1 1 Three containment vacuum breakers shall be OPERABLE and four containment vacuum breakers shall be closed. APPLICABILITY: IVIODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment. ACTIONS NOTE--. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating" when the containment vacuum relief subsystem leakage results in exceeding overall containment leakage acceptance criteria. CONDITION REQUIRED ACTION COTUPLETION TIME A, -NOTE.. 4.1 Close the associated motor 4 hours Separate Condition entry operated isolation valve. is allowed for each containment vacuum AND breaker. 4.2 Restore required 72 hours containment vacuum One or two containment breaker to OPERABLE vacuum breakers not status. closed. OR One required containment vacuum breaker inoperable for other reasons. (contlnued) PERRY - UNIT 1 3.6-31 Amendment No.

TECH NICAL SPECIFICATION INFORMATION ONLY Containment Vacuum Breakers GLEAN, RETYPED PAGE 3.6.1 .11 ACTIONS continued CONDITION REQUIRED ACTION COTVIPLETION TITVIE B. Required Action and .NOT associated Completion Only applicable in MODE 1, 2 or 3. Time of Condition A not met. 8.1.1 Be in MODE 3. 12 hours OR AND Three or more containment vacuum 8.1.2 Be in MODE 4. 36 hours breakers not closed. AND OR

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

Two or more required Only applicable during movement containment vacuum of recently irradiated fuel breakers inoperable for assemblies in the primary other reasons. containment. 8.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment. PERRY - UNIT 1 3.6-32 Amendment No.

TECH NICAL S PECI FICATION INFORMAT]ON ONLY Containment H umidity Control CLEAN, RETYPED PAGE 3.6.1 .12 3.6 CONTAINMENT SYSTETVIS 3.6.1.12 Containment Humidity Control LCO 3.6.1 .12 Containment average temperature-to-relative humidity shall be maintained within Iimits. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment. ACTIONS CONDITION REQUIRED ACTION COIMPLETION TITVIE A. Requirements of LCO not 4.1 Restore containment I hours met. average temperature-to-relative humidity to within limits. (continued) PERRY _ UNIT 1 3.6-34 Amendment No.

TECH NICAL SPECI FICATION INFORMATION ONLY Containment Humidity Control CLEAN, RETYPED PAGE 3.6.1.12 ACTIONS conti CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met in [\4ODE 1, 2, or 3. 8.2 Be in MODE 4. 36 hours C. Required Action and c.1 Suspend movement of Immediately associated Completion recently irradiated fuel Time of Condition A not assemblies in the primary met during movement of containment. recently irradiated fuel assemblies in the primary containment. SU RVEI LLANCE REQUIREMENT SURVEILLANCE FREQUENCY sR 3.6.1 .12.1 Verify containment average temperature-to-relative In accordance humidity to be within Iimits. with the Surveillance Frequency Control Program PERRY _ UNIT 1 3.6-35 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY Secondary Containment CLEAN, RETYPED PAGE 3.6.4.1 3.6 CONTAINTVIENT SYSTEI/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 recently irradiated fuel assemblies in the primary containment. ACTIONS CONDIT!ON REQUIRED ACTION COMPLETION TIME A. Secondary containment 4.1 Restore secondary 4 hours inoperable in MODE 1, 2, containment to or 3. OPERABLE status. B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met. 8.2 Be in tvlODE 4. 36 hours (continued) PERRY - UNIT 1 3.6-51 Amendment No.

TEC HNICAL SPECI FICATION INFORMATION ONLY Secondary Containment CLEAN, RETYPED PAGE 3.6.4.1 ACTIONS contin CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment c.1 Suspend movement of lmmediately inoperable during recently irradiated fuel movement of recently assemblies in the primary irradiated fuel assemblies containment. in the primary containment. SU RVE I LLANCE REQ U I R EIVI ENTS SURVEILLANCE FREQUENCY sR 3.6.4.1.1 Verify secondary containment vacuum is I 0.66 inch ln accordance of vacuum water gauge. with the Surveillance Frequency Control Program sR 3.6.4.1 .2 Verifo the primary containment equipment hatch is ln accordance closed and sealed and the shield blocks are with the installed adjacent to the shield building. Surveillance Frequency Control Program sR 3.6.4.1.3 Verify each secondary containment access door is ln accordance closed, except when the access opening is being with the used for entry and exit. Surveillance Frequency Control Program PERRY - UNIT 1 3.6-52 Amendment No.

TECHNICAL SPECI FICATION INFORMATION ONLY SCIVs CLEAN, RETYPED PAGE 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment lsolation Valves (SClVs) LCO 3.6.4.2 Each SCIV shall be OPERABLE. APPLICABILITY MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment. ACTIONS

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

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

CONDITION REQUIRED ACTION COI\4PLETION TIME A. One or more penetration A.1 lsolate the affected I hours flow paths with one SCIV penetration flow path by inoperable. use of at least one closed manual valve or blind flange. AND continued PERRY _ UNIT 1 3.6-53 Amendment No.

TEGHNICAL SPECI FICATION INFORMATION ONLY GLEAN, RETYPED PAGE SCIVs 3.6.4.2 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Suspend movement of lmmediately associated Completion recently irradiated fuel Time of Condition A or B assemblies in the primary not met during movement containment. of recently irradiated fuel assemblies in the primary containment. SU RVEI LLANCE REQ U I REIVI E NTS SURVEILLANCE FREQUENCY sR 3.6.4.2.1 .---------NOT

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

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

Verify each secondary containment isolation manual ln accordance valve and blind flange that is not locked, sealed, or with the othenuise secured and is required to be closed Surveillance during accident conditions is closed. Frequency Control Program PERRY - UNIT 1 3.6-55 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY AEGT System CLEAN, RETYPED PAGE 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6,4.3 Annulus Exhaust Gas Treatment (AEGT) System LCO 3.6.4.3 Two AEGT subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TITVIE A. One AEGT subsystem 4.1 Restore AEGT subsystem 7 days inoperable. to OPERABLE status. B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met in MODE 1, 2, or 3. 8.2 Be in MODE 4. 36 hours C. Required Action and c.1 Place OPERABLE AEGT Immediately associated Completion subsystem in operation. Time of Condition A not met during movement of OR recently irradiated fuel assemblies in the primary containment. contin PERRY - UNIT 1 3.6-56 Amendment No

TECHNICAL SPECIFICATION INFORMATION ONLY AEGT System CLEAN, RETYPED PAGE 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) c.2 Suspend movement of Immediately recently irradiated fuel assemblies in the primary containment. D. Two AEGT subsystems D.1 Enter LCO 3.0.3, lmmediately inoperable in MODE 1, 2, or 3. E. Two AEGT subsystems E.1 Suspend movement of lmmediately inoperable during recently irradiated fuel movement of recently assembles in the primary irradiated fuel assemblies containment. in the primary containment. PERRY - UNIT 1 3.6-57 Amendment No.

TECHNICAL SPECI FICATION INFORMATION ONLY CLEAN, RETYPED PAGE CRER System 3.7.3 3.7 PLANT SYSTEIVIS 3.7.3 Control Room Emergency Recirculation (CRER) System LCO 3.7.3 Two CRER subsystems shall be OPERABLE. NOT The Control Room Envelope (CRE) boundary may be opened intermittently under administrative control. APPLICABILITY: IVIODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building. ACTIONS CONDITION REQUIRED ACTION COIflPLETION TIIVIE A. One CRER subsystem 4.1 Restore CRER subsystem 7 days inoperable for reasons to OPERABLE status. other than Condition B. B. One or more CRER 8.1 Initiate action to implement lmmediately subsystems inoperable mitigating actions. due to inoperable CRE boundary in fVlode 1, 2, AND or 3. 8.2 Verify mitigating actions 24 hours ensure CRE occupant radiological exposures will not exceed limits, and CRE occupants are protected from chemical and smoke hazards. AND 8.3 Restore CRE boundary to 90 days OPERABLE status (continued) PERRY _ UNIT 1 3.7-4 Amendment No.

TECHNICAL SPECI FICATION INFORMATION ONLY CRER System CLEAN, RETYPED PAGE 3.7.3 ACTIONS continu CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B AND not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours D. Required Action and associated Completion LCO 3.0.3 is not applicable Time of Condition A not met during movement of recently irradiated fuel D.1 Place OPERABLE CRER Immediately assemblies in the primary subsystem in emergency containment or fuel recirculation mode. handling building. OR D.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment and fuel handling building. E. Two CRER subsystems E.1 Enter LCO 3.0.3 lmmediately inoperable in MODE 1, 2, or 3 for reasons other than Condition B. (continued) PERRY - UNIT 1 3.7-5 Amendment No

TECHNICAL SPECIFICATION INFORMATION ONLY CRER System CLEAN, RETYPED PAGE 3.7.3 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME F. Two CRER subsystems F.1 Suspend movement of Immediately inoperable during recently irradiated fuel movement of recently assemblies in the primary irradiated fuel assemblies containment and fuel in the primary handling building. containment or fuel handling building. OR One or more CRER subsystems inoperable due to inoperable CRE boundary during movement of recently irradiated fuel assemblies in the primary containment or fuel handling building. SU RVEI LLANCE REQ U I RETVI E NTS SURVEILLANCE FREQUENCY sR 3.7.3.1 Operate each CRER subsystem for > 10 continuous ln accordance hours with the heaters operating. with the Surveillance Frequency Control Program sR 3.7.3.2 Perform required CRER filter testing in accordance ln accordance with the Ventilation Filter Testing Program (VFTP). with the VFTP (continued) PERRY _ UNIT 1 3.7-6 Amendment No.

TECHNICAL SPEGIFICATION INFORMATION ONLY Control Room HVAC System CLEAN, RETYPED PAGE 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Heating, Ventilating, and Air Conditioning (HVAC) System LCO 3.7.4 Two control room HVAC subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One control room HVAC 4.1 Restore control room 30 days subsystem inoperable. HVAC subsystem to OPERABLE status. B. Two control room HVAC 8.1 Verify control room air Once per 4 hours subsystems inoperable. temperature is < 90'F. AND 8.2 Restore one control room 7 days HVAC subsystem to OPERABLE status. C. Required Action and c.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B AND not met in MODE 1,2, or

3. C.2 Be in MODE 4. 36 hours (continued)

PERRY - UNIT 1 3.7-8 Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY Control Room HVAC System GLEAN, RETYPED PAGE 3.7.4 ACTIONS co CONDITION REQUIRED ACTION COMPLETION TIME D, Required Action and associated Completion LCO 3,0.3 is not applicable. Time of Condition A not met during movement of recently irradiated fuel D.1 PIace OPERABLE control lmmediately assemblies in the primary room HVAC subsystem in containment or fuel operation. handling building. OR D.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment and fuel handling building. (continued) PERRY - UNIT 1 3.7-9 Amendment No

TECHNICAL SPECI FICATION INFORMATION ONLY Control Room HVAC System CLEAN, RETYPED PAGE 3.7,4 ACTIONS contin 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 recently irradiated fuel E.1 Suspend movement of lmmediately assemblies in the primary recently irradiated fuel containment or fuel assemblies in the primary handling building. containment and fuel handling building. SU RVEI LLANCE REQU!REMENTS SURVEILLANCE FREQUENCY sR 3.7.4.1 Verify each control room HVAC subsystem has the ln accordance capability to remove the assumed heat load. with the Surveillance Frequency Control Program PERRY - UNIT 1 3.7-10 Amendment No

TECHNICAL SPEGIFICATION AC Sout"ces - Shutdorm 3.8. e PROVIDED FOR GONTEXT 3.8 ELECTRICAL PO.IER sYsTEtfi 3.8.2 AC Sourtres

                     - Shutdorrn LCO    3.8.2         The following AC        electrical pe{er sources shall be 0PERABLE:

a One gualified circuit between the offsite tra nsmr 551 0n neturirk and the onsite Class tE AC electrical distribution subsystem(s) required by LCO 3.8 ilT" "Di stri buti on Systems - Shutdinn' : b One diesel generator (DG) capable of supplying one di vi si on of-the Di vi si on 1 o'r e onsi te LJ a-ss If AC

eI ectri cg 1 po,rer di stri buti on subsystem( s requi red by LCO 3,8.8; bnd c One gual i fi ed ci rcu'i t , other than the ci rcui t i n LCO S.8.2.a. between the offsite transmission network and the Division 3 onsite Class lE electrical porler distribution subsystem, or the Division 3 DG capable of supp'ly'ing the Di vi si on 3 onsi te Cl ass IE AC el ebtri cal por{er distribution subsystem. when the Division 3 onsite Ctrass lE electrical pcuer distribution subsystem is requi red by LCO 3.8. B. APPL ICABI LITY : IffiDES 4 and 5 , Duri ng movement of recently i rradi ated fuel assembJ i es 'in the pri rTlary contai nment 0r fuel handl i ng bui I di ng . PffiY - I}IIT 1 3,8-17 AlEndttEnt th. 102

TECHNICAL SPECIFICATION INFORMATION ONLY AC Sources - Shutdown CLEAN, RETYPED PAGE 3,8.2 ACTIONS NOTE LCO 3.0.3 is not applicable. CONDITION REQUIHED ACTION CO]\4PLETION TIME A. LCO ltem a not met. -------NOTE-------- Enter applicable Condition and Required Actions of LCO 3.8.8, when any required division is de-energized as a result of Condition A. A.1 Declare required feature(s) lmmediately with no offsite power available from a required circuit inoperable. OR 4.2.1 Suspend CORE lmmediately ALTERATIONS, AND 4.2.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment and fuel handling building. AND (continued) PERRY _ UNIT 1 3.8- 1 B Amendment No.

TECHNICAL SPECIFICATION INFORMATION ONLY AC Sources - Shutdown CLEAN, RETYPED PAGE 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (Continued) 4.2.3 lnitiate action to restore Immediately required offsite power circuit to OPERABLE status. B. LCO ltem b not met. 8.1 Suspend CORE lmmediately ALTERATIONS. AND 8.2 Suspend movement of lmmediately recently irradiated fuel assemblies in primary containment and fuel handling building. AND B.3 Initiate action to restore lmmediately required DG to OPERABLE status. C. LCO ltem c not met. c.1 Declare High Pressure 72 hours Core Spray System inoperable. PERRY - UNIT 1 3.8-19 Amendment No.

TECHNICAL SPECI FICATION INFORMATION ONLY AC Sources - Shutdown GLEAN, RETYPED PAGE 3.8.2 SU RVEI LLANCE REQU I REMENTS SURVEILLANCE FREQUENCY sR 3.8.2.1

                   ; il;];i;;,ilfi; performed:

I8l:i;;;il;; ;; SR 3.8.1.3, SH 3.8.1 .8 through SH 3.8.1 .'l 6, SH 3.8.1 .1 8, and SR 3.8.1 .1 9.

2. SR 3.8.1.12 and SR 3.8.'l .19 are not required to be met when the associated ECCS subsystem(s) are not required to be OPERABLE per LCO 3.5.2, "Reactor Pressure Vessel (RPV)

Water lnventory Control." In accordance For AC sources required to be OPERABLE, the with applicable following SRs are applicable: SRs sR 3.8.1.1 sR 3.8.1.7 sR 3.8.1 .1 4 sR 3.8.1 .2 sR 3.8.1 .9 sR 3.8.1 .15 sR 3.8.1.3 sR 3.8.1.10 sR 3.8.1.16 sR 3.8.1.4 sR 3.8.1.11 sR 3.8.1.18 sR 3.8,1.5 sR 3.8.1 .12 sR 3.8.1.19 sR 3.8.1.6 sFr 3.8.1 .1 3 PERRY _ UNIT 1 3.8-20 Amendment No.

TECHNICAL S PECIFICATION I PROVIDED FOR GONTEXT DC Sources-Shutdown 3.8.5 3 .8 "TLECTRICAL POIdER SYSTEHS 3.8.5 DC Sources

                    -Shutdown LCo 3"8.5          The fo I 'lowi ng DC el ectri cal polver subsystems sha I I be OPIRABLE:

a. One Class lE DC electrical power subsystem capable of supply'ing one div'ision of the D'ivision 1 0r 2 onsite Class IE electrical pourer distribution subsystem(s) requ j red by LCO 3 8 .8 , "Dl stri but'ion Systems

Shdtdowr't":*

b. One Class IE battery 0r battery charger. other than the power subsystem in LCO 3.8.5.a, capable of DC electrical supplying the remainjng Division 1 or Division 2 onsite Class lt 0C electr"ical polver distribution subsystem when requi red by LCO 3. B. B; and

c. The D'ivision 3 DC electrical power subsystem capable of supplying the Divlsion 3 onsite Class 1E DC elebtrical t

poi+er di stri but i on subsystem whdn the Dj v"isi on 3 onsi te Class lt DC electrical power distrtbution subsystem is requi red by LCO 3.8.8. APPLICABILITY: lffiD[S 4 and 5, During movement of recently 'i rrad'iated fuel assembl'ies in fhe primary containment or fuel handling building. I I PERRY . UNIT I 3. B-28 Amendment No. Io2,

TECH NICAL SPECIFICATION INFORMATION ONLY DC Sources - Shutdown CLEAN, RETYPED PAGE 3.8.5 ACTIONS LCO 3.0.3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required DC A.1 Declare affected required Immediately electrical power feature(s) inoperable. subsystems inoperable. OR A.2.1 Suspend CORE lmmediately ALTERATIONS. AND 4.2.2 Suspend movement of lmmediately recently irradiated fuel assemblies in the primary containment and fuel handling building. AND 4.2.3 lnitiate action to restore lmmediately required DC electrical power subsystems to OPERABLE status. PERRY - UNIT 1 3.8-29 Amendment No.

TECHNICAL SPEGIFICATION Distribution Systems

                                                                                    - Shutdown 3.8.8 PROVIDED FOR GONTEXT 3.8    ELECTHICAL POltlER SYSTEI'IS 3.8,8 Distributisn          Systems
                                    - Shutdown LCo 3,8.8              The necessary portions of the Division 1, Division 2, and Division 3 Af and DC electrical poiler distribution suhsystems shall be 0PERABLE to iupport equipnent requ'ired to be OPERABLE.

APPLICABILITY: lfi)DES 4 and 5, During movement of recently irradiated fuel assemblies in the primary containment or fuel handling buildtng. ACTI0I'lS HOTE LCO 3"0.3 is not aBplicable. C0I'IDITI0N REQUIRED ACTIOH COHPLETION TIHE A One or more requi red A. t Declare assocjated I nnedi ate l y AC or DC electrical supported requ'i red ps{er distribution feature( s ) subsystems i noperable. i noperabl e . OE 4.2.1 Susuend Cmt kmedi ately ALTERATXONS. ax! A.?.? Suspend movernent of Inrnedi ate'ly recently i rrad'iated fuel assembl 'i es i n the primary containment and fuel handl i ng bui I d'ing = AND (cont'inued) PffiRY . IT{IT 1 3.8-38 PneIffi l-h. lo2

TECHNICAL SPECIFICATION INFORMATION ONLY Distribution Systems - Shutdown CLEAN, RETYPED PAGE 3.8.8 ACTIONS CONDITION REQUIHED ACTION COh/PLETION TIME A. (continued) 4.2.s Initiate actions to restore lmmediately required AC and DC electrical power distri bution subsystems to OPERABLE status. AND 4.2_4 Declare associated lmmediately required shutdown cooling subsystem (s) i noperable and not in operation. SURVEI LLANCE REQU I REhTENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and voltage to ln accordance required AC and DC electrical power distribution with the subsystems. Surveillance Frequency Control Program PERRY _ UNIT 1 3.8-39 Amendment No.

Attachment 4 L-17-045 PROPOSED TECHNTCAL SPECIFICATlON BASES CHANGES (MARK-UP) (PROVIDED FOR INFORMATION ONLY) (127 pages follow)

TS BASES MARK.UP . PROVIDED PERRY NUCLEAR POWER PLANT FOR INFORMATION ONLY Technical Specifications Bases (TSB) Table of Contents B 2.0 SAFETY LIMITS (SLs) B 2.1 .1 Reactor Core SLs... . B 2.0-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL B 2.0-7 B 3.0 LIMlTING CONDITTON FOR OPERATTON (LCO) APPLICABILITY ..... B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY. ...8 3.0-10 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGTN (SDM)... B 3.1-1 B 3.1.2 Reactivity Anomalies... .... B 3.1-8 B 3.1.3 Control Rod OPERABI LITY. B 3.1-1 3 B 3.1.4 Control Rod Scram Times B 3.1-22 B 3.1 .5 Control Rod Scram Accumulators" B 3. 1-29 B 3.1.6 Control Rod Pattern B 3. 1-34 B 3.1.7 Standby Liquid Control (SLC) System B 3. 1-39 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves.. B 3.1-45 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)... ...8 3.2-1 B 3.2.2 MtNlMUrvr cRlTlcAL POWER RATIO (MCPR) . .... B 3,2-6 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR). B 3.2-10 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) lnstrumentation B 3.3-1 B 3.3.1.2 Source Range Monitor (SRM) lnstrumentation.. B 3.3-33 B 3.3.1 .3 Oscillation Power Range Monitor (OPRM) lnstrumentation.. B 3.3-41a B 3.3.2.1 Control Rod Block Instrumentation B 3.3-42 B 3.3.3.1 Post Accident Monitoring (PAM) lnstrumentation. B 3.3-51 B 3.3.3.2 Remote Shutdown System. ..... B 3.3-63 B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation ..... 83.3-68 B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) I nstrumentation B 3.3-79 B 3.3.s.1 Emergency Core Cooling System (ECCS) lnstrumentation B 3.3.88 B 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventorv Control lnstrumentation . E_3-31-3_Reactor Core Isolation Cooling (RCIC) System lnstrumentation ...8 3.3-124 B 3.3.6.1 Primary Containment and Drywell Isolation lnstrumentation. B 3.3-136 B 3.3.6.2 Residual Heat Removal (RHR) Containment Spray System lnstrumentation. ... B 3.3-174 B 3.3.6.3 Suppression Pool Makeup (SPMU) System Instrumentation B 3.3-185 B 3.3.6.4 Relief and Low-Low Set (LLS) lnstrumentation B 3.3-196 B 3.3.7.1 Control Room Emergency Recirculation (CRER) System lnstru mentation. B 3.3-202 B 3.3.8.1 Loss of Power (LOP) Instrumentation . B 3.3-212 B 3.3.8.2 Reactor Protection System (RPS) Electric Power ltIonitoring..... . B 3.3-219 PERRY _ UNIT 1 Revision 0 I Page i

TS BASES MARK.UP . PROVIDED TSB Table of Contents (continued) FOR INFORMATION ONLY B 3.4 REACTOR COOLANT SYSTETVT (RCS) B 3.4.1 Recirculation Loops Operating B 3.4-1 B 3.4.2 Flow Control Valves (FCVs) B 3.4-9 B 3.4.3 Jet Pumps..... B 3.4-13 B 3.4.4 Safety/Relief Valves (S/RVs) B 3.4-18 B 3.4.5 RCS Operational LEAKAGE.. B 3.4-23 B 3.4.6 RCS Pressure lsolation Valve (PIV) Leakage B 3.4-28 B 3.4.7 RCS Leakage Detection lnstrumentation B 3.4-33 B 3.4.8 RCS Specific Activity. B 3.4-40 B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown... .. B 3.4-44 B 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown... .. B 3.4-49 B 3.4.1 1 RCS Pressure and Temperature (PfD Limits. B 3.4-54 B 3.4.12 Reactor Steam Dome Pressure B 3.4-64 B 3.5 ETVIERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONIEOL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating.. .,. B 3.5-1 B 3.5.2 B 3.5-15 B 3.5.3 RCIC System .. B 3.5-21 B 3.6 CONTAINMENT SYSTETUS B 3.6.1.1 Primary Containment - Operating......... .B 3.6-1 B 3.6.1.2 Primary Containment Air Locks. ..8 3.6-7 B 3.6.1 3 Primary Containment Isolation Valves (PClVs).... ....... B 3.6-17 B 3.6.1 4 Primary Containment Pressure. ... ... .... B 3.6-33 B 3.6.1 5 Primary Containment Air Temperature. ....... B 3.6-36 B 3.6.1 6 Low-Low Set (LLS) Valves. .... ....8 3.6-39 B 3.6.1 7 Residual Heat Removal (RHR) Containment Spray System.... .. ... ... B 3.6-43 B 3.6.1 I Feedwater Leakage Control System (FWLCS). . B 3.6-48 B 3.6.1 I Main Steam Shutoff Valves B 3.6-51 B 3.6.1 10 Primary Containment - Shutdown..... B 3.6-55 B 3.6.1 11 Containment Vacuum Breakers. B 3.6-59 B 3.6.1 12 Containment Humidity Control.. B 3.6-65 B 3.6.2.1 Suppression Pool Average Temperature. B 3.6-70 B 3.6.2.2 Suppression Pool Water Level B 3.6-75 B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling B 3.6-79 B 3.6.2.4 Suppression Pool Makeup (SPMU) System B 3.6-83 B 3.6.3.1 Deleted B 3.6.3.2 Primary Containment and Drywell Hydrogen Igniters B 3.6-95 B 3.6.3.3 Combustible Gas l\4ixing System B 3.6-101 B 3.6.4.1 Secondary Containment. B 3.6-106 B 3.6.4.2 Secondary Containment Isolation Valves (SClVs) .... .,.. B 3.6-111 B 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System .... .... B 3.6-1 18 B 3.6.5.1 Drywell B 3.6-123 B 3.6.5.2 Drywell Air Lock. ..8 3.6-128 B 3.6.5.3 Drywell lsolation Valves B 3.6-136 B 3.6.5.4 Drywell Pressure B 3.6-145 B 3.6.5.5 Drywell Air Temperature. B 3.6-148 B 3.6.5.6 Drywell Vacuum Relief System. B 3.6-151 PERRY - UNIT 1 Revision O Page ii

TS BASES MARK-UP . PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 B 3.3 INSTHUh/ENTATION B 3.3.5.'l Emergency Core Cooling System (ECCS) lnstrumentation BASES BACKGROUND The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that fuel is adequately cooled in the event of a design basis accident or transient. For most anticipated operational occurrences (AOOs) and Design Basis Accidents (DBAs), a wide range of dependent and independent parameters are monitored. Portions of this ECCS instrumentation actuate the Annulus Exhaust Gas Treatment (AEGT) subsystems and the diesel generators (DGs), in addition to the ECCS subsystems (Low Pressure Core Spray (LPCS), Low Pressure Coolant Injection (LPCI), High Pressure Core Spray (HPCS), and Automatic Depressurization System (ADS)). The supported systems are described in the Bases for:

                      -     LCO 3.5.1 and 3.5.2
                                                  .ECCs-Operating" and Reactor Pressure V LnPV) Water lnvento
                      -     LCO 3.6.4.3 "Annulus Exhaust Gas Treatment (AEGT) System," and
                      -     LCO 3.8.1 and 3.8.2 "AC Sources-Operating" and "AC Sources-Shutdown".

Low Pressure Core Sprav Svstem The LPCS System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level - Low Low Low, Level 1 or Drywell Pressure - High. Each of these diverse variables is monitored by two redundant transmitters, which are, in turn, connected to two trip units. The outputs of the four trip units (two trip units from each of the two variables) are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. The initiation signal is a sealed in signal and must be manually reset. The logic can also be initiated by use of a manual push button. Upon receipt of an initiation signal, the LPCS pump is started immediately after power is available. The LPCS test valve to suppression pool, which is also a primary containment isolation valve (PCIV), is closed on a LPCS initiation signal to allow full system flow assumed in the accident analysis and maintains containment isolation in the event LPCS is not operating. (continued) PERRY _ UNIT 1 B 3.3-88 Revision No. +

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE Function must have a required number of OPERABLE channels, with SAFETY their setpoints within the specified Allowable Values, where appropriate. ANALYSES, LCO, The actual setpoint is calibrated consistent with applicable setpoint and APPLICABILITY methodology assumptions. Each ECCS subsystem must also (continued) respond within its assumed response time. Table 3.3.5.1-1 is modified by a&smote three feetnetes, Feetnete (a) is added te speei{y that the

                   +ni Footnote (b) is added to show that certain ECCS instrumentation Functions also perform DG and AEGT subsystem initiation. eee+nete-((}

i Allowable Values are specified for each ECCS Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.9., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.9., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. (continued) PERRY _ UNIT 1 B 3.3-95 Revision No. I I

TS BASES MARK-UP. PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE Low Pressure Core Sprav and Low Pressure Coolant lniection Svstems SAFETY 1 .a. 2.a. Reactor Vessel Water Level - Low Low Low, Level

                                                                                   'l ANALYSES, LCO, and APPLICABILITY (continued)      Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should FIPV water level decrease too far, fuel damage could result. The low pressure ECCS and associated DGs are initiated at Level 1 to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. The AEGT System also receives Level 1 initiation signals to ensure a subsystem will operate following events that challenge core coverage.

The Reactor Vessel Water Level - Low Low Low, Level 1 is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in References 1 and 3. ln addition, the Reactor Vessel Water Level - Low Low Low, Level 1 Function is assumed in the analysis of the DBA LOCA (Ref. 2). The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level - Low Low Low, Level 1 signals are initiated from four Ievel 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 Allowable Value is chosen to allow time for the low 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 required to be OPERABLE when the associated ECCS or AEGT subsystem is required to be OPERABLE, to ensure that no single instrument failure can preclude system initiation. (Two channels input to Division 1 , while the other two channels input to Division 2.) Per foetnete (a) te Tab]e 3,3,5,1-1,-this trCeS Funetien is enly required te be OPERABTE te suBpert ECS initiatien in MODES 'l tGO 3,5,2, Eeeause pertiens ef the EGGS instrumentatien alse aetuate the AEGT subsystems; feetnete (f) te Table 3,3,5,1 1 requires this EeeS Funetien te be OPERABTE when the AEGT subsystems are required te Refer to LCO 3.5.1, "fccs-Operatin S.n*deu+nrfor Applicability Bases for the low pressure ECCS subsystems_

                                                                               ;-and LCO 3.6.4.3, "Annulus Exhaust Gas Treatment (AEGT) System," for Applicability Bases for AEGT System.

PERRY _ UNIT 1 B 3.3-96 Revision No. I I

TS BASES MARK-UP . PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE 1.c, 2.c. Low Pressure Coolant lniection Pump A and Pump B Start - SAFETY Time Delav Relav (continued) ANALYSES, LCO, and APPLICABILITY However, since the time delay does not degrade ECCS operation, it remains in the pump start logic at all times. The LPCI Pump Start - Time Delay Relays are assumed to be OPERABLE in 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 power sources. There are two LPCI Pump Start - Time Delay Flelays, one in each of the RHR "A" and RHR "B" pump start Iogic 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 same ESF bus, to perform their intended function within the assumed ECCS RESPONSE Tlt\4ES (e.g., as in the case where both ECCS pumps on one ESF bus start simultaneously due to an inoperable time delay relay). This still Ieaves two of the four low pressure ECCS pumps OPEFTABLE;thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Value forthe LPCI Pump Start - Time Delay Relay is chosen to be long enough so that most of the starting transient of the first pump is complete before starting the second pump on the same 4.16 kV emergency bus and short enough so that ECCS operation is not degraded. Each LPCI Pump Start - Time Delay Relay Function is only required to be OPERABLE when the associated LPCI subsystem is required to be OPERABLE. Per feetnete (a) te Table 3,3,5,1 1' this EeeS Fun6+ien is

                                                                                       . Refer to LCO 3.5.1    #or                   Applicability Bases for the LPCI subsystems.

1.d. 1,e. 2.d. Reactor Vessel Pressure - Low (lniection Valve Permissive) Low reactor vessel 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 Vessel Pressure - Low (lnjection Valve Permissive) is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. ln addition, (continued) PERRY _ UNIT 1 B 3.3-98 Revision No. g

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE 1.d, 1.e, 2.d. Reactor Vessel Pressure - Low (lniection Valve Permissive) SAFETY (continued) ANALYSES, LCO, and APPLICABILITY the Reactor Vessel Pressure - Low (lnjection Valve Permissive) Function is directly assumed in the analysis of the recirculation line break (Ref. 2). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the Iimits of 10 CFR 50.46. The Reactor Vessel Pressure - Low (lnjection Valve Permissive) signals are initiated from one pressure transmitter for each low pressure ECCS System that senses the reactor pressure. The Allowable Value is low enough to prevent overpressurizing the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of '10 CFR 50.46. One channel of Reactor Vessel Pressure - Low (lnjection Valve Permissive) Function per associated low pressure ECCS subsystem is required to be OPERABLE when the associated ECCS is required to be OPERABLE. Perfeetnete (a) te Table 3,3,5,1 1, this EGCS Funetien is vr rrJ r vYsrr vv Lv

                                                                                     . Refer to LCo3.5.1#orApplicabilityBasesforthelowpresSUre ECCS subsystems.

1 .f. 1 .o. 2.e. Low Pressure Coolant lniection and Low Pressure Core Sprav Pumn Discharqe Flow - Low {Bvpass) The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow 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 OPEFTABLE and capable of closing the minimum flow valves 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. One flow transmitter per ECCS pump is used to detect the associated subsystems' flow rates. continu PERRY _ UNIT 1 B 3.3-99 Revision No. I

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.f, 1.q. 2.e. Low Pressure Coolant lniection and Low Pressure Core SAFETY Sprav Pump Discharoe Flow - Low (Bvpass) (continued) ANALYSES, LCO, and APPLICABILITY The logic is arranged such that each transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for I seconds after the transmitters and associated trip units detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the RHR shutdown cooling mode (for RHH A and RHR B), 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 core. Each channel of Pump Discharge Flow - Low (Bypass) Function (one LPCS channel and three LPCI channels) is only required to be OPEHABLE when the associated ECCS is required to be OPERABLE, to ensure that no single instrument failure can preclude the ECCS function. Refer to LCO 3.5.1 a*#l4o-.53 for Applicability Bases for the low pressure ECCS subsystems. 1.h, 2.f. Jvlanual lnitiation The lVanual Initiation push button channels introduce signals into the appropriate ECCS logic to provide manual 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 B and LPCI C). The Manual lnitiation Function is not assumed in any accident or transient analyses in the USAR. 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. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. Each channel of the Nlanual lnitiation Function (one channel per Division) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE. Per feetnete (a) te Table 3,3,5,'l 1' this EeeS Refer toLCo3.5.1#orApplicabilityBasesforthelowpresSure ECCS subsystems. (continued) PEFIRY _ UNIT 1 B 3.3-100 Revision No. I I

TS BASES MARK-UP. PROVIDED FOR INFORMATION ONLY ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE Hiqh Pressure Core Sprav Svstem SAFETY ANALYSES, LCO, 3.a. Reactor Vessel Water Level - Low Low. Level 2 and APPLICABILITY (continued) Low FIPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCS System and associated DG is initiated at Level 2 to maintain level above the top of the active 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 Fleferences 1 and 3. The Reactor Vessel Water Level - Low Low, Level 2 Function associated with HPCS is assumed in the analysis of a DBA LOCA (Ref. 2). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the f uel peak cladding temperature remains below the Iimits of 10 CFR 50.46. Reactor Vessel Water Level - Low Low, Level 2 signals are initiated from four Ievel 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, Level 2 Allowable Value is chosen such that for complete loss of feedwater flow, the Fleactor Core lsolation Cooling (HCIC) System flowwith HPCS assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Reactor Vessel Water Level - Low Low Low, Level 1. 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 anC{4O-S+for HPCS Applicability Bases. 3,b. Drvwell Pressure - High High pressure in the drywell could indicate a break in the RCPB. The HPCS System and associated DG are initiated upon receipt of the Drywell Pressure - High Function in order to minimize the possibility of fuel damage. The Drywell Pressure - High Function is assumed in the analysis of (continued) PERRY _ UNIT 1 B 3.3-1 01 Revision No. 3

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE 3.c. Heactor Vessel Water Level - High. Level 8 (continued) SAFETY ANALYSES, LCO, measurement instrumentation. The instruments are arranged in a one-and APPLICABILITY out-of-two taken twice logic. This ensures that no single instrument failure can preclude HPCS initiation. The Reactor Vessel Water Level - High, Level B Allowable Value is chosen to isolate flow from the HPCS System prior to water overflowing into the tt/Sls, I Four channels of Reactor Vessel Water Level - High, Level Function are only required to be OPERABLE when HPCS is required to be OPERABLE. Refer to LCO 3.5.1 and{G0=5* for HPCS Applicability Bases. 3.d. Condensate Storaqe Tank Level - Low Low Ievel in the CST indicates the unavailability of an adequate supply of makeup water from this normal source. Normally the suction valve between HPCS and the CST is open and, upon receiving a HPCS initiation signal, water for HPCS injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valve automatically opens, and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the HPCS pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valve must be open before the CST suction valve automatically closes. The Function is implicitly assumed in the accident and transient analyses (which take credit for HPCS) since the analyses assume that the HPCS suction source is the suppression pool. Condensate Storage Tank Level - Low signals are initiated from two Ievel transmitters. The logic is arranged such that either transmitter and associated trip unit can cause the suppression pool suction valve to open and the CST suction valve to close. The Condensate Storage Tank Level

                   - Low Function Allowable Value of 90,300 gallons (elevation 626 ft. I inches) is high enough to ensure adequate pump suction head while water is being taken from the CST.

Two channels of the Condensate Storage Tank Level - Low Function are only required to be OPERABLE when HPCS is required to be OPERABLE to ensure that no single instrument failure can preclude HPCS swap to suppression pool source. Thus, the Function is required to be OPERABLE (continued) PERRY _ UNIT 1 B 3.3-103 Revision No. 3

TS BASES MARK.UP - PROVIDED ECCS Instrumentation FOR INFORMATION ONLY B 3.3.5.1 BASES APPLICABLE 3.d. Condensate Storaqe Tank Level - Low (continued) SAFETY ANALYSES, LCO, in tvlODES 1, 2, and 3. ln MODES I and 5' the Funetien is required te be and APPLICABILITY water level is within the lirnits ef SR 3,5,2,2, With GST water level within eensequenees ef a vessel draindewn event, Refer to LCO 3.5.1 afld t.5*for H PCS Applicability Bases. 3.e. Suppression Pool Water Level - Hioh Excessively high suppression pool water could result in the loads on the suppression pool exceeding design values should there be a blowdown of the reactor vessel pressure through the SiRVs. Therefore, signals indicating high suppression pool water level are used to transfer the suction source of HPCS from the CST to the suppression pool to eliminate the possibility of HPCS continuing to provide additional water from a source outside containment. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valve must be open before the CST suction valve automatically closes. This Function is implicitly assumed in the accident and transient analyses (which take credit for HPCS) since the analyses assume that the HPCS suction source is the suppression pool. Suppression Pool Water Level - High signals are initiated from two level transmitters. The logic is arranged such that either transmitter and associated trip unit can cause the suppression pool suction valve to open and the CST suction valve to close. The Allowable Value for the Suppression Pool Water Level - High Function is chosen to ensure that HPCS will be aligned for suction from the suppression pool before the water level reaches the point at which suppression pool design loads would be exceeded. Two channels of Suppression Pool Water Level - High Function are only required to be OPERABLE in t\4ODES 1,2, and 3 when HPCS is required to be OPERABLE to ensure that no single instrument failure can preclude HPCS swap to suppression pool source. ln tt4ODES 4 and 5, the Function is not required to be OPERABLE since the reactor is depressurized and vessel (continued) PERHY _ UNIT 1 B 3.3-104 Flevision No. I

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE 3.e. Suppression Pool Water Level - Hioh (continued) SAFETY ANALYSES, LCO, blowdown, which could cause the design values of the containment to be and APPLICABILITY exceeded, cannot occur. Refer to LCO 3.5.1 for HPCS Applicability Bases. 3.f. 3.q. HPCS Pump Discharoe Pressure - Hioh (Bvpass) and HPCS Svstem Flow Hate - Low (Bvpass) 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 fully open. The minimum flow 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). The HPCS System Flow Rate - Low (Bypass) and HPCS Pump Discharge Pressure - High (Bypass) 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 is met. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. One flow transmitler is used to detect the HPCS System's flow rate. The logic is arranged such that the transmitter causes the minimum flow valve to open, provided the HPCS pump discharge pressure, sensed by another transmitter, 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) and HPCS Pump Discharge Pressure - High (Bypass) Allowable Value is 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 core. 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. Refer to LCO 3.5.1 and+G0-5* for HPCS Applicability Bases. (continued) PERRY _ UNIT 1 B 3.3-105 Revision No. +

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY ECCS lnstrumentation B 3.3.5.1 BASES APPLICABLE 3.h. fvlanual lnitiation SAFETY ANALYSES, LCO, The lvlanual lnitiation push button channel introduces a signal into the and APPLICABILITY HPCS logic to provide manual initiation capability and is redundant to the (continued) automatic protective instrumentation. There is one push button for the HPCS System. The [/anual lnitiation consists of a single channel in a single trip system. This Function is not considered to be inoperable with indicated reaclor vessel water Ievel on the wide range instrument greater than the Level I setpoint coincident with the reactor steam dome pressure < 450 psig since the HPCS System would provide the necessary injection if required (i.e., if the water level reaches the low water level initiation setpoint). The [/anual lnitiation Function is not assumed in any accident or transient analysis in the USAR. However, the Function is retained for overall redundancy and diversity of the HPCS 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 the lvlanual lnitiation Function is only required to be OPERABLE when the HPCS System is required to be OPERABLE. Refer to LCO 3.5.1 end-I4=5+for HPCS Applicability Bases. Automatic Depressurization Svstem 4-a- 5.a. Fleactor Vessel Wa I eve - Low Low Low- Level 1 Low FiPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, ADS receives one of the signals necessary for initiation from this Function. The Reactor Vessel Water Level - Low Low Low, Level 1 is one of the Functions assumed to be OPERABLE and capable of initiating the ADS during the accidents analyzed in Reference 2. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. (continued) PERRY - UNIT 1 B 3.3-106 Revision No. +

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY ECCS lnstrurnentation B 3.3.5.1 BASES ACTIONS 8.1. B.2. and 8.3 (continued) For Required Action 8.1, redundant automatic initiation capability is lost for a feature if either (a) one or more of its Function 1.a channels and one or more of its Function 2.a channels are inoperable and untripped, or (b) one or more of its Function 1 .b channels and one or more of its Function 2.b channels are inoperable and untripped. Since Required Action 8.1 is only applicable if channels supporting both Divisions of a feature are inoperable and untripped, the affected portions of both Divisions of ECCS, DG and AEGT are declared inoperable concurrently (within '1 hour of discovery). For Required Action 8.2, redundant automatic initiation capability is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system. ln this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action 8.3 is not appropriate and the feature(s) associated with the inoperable, untripped channels must be declared inoperable within t hour.

                @RequiredActionB.1andRequiredActionB.21,15g
                @iens_are                    only applicable in tvloDES 1, 2, and 3. ln MODES 4 and 5, the speeifie initiatien time ef the ECS is net assumed ien inc Notes are also provided (Note 3 to Required Action B.1 and Required Action 8.2) 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.

(continued) PERRY _ UNIT 1 B 3.3-1 1 1 Revision No. 7 I

TS BASES MARK.UP - PROVIDED ECCS lnstrumentation FOR INFORMATION ONLY B 3.3.5.1 BASES ACTIONS C.1 and C.2 (continued) Flequired Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function (or in some cases, within the same monitored parameter) result in redundant automatic initiation capability being lost for the feature(s). Required Action C.1 features would be those that are initiated by Functions 1.c, 1.d, 1.e, 2.c, and 2.d (i.e., Iow pressure ECCS). For Functions 1.c and 2.c, redundant automatic initiation capability is Iost if the Function 1.c and Function 2.c channels are inoperable. For Functions 1.d, 1.e, and 2.d, redundant automatic initiation capability is lost if the Function 1.d and 1.e channels and the Function 2.d channels are inoperable. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated Division to be declared inoperable. However, since channels in both Divisions are inoperable, and the Completion Times started concurrently for the channels in both Divisions, this results in the affected portions in both Divisions being concurrently declared inoperable. For Functions 1.c and 2.c, the affected portions of the Division are LPCI A and LPCI B, respectively. For Functions 'l .d, 1 .e, and 2.d, the affected portions of the Division are the low pressure ECCS pumps (Divisions 1 and 2, respectively). In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 'l hour. e initiatien time ef the EGCS is net assumed and the prebability ef a teGA

Tte-Note 2 states that Required Action C.1 is only applicable for Functions 1.c, 1.d, 1 .e,2.c, and 2.d. The Required Action is not applicable to Functions 1.h, 2.f, and 3.h (which also require entry into this Condition if a channel in these Functions is inoperable), since they are the N/anual lnitiation Functions and are not assumed in any accident or (continued) PERRY _ UNIT 1 B 3.3-1 13 Revision No. 4 I

TS BASES MARK.UP - PROVIDED ECCS lnstrumentation FOR INFORMATION ONLY B 3.3.5.1 BASES ACTIONS E.1 and E.2 (continued) channels within the LPCS and LPCI Pump Discharge Flow - Low (Bypass) Functions result in redundant automatic initiation capability being lost for the feature(s). For Required Action E.1, the features would be those that are initiated by Functions 1.f, 1.9, and 2.e (e.9., low pressure ECCS). Fledundant automatic initiation capability is lost if three of the four channels associated with Functions 1.f, 1.9, and 2.e are inoperable. Since each inoperable channel would have Flequired Action E.1 applied separately (refer to ACTIONS Note), each inoperabte channel would only require the affected low pressure ECCS pump to be declared inoperable. However, since channels for more than one Iow pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable. In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is not appropriate and the feature(s) associated with each inoperable channel must be declared inoperable within t hour after discovery of loss of initiation capability for feature(s) in both Divisions. Required Aetien E,:l is enly applieable in MODES 1; 3; and 3, In MODES I and 5; the speeifie initiatien time ef the lew pressure EGGS is ef initiatien eapability fer 7 days is allewed by Required Aetien E,2 during Note is also provided (Ihe-Note 2to Required Action E.1)to delineate that Required Action E.1 is only applicable to low pressure ECCS Functions. Required Action E.1 is not applicable to HPCS Functions 3.f and 3.9 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 Hequired Action E.2. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action E.1, the Completion Time only begins upon discovery that three channels of the Function (Pump (continued) PERRY _ UNIT 1 B 3.3-1 1 6 Revision No. 4 I

TS BASES MARK-UP. PROVIDED nPV Water lnvento FOR INFORMATION ONLY B 3.3.5.2 B 3.3 INSTRUIVENTAT A 3.3.5.2 Reactor P EASES BACKGROUND The RPV nave tne potential It tne water level sh neat is reduceO. wn clad pertoration. above the too of t etevated ctadding Technical Soecifl satetv svstem sett tunctions. LSSS ar specified for a variab setting must be cho tne annormat situa nnatvticat fimit is-initiated to en l ion action that occurs o tne St is ngt excee@ automatic orotection channels must be chosen to be more conservative than the Anatvtica relateO to tne sett actuattv occur. fn are the same as thos and S in LCO 3.3.5.1 tnstrumentation, lsolation nstru mentation. " I Witn tne unit in nlOU to mitigate anv eve RPV water inventor Safetv Limit e.t.l radioactive mater DRAIN TIUE , some oe DRAIN_TlME calcul automaticattv witn equat to tne fnf wne instrumentation PERRY _ UNIT 1 B 3.3-123a Revision No, TBD

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY RPV Water lnventorv B 3.3.5.2 EASES BACKGROUND The pur fcontinuedl srrnnort the reorriremenls nf I CO 5 ? "ReRetor Pressrrre Vessel fRPV\ Water Inventorv Control." and the "definition of DRAIN TIME. There are tunctions tnat are iniectiontsorav s otner tunctions tn suUsvstem anO neao on low RPV water levd fne nPV Water lnvent tow oressure core s anO high pressure o of these svstems ls nppltcnelE with rh SnfffY lco. to mitioate nrunlvsrs. npv and nPPftCnelLITY raOioactive mater n UouUle-enOeO gul not oostulated in M reduced oioino str postutateO in wnlc Oraining of the RPV with the highest fl penetration flow-seimic event, toss o Uasea on engineeri iniection/sorav subsvstem can be manuallv initiated to maintain adeouate reactor vessel water level. ns OiscusseO in net sttowUFPV water inve fnerefore, nPY Water t o CFR so.s6(c)(z)ffi Permissive and int vatues without reg Satetv nnatvses, l on a function nv fun PERRY _ UNIT 1 B 3.3-123b Revision No. TBD

TS BASES MARK.UP. PROVIDED RPV Water lnventor FOR INFORMATION ONLY B 3.3.5.2 BASES nPPtlCneLE Low Pre SAFEry nrunlvsrs, lco, t.a and APPLICABILITY (continueOt Low re pressure ECCS suUs iniection valves o pressure nas talte pressure. Wnite it vessel oressure will be below the ECCS maximum desiqn oressure, the Reactor Vessel Pre capaUe ot permitt fne neactor Vessel are-initiated from on svstem that senses The Allowable Valu equipment in the lo One cnannel ot neac Permissivel Funct reqtlred to be OPERABL support tne manual required when the asso OPERABLE bv LCO s.s.Z t .b. t .O. e.b. Low P Sprav Pumo Oiscnar fne minimum tlow in Dressure ECCS oumo from overheatino when the oumo is ooeratino and the associated inl is opened when low f wnen tne ttow rate l One tlow transmitt subsvstems' flow r eauscgits assogiated ml Ine minimum tlow va mhimum flow val tor A seconds after low. fne time Oela Ouring the startu itrg mode {for RHR A and RHR B). PEFIRY _ UNIT 1 B 3.3-123c Revision No. TBD

TS BASES MARK-UP - PROVIDED RPV Water lnventor FOR INFORMATION ONLY B 3.3.5.2 EASE nppllcnelE 1.b, 1. SnFEfY Sprav Pumo D ANALYSES. LCO. anU nPPtlCnSILlTY enough to ensure th Wt tow enough to en is initiated to allow fach channel of Pum IPCS cnannel anO tn OPERABLE in I\XODTS + Pumgisrcquired to be OPEBA are capabte of inie inilialed. t.e. e.c. lt/anual ln fne Uanuat Initiat appropriate ECCS l reOunOant to tne au nutton tor eacn ot t ECCS, I-PCS anO fPCl The l\tlanual lnitiat analvses in tne USn reOunOancv anO Oiv nvtne runC in tne pl@is-Inere is no nttowan mechanicallv actu Each channel of the is onlv reouired to be OPERABLE when the associated ECCS is reouired to be OPERABLE bv LCO 3.5.2. Hlgh Pressure Core g.a. ConOensate St fow tevet in tne CSl makeuo water from this normal source. Normallv the suction valve Uetween HPCS anO tn initiation signal However, it tne wat tne suppression po Innntinrred\ PERRY _ UNIT 1 B 3.3-123d Revision No. TBD

TS BASES MARK-UP. PROVIDED nPV Water lnventor FOR INFORMATION ONLY B 3.3.5.2 BASES nPPftCnSLE 3.a. Co SAEEIY nrunrYsrs. lco. csl anO nPPLlCnelLffi losing suction to t supprcssjon pool suction im valve automatica!ffi ConOensate Storag transmitters. fne associated trip unit ca anA tne CSf suction

                   -  Low Function Allo 0e0 ft. S inches) is wnite water is nein fwo cnannels ot tne ontv requireO to Ue OPf nnSLE to ensure HPCS swao to tne sup be OPEHABLE onlv wh tne requirements o wa]et levet is not wl water teyLis withl limlts, a sutticie consequences of a v S.b, S.c. HPCS Pump Svstem Flow Rate - L fne   minimum tto=w ins from overheating w vatve is not tullv o and high pump disch automaticattv ctos or the discharge pr operalinst One tlow transmitt togic is arranged s to open, proviOeO t anotner transmitt fhe togic will clos exceeOeO. fine val pressure decreasl PERRY _ UNIT 1                         B 3.3-123e                 Revision No. TBD

TS BASES MARK-UP. PROVIDED FOR INFORMATION ONLY nPV Water lnventorv B 3,3.5.2 EASES APPLICABLE S.b. 3. SAfEfY Svstem flow ANALYSES, LCO. anO nPPtlCnelLlTY Discharge Pressur ensure tnat oump tl enough to ensure th allow full flow int (Bvpass) Allowable Value is set hioh enouoh to ensure that the valve will not be ooen when the oumo is not ooeratino. One channet ot eacb f is-rcquired to be OPERAB RHR Svstem Isolatlm 4.a. Reactor Vessel Water Level - Low. Level 3 fne detinition ot n ttow oathg tnat are level isotation in Ihe TAL The Reactor Vessel to be OPEnneLE when penetration ttow o* The Reactor Vessel RHn Snutdown Cool in-ary transient or accid laroe breaks such as MSLBS. The RHR Shutdown Coolino Svstem isolation on Level Ooes not droo below event through the l (e.g., pipe break o Cooting Svstem. Reactor Vessel Wat levet transmitter constant cotumn ot actual water tevel cnannels per trip tevel S Functiongre a Water l-evet - t-ow. I IUODES 4 and S (both c (continued) PERRY _ UNIT 1 B 3.3-123f Revision No. TBD

TS BASES MARK.UP. PROVIDED RPV Water lnventor FOR INFORMATION ONLY B 3.3.5.2 BASS nPPLtCneLE 4.a. Re SAFEry nfUntVSES, tCO. fne anO nPPttCnelLlT l-evel S Allowable V tuel mav be tnreate fnis function isol neactor Water Ctean 5.a. Reactor Vessel Water Level - Low Low, Level 2 Tne Oetinition ot O tlow oatns tnat are tevet isotation in the fnf. fne neacto assoeiated with RWCU Sv isotation ot penet fow nPV water level threatened. Shoul resutt. fneretore isolate the ootential sources of a break. neactor Vesset Wate tourlevel tr3nsmitt to a constant cotum actual water level cnannets per trio s Level e function ar svstem) are reouired to be OPERABLE. The neactor Vessel chosen to be the sam t-ow, l-evet Z nltowa the tuet mav be thre fne neactor Vessel resdred to be OPERABL penetration ttow pat fnis Function isol fcontinrredl PERRY _ UNIT 1 B 3.3-1239 Revision No. TBD

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY RPV Water lnventorv B 3.3.5.2 BASES (continued) nCf tOruS n f'lote nas O Inventorv Control fimes, specities t Oivisions, suUsvs ConOition Oiscove separate entrv int Required Actions c Completion fimes n Reouired Actions for inoperable RPV Water !nventorv Control instrumentation c separate inqoerab Inventorv Control AJ Reouired Action A.1 directs entrv into the aoprooriate Condition retergnced in faOl Table is Function d inoperaUte, ConOl transter to tne ap ion 8.1 and B.2 RHH Svstem tsolatio neactor Water Ctean I-evet e tunctions a associateO penetratio the instrumentatio immeUiate Oeclara incaoable of automatic isolation. Reouired Action 8.2 directs calculation ot =nnntN ftHllf . fne o affected penetrat (eontinrredl PERRY _ UNIT 1 B 3.3-123h Revision No. TBD

TS BASES MARK-UP. PROVIDED nPV Water lnventor FOR INFORMATION ONLY B 3.3.5.2 BASES ACTIONS CI

 -Jesxtrnuedl Low reactor steam d tne nw pressure f                                         iu tunctions. lt tnis proniUited. f ner                                                ion within l hour. With mav Ue pertormeO. P condition. the ooe iniection valve to fne   Comptetion   fim evaluate anv   Oisc                                            ig DlandD2 neouired nctbns n actions are taken    l function result in from tne conOensat svstem must Ue Oecl suctlon must Ue ali tunction,iq alrea fne t   nour Completl ot HPCS Oeing need                                              itrg time for restoratl zuppre.sgion-paaL E.1 It an tPCI or LPCS Ol Svstem Oischarge P lnoperanle, there wnen tne pump is ope open. ln tnis conOl and the iniection v manuat initiation ne starteO manuatl not tne preterreO c fne e+ nour Complet to evatuate and re                                           ian fime is appropriat and open the iniect overheat.

PEFIRY _ UNIT 1 B 3.3-123i Revision No. TBD

TS BASES MARK-UP. PROVIDED FOR INFORMATION ONLY nPV Water lnventor B 3.3.5.2 BASES ACTIONIS = , FJ JcsnIEUed) Witn tne nequireO n f C. O. or not met. tn be incapable of per inoperable immedl SURVEILLnNCE As no REQUIREIttIENTS Inve Table 3.3.5.2-1 . sR 3.3.s.2.1 Pertormance ot tne instrumentation ha compzuison of tne parame parameter on otner instrument cnanne apptulmatetv tne same v instrument channe one ot the chgnflels o CHECK ouarantees that undetected outrioht channel failure is limited; thus, it is kev to ve properlv Uetween e Agreement criterl combination of th ion and readabilitv. I that the instrument has drifted outside its limit. fne Surveittance f frequencv Control The CHANNEL CHECX S checks of channels assoclateO wittt tne cna sE-3-?.5.22 n cunr.lnrl rurucrtoru to ensure that the e successtul test ot PERRY _ UNIT 1 B 3.3-123j Revision No, TBD

TS BASES MARK.UP . PROVIDED nPV Water lnventorv FOR INFORMATION ONLY B 3.3.5.2 BASES SURVEILLANCE SH S. HEOUIREIUENTS pertormeO bv tne ve tne retav. fnis cla TEST of a relav. Thi contacts ot tne rel non-fecnnical Spe Anv setooint adius current plant spec The Surveillance F Frequencv Control sR 3.3.5.2.3 The LOGIC SYSTEIV FU OPf nnelLlTY of the svstem functiona js Surveillance to c ign-fEg Eufv=eillance Fre Freouencv Control ne ff nf IUCES t . lnfo tnventorv in goili November 1984. Z. lnformation Not gecause of Nlisalig S. Generic Letter g Vessel Water Level I SO.S+(t)." nugus

                +. NRC Bulletin 9S Water Levet lnstru S. tntormation tUot neactor Vesset nra PERRY _ UNIT 1                         B 3.3-123k                  Revision No. TBD

TS BASES MARK.UP . PROVIDED RCIC System lnstrumentation FOR INFORMATION ONLY B 3.3.5.32 B 3.3 INSTRUIUENTATION B 3.3.5.32 Heactor Core lsolation Cooling (RCIC) System lnstrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is unavailable, such that initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LCO 3.5.3, "FlClC System." The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low, Level 2. The variable is monitored by four transmitters that are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement. Once initiated, the RCIC logic seals in and can be reset by the operator only when the reactor vessel water level signals have cleared. The RCIC CST first and second test return valves close on a RCIC initiation signal to allow full system flow. The HCIC System also monitors the water levels in the condensate storage tank (CST) and the suppression pool, since these are the two sources of water for RCIC operation, Reactor grade water in the CST is the normal source. However, only the capability to take suction from the suppression pool is required for OPEFIABILITY. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically signaled to open (it is normally in the open position) unless the pump suction from the suppression pool valve is open. lf the water level in the CST falls below a preselected level, first the suppression pool suction valve automatically opens and then the CST suction valve automatically closes. Two level transmitters are used to detect low waler level in the CST. Either transmitter and associated trip unit can cause the suppression pool suction valve to open and the CST suction valve to close. Similarly, two level transmitters are used to detect high water level in the suppression pool. The suppression pool suction valve also automatically opens and the CST suction valve closes if high water level is detected in the suppression pool. To prevent losing suction to the pump, (continued) PERRY _ UNIT 1 B 3.3-124 Revision No. 5

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY RCIC System lnstrumentation B 3.3.5.3e BASES BACKGROUND the suction valves are interlocked so that one suction path must be open (continued) 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 B) trip (two-out-of-two logic), at which time the RCIC steam supply valve closes (the injection valve also closes due to the closure of the steam supply valve). The HCIC System restarts if vessel level again drops to the low level initiation point (Level 2). APPLICABLE The function of the RCIC System is to provide makeup coolant to the SAFETY reactor in response to transient events. The RCIC System is not an ANALYSES, LCO, Engineered Safety Feature Systern and no eredit is taken in the safety and APPLICABILITY analysis for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the RCIC System, and therefore its instrumentation, are included as required by the NFIC Final Policy Statement on Technical Specification Improvements (58 FR 39132). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the RCIC System instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.92-1. Each Function must have a required number of OPEFIABLE channels with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for each RCIC System instrumentation Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified accounts for instrument uncertainties appropriate to the Function. These uncertainties are described in the setpoint methodology. (continued) PERRY _ UNIT 1 B 3.3- 1 25 Revision No. 1

TS BASES MARK.UP. PROVIDED RCIC System lnstrumentation FOR INFORMATION ONLY B 3.3.5.32 BASES APPLICABLE The individual Functions are required to be OPERABLE in fUODE 1 , and SAFETY in IIIODES 2 and 3 with reactor steam dome pressure > 150 psig, since ANALYSES, LCO, this is when RCIC is required to be OPERABLE. (Flefer to LCO 3.5.3 for and APPLICABILITY Applicability Bases for the RCIC System.) (continued) The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis,

1. Reactor Vessel Water Level - Low Low, Level 2 Low reactor pressure vessel (RPV) water level indicates that 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 could result. Therefore, the RCIC System is initiated at Level 2 to assist in maintaining waler level above the top of the active fuel.

Reactor Vessel Water Level - Low Low, Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. 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 required to be OPERABLE when RCIC is required to be OPEFIABLE 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 - Hioh. Level B High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel.

Therefore, the Level 8 signal is used to close the RCIC steam supply valve to prevent overflow (continued) PERRY _ UNIT 1 B 3,3-126 Revision No. g I

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY RCIC System lnstrumentation B 3.3.5.e2 BASES APPLICABLE 2. Reactor Vessel Vllater Level - Hiqh. Level B (continued) SAFETY ANALYSES, LCO, into the main steam lines (tt/lSls). (The injection valve also closes due to and APPLICABILITY the closure of the steam supply valve.) Fleactor Vessel Water Level - High, Level 8 signals for FICIC are initiated from four level transmitters from the wide range water Ievel measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Water Level - High, Level I 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 tvlSls. Four channels of Reactor Vessel Water Level - High, Level 8 Function are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.

3. Condensate Storaoe Tank Level - Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this norrnal source. Normally the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valve automatically opens and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valve must be open before the CST suction valve automatically closes.

Condensate Storage Tank Level-Low signals are initiated from two level transmitters. The logic is arranged such that either transmitter and associated trip unit can cause the suppression pool suction valve to open and the CST suction valve to close. The Condensate Storage Tank Level-Low Function Allowable Value of 90,300 gallons (elevation 626 ft. I inches) is high enough to ensure adequate pump suction head while water is being taken from the CST. (continued) PEFIRY _ UNIT 1 B 3.3-127 Revision No. 3

TS BASES MARK-UP - PROVIDED FOR INFORMATION ONLY RCIC System lnstrumentation B 3.3.5,32 BASES APPLICABLE 3. Condensate Storaqe Tank Level - Low (continued) SAFETY ANALYSES, LCO, Two channels of Condensate Storage Tank Level - Low Function are and APPLICABILITY required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to the suppression pool source. Refer to LCO 3.5.3 for RCIC Applicability Bases.

4. Suppression Pool Water Level - Hish Excessively high suppression pool water level could result in the loads on the suppression pool exceeding design values should there be a blowdown of the reactor vessel pressure through the safety/relief valves.

Therefore, signals indicating high suppression pool water level are used to transfer the suction source of RCIC from the CST to the suppression pool to eliminate the possibility of RCIC continuing to provide additional water from a source outside primary containment. This Function satisfies Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132). To prevent losing suction to the pump, the suction valves are interlocked so that the suppression poo! suction valve must be open before the CST suction valve automatically closes. Suppression Pool Water Level - High signals are initiated from two Ievel transmitters. The logic is arranged such that either transmitter and associated trip unit can cause the suppression pool suction valve to open and the CST suction valve to close. The Allowable Value for the Suppression Pool Water Level - High Function is chosen to ensure that RCIC will be aligned for suction from the suppression pool before the water level reaches the point at which suppression pool design loads would be exceeded. Two channels of Suppression Pool Water Level - High Function are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to the suppression pool source. lf the automatic transfer of the suction source for RCIC from the CST to the suppression pool, due to a high suppression pool water level signal, is manually overridden by the operator, then the Suppression Pool Water Level-High Functions are considered inoperable. Refer to LCO 3.5.3 for RCIC Applicability Bases. (continued) PERRY _ UNIT 1 B 3.3-128 Revision No. +

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY RCIC System lnstrumentation B 3.3.5.32 BASES APPLICABLE 5. lt/anual lnitiation SAFETY ANALYSES, LCO, The Manual lnitiation push button switch introduces a signal into the and APPLICABILITY RCIC System initiation logic that is redundant to the automatic protective (continued) instrumentation and provides manual initiation capability. There is one push button for the RCIC System. The lt/anual lnitiation Function is not assumed in any accident or transient analyses in the USAR. However, the Function is retained for the FICIC function as required by the NFtC 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 lnitiation is required to be OPERABLE when RCIC is required to be OPERABLE. ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels, Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel. 4.1 Hequired Action A.'l directs entry into the appropriate Condition referenced in Table 3.3.5.92-'l in the accompanying 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) PERRY _ UNIT 1 B 3.3-129 Revision No. g

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY RCIC System Instrumentation B 3.3.5.32 BASES ACTIONS 8.1 and B.2 (continued) Required Action 8.1 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. ln this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. ln this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action 8.2 is not appropriate, and the RCIC System must be declared inoperable within t hour after discovery of Ioss of RCIC initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action 8.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two inoperable, untripped Reactor Vessel Water Level - Low Low, Level 2 channels in the same trip system. The t hour Completion Time from discovery of Ioss 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 hours has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status. lf 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 8.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.9., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken. (continued) PERRY _ UNIT 1 B 3.3-130 Revision No. g

TS BASES MARK.UP . PROVIDED RCIC System lnstrumentation FOR INFORMATION ONLY B 3.3.5.92 BASES ACTIONS c.1 (continued) A risk based analysis was peformed and determined that an allowable out of service time of 24 hours (Ref. 1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action 8.1), limiting the allowable out of service time if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level - High, Level I Function. This Condition also applies to the l/anual lnitiation 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 hours is allowed. The Required Action does not allow placing a channel in trip since this action would not necessarily result in the safe state for the channel in all events. D.1. D.2.1. and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if multiple inoperable, untripped channels within the same Function result in automatic component initiation capability being lost for the feature(s). For Required Action D.1, the RCIC System is the only associated feature. In this case, automatic component initiation capability is lost if two Function 3 channels or two Function 4 channels are inoperable and untripped. ln this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is not appropriate, and the RCIC System must be declared inoperable within t hour from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppres$ion pool since, if aligned, the Function is already pedormed. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action D.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The t hour (continued) BASES PERRY _ UNIT 1 B 3.3-131 Revision No. g

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY RCIC System lnstrumentation B 3.3.5.32 ACTlONS D.1 , D.2.1, and D.2.2 (continued) 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 hours has been shown to be acceptable (Ref . 1) to permit restoration of any inoperable channel to OPERABLE status. lf the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1, which performs the intended function of the channel (shifting the suction source to the suppression pool). Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. lf Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. lf it is not desired to perform Flequired Actions D.2.1 and D.2.2 Condition E must be entered and its Required Action taken. E.1 With any Required Action and associated Completion Time of Condition B, C, or D not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately. SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC System REQUIRETUENTS instrumentation Function are found in the SRs column of Table 3.3.5.32-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Flequired Actions may be delayed as follows: (a) for up to 6 hours for Function 5; and (b) for up to 6 hours for Functions 'l , 2, 3, and 4 provided the associated (continued) PERRY _ UNIT 1 B 3.3-132 Revision No. g

TS BASES MARK.UP - PROVIDED RCIC System Instrumentation FOR INFORMATION ONLY B 3.3.5.32 BASES SURVEILLANCE Function maintains trip capability. Upon completion of the Surveillance, REQUIRETVIENTS or expiration of the 6 hour allowance, the channel must be returned to (continued) OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 1) assumption of the average time required to perform channel Surveillance That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary. sR s.s.5.32.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. lt 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 that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. lf a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with the channels required by the LCO. sR 3.3.5.82 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) PERRY - UNIT 1 B 3.3-133 Revision No. 11

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY RCIC System I nstrumentation B 3.3.5.32 BASES SURVEILLANCE SR 3.3.5.82.2 (continued) REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. sR 3.3.5.3e.3 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.Q+1. lf the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be re-adjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.3.5.32.4 and SR 3.3.5.8.6 CHANNEL CALIBRATION is a complete check of the instrument Ioop and the sensor. This test verifies the channe! responds to the measured parameter with the necessary range and accuracy. CHANNEL CALIBRATION Ieaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. sR 3.3.5.32.5 The LOGIC SYSTET\4 FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function. (continued) PERRY _ UNIT 1 B 3.3-134 Revision No. 44

TS BASES MARK-UP - PROVIDED FOR INFORMATION ONLY HCIC System lnstrumentation B 3.3.5.32 BASES SUHVEILLANCE SR 3.3.5.92.5 (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. REFERENCES 1 GENE-770-06-2, "Addendum to Bases for Changes to Surveillance Test lntervals and Allowed Out-of-Service Times for Selected lnstrumentation Technical Specifications," February 1 991 . PEHRY _ UNIT 1 B 3.3-135 Revision No. {4

TS BASES MARK-UP . PROVIDED FOR INFORMATION ONLY Primary Containment and Drywell Isolation lnstrumentation B 3.3.6.1 BASES APPLICABLE specified Allowable Values, where appropriate. A channel is inoperable SAFETY if its actual trip setpoint is not within its required AIIowable Value. The ANALYSES, LCO, actual setpoint is calibrated consistent with applicable setpoint and APPLICABILITY methodology assumptions. Each channel must also respond within its (continued) assumed response time, where appropriate. Allowable Values are specified for each Primary Containment and Drywell lsolation Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. 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.9., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained frorn the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.9., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. 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 lnstrumentation," and LCO 3.3.5.32, "RCIC lnstrumentation," and are not included in this LCO. ln 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-Operating" or LCO 3.6.5.1, "Drywell," as (continued) PERRY _ UNIT 1 B 3.3-140 Revision No. O I

TS BASES MARK.UP . PROVIDED Primary Containment and Drywell Isolation lnstrumentation FOR INFORMATION ONLY B 3.3.6.1 BASES APPLICABLE 2,a. 2.e. Reactor Vessel Water Level-Low Low, Level 2 SAFETY (continued) ANALYSES, LCO, and APPLICABILITY since isolation of these valves is not critical to orderly plant shutdown This Funetien is required te be OPERABtE during eperatiens with a eFiselatingFpetential seurees ef Ieakage must be previded te ensure that is Funetien is net required te be OPERABTE= This Function isolates the 1E22-F023 Valve (Function 2.e\, and the Group 1, 5,7, and I valves (Function 2.a). 2.b, 2.d, 2.f Drvwell Pressure-Hiqh High drywell pressure can indicate a break in the RCPB. The isolation of some of the PCIVs on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 50.67 are not exceeded (Ref. 8). The Drywell Pressure-High Function associated with isolation of the primary containment is implicitly assumed in the USAR accident analysis as these Ieakage paths are assumed to be isolated post LOCA. ln addition, Functions 2.b and 2.d provide isolation signals to certain drywell isolation valves. The isolation of drywell isolation valves, in combination with other accident mitigation systems, functions to ensure that steam and water releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the drywell. High drywell pressure signals are initiated from four pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High per Function are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. Function 2.f (Division 3) has only one trip system consisting of four channels logically combined in a one-out-of-two twice configuration. The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High AIlowable Value (LCO 3.3.5.1), since (continued) PERRY - UNIT 1 B 3.3-146 Revision No. {S

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY Primary Containment and Drywell lsolation Instrumentation B 3.3.6.1 BASES APPLICABLE 2.c. Reactor Vessel Water Level-Low Low Low, Level 1 (continued) SAFETY ANALYSES, LCO, This Funetien is required te be OPERABTE during eperatiens with a and APPLICABILITY r vessel (OPDRVs) beeause the eapability et isetating pet effsite dese limits are net exeeeded if eere damage eeeurs, Hewever, OPDRVs assume that ene er mere fuel assemblies are leaded inte the eere, Therefere; if the fuel is fully eff Ieaded frem the reaeter vessel, this This Function isolates the Group 2 isolation valves. 2.9. Containment and Drvwell Puroe Exhaust-Plenum Radiation-Hiqh High purge exhaust plenum radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the HCPB. When Purge Exhaust-Plenum Ftadiation-High is detected, valves whose penetrations communicate with the primary containment atmosphere are isolated to limit the release of fission products. ln addition, this Function provides an isolation signal to certain drywell isolation valves. The isolation of drywell isolation valves, in combination with other accident mitigation systems, functions to ensure that steam and water releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the drywell. The Purge Exhaust-Plenum Radiation-High signals are initiated from four radiation detectors that are located on the purge exhaust plenum ductwork coming from the drywell and containment. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. (continued) PERRY _ UNIT 1 B 3.3- 1 48 Revision No. 7

TS BASES MARK.UP - PROVIDED Primary Containment and Drywell lsolation Instrumentation FOR INFORMATION ONLY B 3.3.6.1 BASES APPLICABLE 2.s Containment and Drvwell Purse Exhaust-Plenum Radiation-Hiqh SAFETY (continued) ANALYSES, LCO, and APPLICABILITY Four channels of Containment and Drywell Purge Exhaust-Plenum Radiation High Function are required to be OPERABLE to ensure that no single instrument failure Gan preclude the isolation function. Containment and Drywell Purge System inboard and outboard isolation valves each use a separate two-out-of-two isolation logic. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding and to ensure offsite doses remain below 10 CFR 20 and 10 CFR 50.67 limits. The Funetien is required te be OPERABtE during erer#isns with a must be previded te ensure effsite dese limits are net exeeeded, Hewever; OPDRVS assume that ene er mere fuel assemblies are leaded inte tne eere, fner vessel, this Funetien is net required te be OPERABtE, Due to radioactive decay, handling of fuel only requires OPERABILITY of this Function when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 12). These Functions isolate the Group I valves. 2.h. Manual lnitiation The Manual lnitiation push button channels introduce signals into the primary containment and drywell isolation logic that (continued) PERRY - UNIT 1 B 3.3-149 Revision No. 1S I

TS BASES MARK-UP . PROVIDED Primary Containment and Drywell Isolation lnstrumentation FOR INFORMATION ONLY B 3.3.6.1 BASES APPLICABLE 2.h. Manual lnitiation (continued) SAFETY ANALYSES, LCO, are redundant to the automatic protective instrumentation and provide and APPLICABILITY manual isolation capability. There is no specific USAR safety analysis that takes credit for this Function. lt is retained for the isolation function as required by the NRC in the plant licensing basis. There are four push buttons for the logic, two manual initiation push buttons per trip system. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. Four channels of the Manual lnitiation Function are required to be OPERABLE in MODES 1, 2, and 3, and during PDRVs++-movement of recently irradiated fuel assemblies in primary containment, since these are the MODES in which the Primary Containment and Drywell lsolation automaticFunctionSarerequiredtobeoPERABLE.fu that ene er mere fuel assemblies are loaded into the eere, Therefere; if reg{*i Due to radioactive decay, handling of fuel only requires OPERABILITY of this Function when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 12). The manual initiation channels for the RCIC System is discussed in Section 3.k below, and for the HPCS System is discussed in the Bases description for ECCS lnstrumentation (LCO 3.3.5.1). (continued) PERRY _ UNIT 1 B 3.3-150 Revision No. 4 I

TS BASES MARK.UP . PROVIDED Primary Containment and Drywell lsolation lnstrumentation FOR INFORMATION ONLY B 3.3.6.1 BASES APPLICABLE 5.b. Reactor Vessel Water Level-Low. Level 3 (continued) SAFETY ANALYSES, LCO, Reactor Vessel Water Level - Low, Level 3 signals are initiated from four and APPLICABILITY level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable Ieg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water Level - Low, Level 3 Function are required to be OPERABLE to ensure that no single instrumentfailurecanprecludetheisolationfunction,@

                       '1 and 5 (beth ehannels must inBut inte the same trip system) previded the
                       +s+ai                                                                   ing pertermea tnat nas th+ys'te+

The Reactor Vessel Water Level - Low, Level 3 Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level - Low, Level 3 Allowable Value (LCO 3.3.1.1) since the capability to cool the fuel may be threatened. The Reactor Vessel Water Level - Low, Level 3 Function is required to be OPERABLE in IIODES 2 and 3 when the reactor vessel steam dome preSSUreislessthantheRHRcutinpermissiVepreSSureM I and 5 to prevent this potential flow path from lowering reactor vessel Ievel to the top of the fuel. ln tt/ODE 1 , and lvlODES 2 and 3 when the reactor vessel steam dome pressure is above the RHR cut in permissive pressure, the Reactor Steam Dome Pressure-High Function will maintain the RHR System isolated. This Function isolates the Group 3 and 4 valves. 5.c. Reactor Vessel Steam Dome Pressure-Hioh The Reactor Vessel Steam Dome Pressure-High Function is provided to isolate the shutdown cooling portion of the HHR System. This interlock (RHR cut in permissive) is provided only for equipment protection to prevent an intersystem LOCA scenario and credit for the interlock is not assumed in the accident or transient analysis in the USAR. (continued) PERRY _ UNIT 1 B 3.3-161 Revision No. 4

TS BASES MARK.UP - PROVIDED Primary Containment and Drywell lsolation lnstrumentation FOR INFORMATION ONLY B 3.3.6.1 BASES ACTIONS J.1 (continued) lf the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be isolated (e.9., close either 1E12-F008 or 1E12-F009). However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status. Actions must cont OPf nnelE status. eaBability and subsequently iselate the RHR Shutdewn Geeling System; is

iselatien valve and asseeiated instrumentatien are OPERAETE er ether i ine if the emenents a i OPEFIAEItITY ef the eempenents, lf hewever' any required eempenent is-+neeerabte; ttre ' the Surveillanees may need te be perfermed te restere the eempenent te SPERABtE status= ln additien; at least ene deer in eaeh primary eentainrnent air leeh must ive een+rels hewever' the elesed deer ean be epened intermittently fer entry weuld net be expeeted te result in the immediate release ef appreeiable fissien preduets te the eentainment atmesphere, Aetiens must eentinue (continued) PERRY _ UNIT 1 B 3.3-168 Revision No. 4 I

TS BASES MARK.UP . PROVIDED Primary Containment and Drywell lsolation lnstrumentation FOR INFORMATION ONLY B 3.3.6.1 BASES ACTIONS K.1_an-d, K.zJ-anCJ6.E2 (continued) If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path(s) should be isolated (Required Action K.1). Isolating the affected penetration flow path(s) accomplishes the safety function of the inoperable instrumentation. Alternately, the plant must be placed in a condition in which the LCO does not apply. If applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe condition Als# fer fissien preduetien release= Aetiens must eentinue until ePDRVs are suspende* tJ fer fissien preduet release= Aetiens must eentinue until OPDRVs are suspenge+ SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Primary REQUIREMENTS Containment and Drywell lsolation lnstrumentation Function are found in the SRs column of Table 3.3.6.1-1. The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains primary containment isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. (continued) PERRY - UNIT 1 B 3.3-169 Revision No. 2 l

TS BASES MARK-UP . PROVIDED CRER System lnstrumentation FOR INFORMATION ONLY B 3.3.7,1 BASES APPLICABLE 1. Reaetorl(essel Water Level - Low Low Low, Level 1 SAFETY ANALYSES, LCO, Low reactor pressure vessel (HPV) water level indicates that the and APPLICABILITY capability to cool the fuel may be threatened. A low reactor vessel water (continued) level could indicate a LOCA, and will automatically initiate the CRER System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel. Reactor Vessel Water Level - Low Low Low, Level 1 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Fleactor Vessel Water Level - Low Low Low, Level 'l Function (two channels per trip system) are required to be OPERABLE to ensure that no single instrument failure can preclude CRER System initiation. The Allowable Value for the Reactor Vessel Water Level - Low Low Low, Level 1 is chosen to be the same as the Reactor Vessel Water Level - Low Low Low, Level 1 Allowable Value (LCO 3,3.5.1, "ECCS lnstru mentation"). The Reactor Vessel Water Level - Low Low Low, Level 1 Function is required to be OPERABLE in [\4ODES 1, 2, and 3,@iens with a petential fer draining the reaeter vessel (OPDRVs)' to ensure that the control room personnel are protected. Hewever, OPDRVs assume that ene er mere fuel assernblies are Ieaded inte the eere, Therefere, if the fuel is fully eff leaded frem the reaeter vessel' this funetien is ne+ requ4 lntvloDES4and5,ffi the Control Room Ventilation Radiation lt/onitor Function provides adequate protection.

2. Drvwell Pressure - Hiqh High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). A high drywell pressure signal could indicate a LOCA and will automatically initiate the CRER System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

(continued) PERRY _ UNIT 1 B 3.3-204 Revision No. I

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY CRER System lnstrumentation B 3.3.7.1 BASES APPLICABLE 2. Drvwell Pressure - Hiqh (continued) SAFETY ANALYSES, LCO, Drywell Pressure - High signals are initiated from four pressure and APPLICABILITY transmitters that sense drywell pressure. Four channels of Drywel! Pressure - High Function (two channels per trip system) are required to be OPERABLE to ensure that no single instrument failure can preclude CHER System initiation. The Drywell Pressure - High Allowable Value was chosen to be the same as the ECCS Drywell Pressure - High Allowable Value (LCO 3.3.5.1). The Drywell Pressure - High Function is required to be OPEHABLE in IIODES 1, 2, and 3 to ensure that control room personnel are protected during a LOCA. ln ttflODES 4 and 5, the Drywell Pressure - High Function is not required since there is insufficient energy in the reactor to pressurize the drywell to the Drywell Pressure - High setpoint.

3. Control Room Ventilation Radiation [/onitor The Control Room Ventilation Ftadiation Jtlonitor measures radiation levels downstream of the supply plenum discharge of the control room. A high radiation level may pose a threat to control room personnel; thus, the Control Room Ventilation Ftadiation lvlonitor Function will automatically initiate the CREH Systern.

The Control Room Ventilation Radiation [4onitor Function consists of one noble gas monitor. One channel (which provides input to both Trip Systems) of the Control Room Ventilation Radiation Monitor is required to be OPERABLE. Since a LOCA signal will also initiate the CRER System isolating the control room from the environment, and considering the fact that a LOCA signal itself incorporates sufficient redundancy, the airborne radiation monitor signal is considered a diverse signal, and does not require redundancy. The Allowable Value was selected to ensure protection of the control room personnel. The Control Room Ventilation Radiation hlonitor Function is required to be OPEFIABLE in IVIODES 1,2, and 3, and during lPDEI/s+nd movement of recently irradiated fuel in the primary containment or Fuel Handling Building to ensure (continued) PEFIRY _ UNIT 1 B 3.3-205 Revision No. 2 I

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY CRER System lnstrumentation B 3.3.7.1 BASES APPLICABLE 3. Control Room Ventilation Radiation Monitors (continued) SAFETY ANALYSES, LCO, that control room personnel are protected during a LOCA_orrafuel and APPLICABILITY handling event involving recently irradiated fuel; er a vessel draindewn event. Due to radioactive decay, handling of fuel only requires OPERABILITY of this Function when the fuel being handled is recently irradiated, i.8., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handting before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 9). fuel assemblies are leaded inte the eere, Therefere; if the fuel is fully eff eIPEEAEtE=-During MODES 4 and 5, when these specified conditions (i.e.. movement of recentlv irradiated fuel assemblies) is ar+-not in progreSS@,theprobabilityofaLoCAorsignificantfueJ damage is low; thus, the Function is not required. ACTIONS A Note has been provided to modify the ACTIONS related to CRER System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CRER System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CRER System instrumentation channel. 4.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.7.1-1. The applicable Condition specified in the Table is Function dependent. (continued) PERRY _ UNIT 1 B 3.3-206 Revision No. 4

TS BASES MARK.UP - PROVIDED ECCS - Operating FOR INFORMATION ONLY B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CUIEAL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network is composed of the High Pressure Core Spray (HPCS) System, the Low Pressure Core Spray (LPCS) System, and the Iow pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System. The ECCS also consists of the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS. Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for the HPCS System. On receipt of an initiation signal, each associated ECCS pump automatically stafts; simultaneously the system aligns, and the pump injects water, taken either from the CST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed by a timer, allowing the operator to interrupt the timed sequence if the system is not needed. The HPCS pump discharge pressure almost immediately exceeds that of the RCS, and the pump injects coolant into the spray $parger above the core. lf the break is small, HPCS will maintain coolant inventory, as well as vessel Ievel, while the RCS is still pressurized. lf HPCS fails to maintain water level above Level 1, it is backed up by automatic initiation of ADS in combination with LPCI and LPCS. In this event, the ADS would time out and open the selected safety/relief valves (S/RVs), depressurizing the RCS and allowing the LPCI and LPCS to overcome RCS pressure and inject coolant into the vessel. lf the break is large, RCS pressure initially drops rapidly, allowing the LPCI and LPCS systems to inject coolant into the core. Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool may be circulated through two heat exchangers in series cooled by the Emergency Service Water (ESW) System. Depending on the Iocation and size of the break, portions of the ECCS may be ineffective; however (continued) PERRY _ UNIT 1 B 3.5-1 Revision No. g

TS BASES MARK-UP - PROVIDED ECCS - Operating FOR INFORMATION ONLY B 3.5.1 BASES (continued) LCO Each ECCS injection/spray subsystem and eight ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are the three LPCI subsystems, the LPCS System, and the HPCS System. The ECCS injection/spray subsystems are fufther subdivided into the following groups: a) The Iow pressure ECCS injection/spray subsystems are the LPCS System and the three LPCI subsystems; b) The ECCS injection subsystems are the three LPCI subsystems; and c) The ECCS spray subsystems are the HPCS System and the LPCS System. With less than the required number of ECCS subsystems OPERABLE during a Iimiting design basis LOCA concurrent with the worst case single failure, the Iimits specified in 10 CFR 50.46 (Ref. 10) could potentially be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by 10 CFR 50.46 (Ref. 10). LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not othenrvise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary. APPLICABILITY All ECCS subsystems are required to be OPERABLE during I4ODES 1, 2, and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the reactor coolant pressure boundary. In MODES 2 and 3, the ADS function is not required when pressure is < 150 psig because the low pressure ECCS subsystems (LPCS and LPCI) are capable of providing flow into the RPV below this pressure. ECCS requirements for MODES 4 and S are specified in LCO 3.5.2, Vessel (nPV) Water (continued) PERRY - UNIT 1 B 3.5-5 Revision No. O I

TS BASES MARK.UP - PROVIDED ECCS - Operating FOR INFORMATION ONLY B 3.5.1 BASES SURVEILLANCE SR 3.5.1 .2 (continued) REQUIREMENTS capable of being manually realigned (remote or local) to the LPCI mode and not othenrvise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3 if necessary. sR 3.5.1 .3 Verification that ADS accumulator supply pressure is < 150 psig assures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The designed pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 7A% of design pressure (Ref. 13). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of 150 psig is provided by the Safety Related Instrument Air System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. sR 3.5.1.4 The performance requirements of the ECCS pumps are determined through application of the 10 CFR 50, Appendix K, criteria (Ref. B). This periodic Surveillance is performed (in accordance with the ASME Code, Section Xl, requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of 10 CFR 50.46 (Ref. 10). The differential pressure for each listed system in the Surveillance Requirements (SRs) 3.5.1.4+nd45*5, is the difference between the containment wetwell pressure and the RPV pressure assumed in the LOCA analyses at the time of injection/spray. In addition to this listed differential pressure, the ECCS pumps also need to overcome (conli NU ed\ PERRY - UNIT 1 B 3.5-10 Revision No. {4 I

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY ffi B 3.5.2 rsot^roN coorrNc (RGrc) SYSTEM B 3,5,2 ECCS Shutdewn BASES BACKGROUND Adeseriptien ef the High Pressure Cere Spray (HPGS) System; tew D.^^^,.va /n^r^ C^ra., /l trlr^C\ Crra*am anrJ la,*r 6rA^^r,ra aaalan* iniaa{lan (tPCl) mede ef the Residual Heat Ftemeval (RHFI) System is previded in is vessel water level, Te previde redundaney; a minimum ef twe ECCS The EGCS satisfy eriterien 3 ef the NRG final Peliey Statement en Teehnieal Speeifieatien Imprevements (58 FFI 39132)= The ECGS inieetien/sBray subsystems are defined as the three tPG! in tPeS System and the three tPe I subsystems; afld e) The ECCS sBray subsystems are the HPCS System and the fPeS+Vs*e+ Ore tPGl subsystem (A er E) may be eensidered OPERAETE during eapable ef being manuelly realigned (+emete er loeal) te the tPGl mode Brevrge cere esel PERRY _ UNIT

               .I B 3.5-'15                                   Revision No.      + I

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY ffi B 3.5.2 ffi i+ MODES 4 and 5 te ensure adequate eeelant inventery and suffieient OPERA,EItITY during MODES 1' 2; and 3 are diseussed in the net required te be OPERABtE during MODE 5 with the reaeter vessel inventery tess pri The Auternatie Depressurizatien Systern is Frs[required te be OPERABTE tPCS' HPCS' and tPel subsystems ean previde eere eeeling witheut any lf any ene required ECC$injeetien/spray subsystem is ineperable; the OPERABTE status within 'l heurs= ln this Cenditien' the remaining OPERABTE subsystem ean previde suffieient RPV fleeding eapability te reeever frem an inadvertent vesse! draindewn. Hewever; everall system reliability is redueed beeause a single failure in the remaining OPERABTE ien Time fer restering the required ECCS injeetien/spray subsystem te OPERABTE status is based en engineering judgment that eensidered the With the ineperable subsystem net restsred te OPERABTE stetus within (OPDRVS) te minimize the prebability ef a vessel draindewn and the inue PERRY - UNIT 1 B 3.5-16 Revision No. +

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY ffi B 3.5.2 ffi aetiens must Ue i restered te OPERABTE ettus with'in { heurs, lf at least ene ECCS injeetien/spray subsystem is net restered te OPERI\BtE status within the 4 heur Cempletien Time; additienal aetiens

                                                                                     +en primarV eentainme required'eernpenent.is ineperable' then it must be restered te OPERABT E restere the eempenent te OPERABTE status, ln additien; at least ene ir leek deer eemBletes the beundary fer eentrel ef petential radieaetive releases, With the appre^riate administrative eentrels; hewever' the i

aeeess and due te the slew pregressien ef events whieh may result frem the Iaek ef available Ee e S, The laek ef available Ee CS during shutdewn PERRY - UNIT 1 B 3.5-17 Revision No. +

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY ffi B 3.5.2 EAffi issien requirements ef this Genditien are satisfied, The t heur Cempletien Time te restere at Ieast ene ECCS injeetien/spray te previde the required eeeling eapaeity er te initiate aetiens te plaee the the envirenment, RESUIREMENTS The minimum water level ef 16 ft 6 inehes required fer the suppressien peel is Beriedieally verified te ensure that the suppressien peel will reeireulatien velume; and vertex preventien, With the suppressien peel CST, velumes), a Suppressien Peel tevel Adjustment Table is eentained in the PIant Data Beek, This table Iists water level adjustment fer varieus drywell te eentainment differentja{ pressures, The table adiustment faeters are used te medify the lndiated suppressien peel wate" level te inee these differential pressures were direetly aeeeunted fer in the shert term analyses; When the suppressien peel level is -- 16 ft 6 inehes; the HPGSystem is pump, Therefere; a verifieatien that either the suppressien peel water leve+is PERRY _ UNIT 1 B 3.5-18 Revision No. +

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY ffi B 3.5.2 EASES sunv=trunNee sn s, REQU IREMENTS ing 150;000 gallens ef water available fer HPGS; equivalent t+a velume ef FIPV. The Surveillanee Frequeney is Gntrelled under the Surveillanee w3 The Eases previded fer SP 3=5,1,1 is aBplieable te SR 3.5,2# SR 3,5.2,'l Verifying the eorreet alignmentfer manual, pewer eperated, and

                                                              +en, This SR dees net apply te valves that are leeked; sealed; er etherwise seeured in pesitien sinee in*

valves petentially eapable ef being mispesitiened are in the eerreet pesj inadvertently PERRY - UNIT 1 B 3.5-19 Revision No. {4

TS BASES MARK-UP . PROVIDED ffi FOR INFORMATION ONLY B 3.5.2 gAffi RESUIREMENTS ln MODES 4 and 5, the RHR System may eperate in the shutdewn Therefere; RHR valves that are required fer tPGl subsystem eperatien that allews ene tPGl subs)'stem te be eensidered OPERABTE during

                                                                                 +ng manually realigned (remete er leeal) te the tPCl mede and net ethenvise ineperable, This will ensure adequate eere eeeling if an inadvertent vessel draindewn sheuld eeeur, The Bases previded fer SR 3,5,1 .'l and SR 3,5,1 =5 are aBBlieable te REFERENCES-    1,   USA.R, Seetien 6,3=3, PERRY _ UNIT 1                        B 3.5-20                            Revision No. 11  I

TS BASES MARK.UP . PROVIDED RPV Water lnventorv FOR INFORMATION ONLY B 3.5.2 8 3.5 E]VIERGENCY COR CONTROL, AND REACTOFI CORE ISOLATION COOLING (RCIC) SYSTETU 9 3.5.2 Reactor Pre EASET BACKGROUND The RPV have the potential tt tne water tevet s neat is reduced, wn ctaO ogrtoration. above the top of the elevated claddino t nppltcnelE wirh th SnfffY to mitigate ANALYSES RPV water Satetv-fimlt e.t . radioactive materi eveflI_occur. A double-ended gu is not postulated in I/ODES 4 and 5 due to the reduced RCS oressure. feduced piping str cqnsider-ed in which sin Oraining of the RPV with tne highest fl penetration ttow p seisntic event, loss o based on engineering iudgment, that while in hdODES 4 and 5, one low pressure fCCSjnje vessel water level. Operating exper public neaftn and s tnventorv Control LCO The RPV water le if an unexpected dr level remains abov Safetv Limit 2.1.1.3. PERRY _ UNIT 1 B 3.5-15 Revision No. TBD

TS BASES MARK.UP. PROVIDED nPV Water tnventor FOR INFORMATION ONLY B 3.5.2 EASES I-CO fne l-imitino Co (continrred) RPV water inventorv to TAF to be > 36 hours. A DRAIN TltvlE of 36 hours is considere unexpecteO Urainl nPV water inventor in greater than S6 h Satetv fimit e.t opelEltitr= One ECCS iniection caoanb ot Ueing ma unexpecteO Urain is OetineO as eitner o suUsvstems, one l-o Pressure Core Sorav LPCS Svstem consis transter water tro effsists of one motor d trom tne suppressl The I CO is mndifiad hv Note whieh allows a reorrire.l I PCI suhsvstem (n or el to be cEnside Oecav neat removal tocat) to ttre tPCt mo operation tor Oeca is not ooeratino or shutdown coollnp m Svstem mav Ue requl remove decav heat a restrictions on On unexpecteO drainl operation to mainta reaching the TAF. NPPLICNEILITY RPV Reouirements on wa in t-COs in Section S ECCS. nCtC. anO nPV controt is require whenever irradiat PERRY - UNIT 1 B 3.5-16 Revision No. TBD

TS BASES MARK.UP - PROVIDED nPV Water Inventor FOR INFORMATION ONLY B 3.5.2 AASES (continued) ACTIONS A.1 and B.1 It tne requireO fCC restoreO to OPf nng controls on DRAIN T drainlnO event cou sunsvstem, noweve iniection/sorav s restorino the reouired ECCS iniection/sorav subsvstem to OPEHABLE status is based on e on DRAIN TlME and th that would result l lf the inoperable E OPf nnelE status wl initiateO immeOia operatino without inctudes tne neces pumps anO vatves ne snoutO an unexpect iniection mav be ma svstems or subsvst capable of maintal hours. lf recircul determining the ne e-LanstG2 with tne unnlru lME hours, compensato imolement mitioatino actions should an unexoected drainino event occur. ShoutO a Oraining e tnere is ootential radioactive mater radioactive mater beino released to t The primarv containment orovides a controlled volume in which fission oroducts can be contained. diluted. and processed prior to release to the ruironment. nequir establish the prim f tME. fhe required containment bound available-PERRY _ UNIT 1 B 3.5-1 7 Revision No. TBD

TS BASES MARK.UP . PROVIDED RPV Water lnventorv FOR INFORMATION ONLY B 3.5.2 MSE=T ACTIONS r. 1 and C 2 (conlinued) Verification that must be performed w administrative ac equipment. Primarv c primarv containme of the capabilitv t in less tnan tne nnn to isotate tne prim and necessarv mate requireO to receiv manuallv within th containment penet witnin + nours. ftle does not require ma O.t, O.Z, and DS With tne Onnlf'l f lUe implementeO in cas tnat it tne OnnllU f l aBplcable= nequired Action D. metnoO ot water inl subsvstem requir im inctuOes tne neces pumps anO vatves ne snouU an unexpect O.t states tnat eit aOOitionat metnoO ottsite electrica manualtv inltiate subsvstems. fne aO access water lnven water tevel above t water iniection an part ot tne same wat occur. it mav be cre (continued) PEHRY _ UNIT 1 B 3.5-18 Revision No. TBD

TS BASES MARK-UP . PROVIDED nPV Water Inventor FOR INFORMATION ONLY B 3.5.2 BASES ACTIONS D.1, D.2. and D.3 (continued) ShoutO a Oraining e tnere is potential raOioactive matet raOioactive mater nelng released to t fne orlmarv contal proOucts can ne con environment. nequ initiateO to estab fne orimarv contal containment nounO rmmeOlatetv init flow path is isotat control room. EJ tt tne nequireO nct or O are not met or it ne initlated immeO conOition, tnere m Oraining event to p ruote tnat nequireO n DFIAIN TIhIF is lesc than t hnrrr SURVEILLANCE SR 3. REOU IREI\TENTS fnis Surveittance the fnf is > S6 hours to iOentitv anO inl ot nPV water inven ing tne fnf in greater t chattenge to Safet plant operation (continued) PERRY _ UNIT 1 B 3.5-'19 Revision No. TBD

TS BASES MARK.UP - PROVIDED nPV Water lnventor FOR INFORMATION ONLY B 3.5.2 BASES sunvHlrnNce sn s. REQU IHEIVENTS The definition of D and drain rates are determined using ion cmsiOerlng the chano Control Rod HPV pe ive Uecnanism removeO cross-sectional a controt roO guide t penetration to adl sectional area ot t The detinitlon ot U penetration ttow p bv manual or automa secureO in tne ctos prevent ttow ot rea blank flange or oth numner ot Uotts to p f artnouat<e. fUorma isotation devlces closeO or isotate iqq plant maintenanc ies. fne nesUua Ueat n consiOereO an inta (neterence 0t nave isotation Oevices sunsvstem is precl considered an inta transferred to nem isolation signals. fne exctusion ot pe f tUf must consiOer initiating event o scattotOing, temp seals. etc.). lt ta drainino event from a closed svstem or between the RPV and the isolation Oevice, D RAIN Tllt/E calculation. (continrred) PEFIRY _ UNIT 1 B 3.5-20 Revision No. fED

TS BASES MARK.UP . PROVIDED RPV Water lnventorv FOR INFORMATION ONLY B 3.5.2 EASES SURVEILLANCE SR 3. REOU IREIVIENTS Surveittance neou pertormances. f ne cnanpe tne DnntlU f l determined= The Surveillance F Frequencv Control SR 3.5.2.2 and SR 3.5.2.3 The minimum water l perioOicaltv verl adeouate net oositive suction head (NPSH) for the ECCS oumo. reeirculation volum water tevet less tn sunsvstem is inope When tne suppress is considered OPfRAg CST water level is s oumo. Therefore. a verification that either the suooression oool water level is > t6 ft O ln o tne csr ano tne csr c tne HPCS Svstem can fne Surveittance f frequencv Control sFl 3.5.2.4 The flow path pipin entraineO air. nlal iniectionlsprav s subsvstem will per followino an ECCS initiation sional. One acceotable method of ensurino that the lines are f fne Surveittance f frequencv Control PERRY

UNIT 1 B 3.5-20a Revision No. TBD I

T$ BASES MARK.UP - PROVIDED nPV Water lnventor FOR INFORMATION ONLY B 3.5.2 BASES SURVEILLANCE ER 3. REOU,IREIVIENTS (contlnueOt Verit automatic valves l assurance that the This SR does not app secured in positio Pqsition orior to loc initiation signal I vatve witt automatl not require anv tes veriticatlon tnat tno are-in the correct pos be inadvertentlv mi fne Surveittance f Freouencv Control sR 3.5.2.6 Verltving that the manuattv starteU a tne subsvgtem is av fCCS iniectioUsora necessarv to avoid operating time of l Tne Surveillance F Frequencv Control sR 3.5.2.7 Veritving that each ttow patn actuates water tevet isotat from dropping belo fne Surveitlance f Frequency Control PERRY _ UNIT 1 B 3.5-20b Revision No. TBD

TS BASES MARK-UP . PROVIDED RPV Water lnventor FOR INFORMATION ONLY B 3.5.2 BASES SUHVEILLANCE SR 3. RFQUTRETUENTS (eontinrred\ The reotrired ECC.S s uhsvstam iq ronrrired tn have e menrrnl start capabilitv. This S cause the required as designed, inclu to their required p from a standbv conf Level S iniection v The Surveiltance l Proffam-fnis Sn is moOltieO the Surveillance. can be demonstrate iniection into the RPV is not reouired durino the Surveillance. nf f e nf f'lCfS t. lnto Inventorv in Boili November 1984.

e. tntormation lUotl gecause of N/isali0 S. Generic Letter g Vessel Water Level I t o cFR so.s+(t),"

                   +. NRC eulletin gS-Water l-evet tnstrum S. lntormation Notl Reactor Vesset Ora O. General Electri Itrtisatignment Duri Eehruaryll83-PERRY _ UNIT    1                         B 3.5-20c                          Revision No. IBD  I

TS BASES MARK.UP . PROVIDED RCIC System FOR INFORMATION ONLY B 3.5.3 B 3.5 ETVTERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONIEOL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTETVI 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 Ioss 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 reactor vessel head spray nozzle. Suction piping is provided from the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV. However, only the capability to take suction from the suppression pool is required for OPERABILITY. lf the CST volume is low, or the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from main steam line A, upstream of the inboard main steam Iine isolation valve. The RCIC System is designed to provide core cooling for a wide range of reactor pressures, 165 psia to 1215 psia. Rated flow is required up to 1 1 18 psia, based on operation of the Safety Relief Valves in the Relief and Low-Low-Set modes (T.S. 3.3.6.4) during the vessel isolation transients for which RCIC is designed. Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increa$es, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test Iine is provided to route water from and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV. (continued) PERRY - UNIT 1 B 3.5-21 Revision No. 5

TS BASES MARK.UP . PROVIDED RCIC System FOR INFORMATION ONLY B 3.5.3 BASES BACKGROUND The RCIC pump is provided with a minimum flow Iine, which discharges (continued) to the suppression pool. The valve in this line automatically opens to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge 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 events by SAFETY providing makeup coolant to the reactor. The RCIC System is not an ANALYSES Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the system is included in the Technical Specifications as required by the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132). LCO The OPERABILITY 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. APPLICABILITY The RCIC System is required to be OPERABLE in MODE 1, and IVIODES 2and 3 with reactor steam dome pressure > 150 psig since RCIC is the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. ln IVIODES 2 and 3 with reactor steam dome oressure < 150 osio. the ECCS iniection/sorav subsvstems can provide sufficl JLanCj+MODES 4 and 5, RCIC is not reouired to be OPERABLE since RPV water inventorv control is reouired bv LCO 3.5.2, "RPV Water Inventorv Control." theCCS i+ ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable RCIC system. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable RCIC system, and the (continued) PERRY _ UNIT 1 B 3.5-22 Revision No. +

TS BASES MARK-UP. PROVIDED RCIC System FOR INFORMATION ONLY B 3.5.3 BASES SURVEILLANCE sR 3.5.3.5 (continued) REQUIRETUENTS automatic pump startup and actuation of all automatic valves to their required positions. This Surveillance test also ensures that the RCIC System will automatically restart on an RPV Iow water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool on a condensate storage tank low water Ievel signal and on a suppression pool high water Ievel signal. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.32, "Reactor Core Isolation Cooling (RCIC) System Instrumentation," overlaps this Surveillance to provide complete testing of the assumed safety 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 full flow test Iine, coolant injection into the RPV is not required during the Surveillance. REFERENCES 1. 10 CFR 50, Appendix A, GDC 33.

2. USAR, Section 5.4.6

3. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),

                     "Recommended lnterim Revisions to LCO's for ECCS Components,"

December 1, 1975. PERRY _ UNIT 1 B 3.5-26 Revision No. 14

TS BASES MARK.UP . PROVIDED Primary Containment Air Locks FOR INFORMATION ONLY B 3.6.1 .2 BASES BACKGROUND DBA. Not maintaining air lock integrity or leak tightness may result in a (continued) leakage rate in excess of that assumed in the unit safety analysis. APPLICABLE The DBA that postulates the maximum release of radioactive material SAFETY within primary containment is a LOCA. In the analysis of this accident, it ANALYSES is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (LJ of 0.20% by weight of the containment and drywell air per 24 hours at the calculated maximum peak containment pressure (Pr) of 7.80 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks. Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the intermediate building. Primary containment air locks satisfy Criterion 3 of the NRC Final Policy Statement on Technical Specification Improvements (58 FR 39132) in IVIODES 1, 2, and 3. During MODES 4 and 5, there are no accident analyses that credit the air locks. However, it was determined that a Specification should remain in place per Criterion 4 to address eperetiens wi uel handling accidents. Criterion 3 would apply if dose calculations are revised to credit the air locks during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. LCO As part of the primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event. The primary containment air locks are required to be OPERABLE. For each air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock ( continu ed) PERRY. UNIT 1 B 3.6-8 Revision No. 7 I

TS BASES MARK.UP - PROVIDED Primary Containment Air Locks FOR INFORMATION ONLY B 3.6,1.2 BASES LCO allows only one air lock door to be open at a time. This provision ensures (continued) that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single OPERABLE door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from primary containment. APPLICABILITY ln MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. ln MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature Iimitations of these MODES. Therefore, maintaining OPERABLE primary containment air locks in MODE 4 or 5 to ensure a control volume is only required during situations for which significant releases of radioactive material can be postulated; such as du+iflg eperatiens with a petential fer draining the reaeter vessel (OPDRVS)' er during movement of recently irradiated fuel assemblies in the primary containment. ies-e+e leaded inte th+eere, Therefere; if the fuel is fully eff leaded frem the ePErqA,BtE. Due to radioactive decay, handling of fuel only requires primary containment air lock OPERABILITY when the fuel being handled is recently irradiated, i.e., fuel that has occupied paft of a critical reactor core within the previous 24 hours. However, even though the air Iocks are not required to be OPERABLE during handling of fuel that is not recently irradiated, there are still controls provided to ensure the ability to close a door in an air lock should the need arise. Closure of a door, even though it is not OPERABLE, would reduce the potential for gross unfiltered leakage. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 4). ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component. PERRY _ UNIT 1 B 3.6-9 Revision No. 4

TS BASES MARK.UP - PROVIDED Primary Containment Air Locks FOR INFORMATION ONLY B 3.6.1.2 BASES ACTIONS D.1 and D.2 (continued) lf the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time while operating in IMODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. E.1+ndE2 lf the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion time during d+t+i++grmovement of recently irradiated fuel assemblies in the primary containment, action is required to immediately suspend activities that represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a Condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. inue SURVEILLANCE sR 3.6.1 .2.1 REQUIREMENTS l/laintaining primary containment air locks OPERABLE requires compliance with the Ieakage rate test requirements of the Primary Containment Leakage Rate Testing Program when in MODES 1, 2, and 3. This SR reflects the Ieakage rate testing requirements with regard to air lock Ieakage (Type B leakage tests). The acceptance criteria were established prior to initial air Iock and primary containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the combined Type B and C primary containment (continued) PERRY - UNIT 1 B 3.6-14 Revision No. -1

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY PCIVs B 3.6.1.3 BASES (continued) APPLICABLE The PCIVs LCO was derived from the assumptions related to minimizing SAFETY the Ioss of reactor coolant inventory, and establishing the primary ANALYSES containment boundary during major accidents. As part of the primary containment boundary, PCIV OPERABILITY supports leak tightness of primary containment. Therefore, the safety analysis of any event requiring isolation of primary containment is applicable to this LCO, The DBAs that result in a release of radioactive material for which the consequences are mitigated by crediting PClVs, are a loss of coolant accident (LOCA), and a main steam line break (MSLB) (Refs. 1 and 2). ln the analysis for each of these accidents, it is assumed that PCIVs are either closed or function to close within the required isolation time following event initiation. This ensures that potential paths to the environment through PCIVs are minimized. Of the events analyzed in Reference 1, the LOCA is the most limiting event due to radiological consequences. lt is assumed that the primary containment is isolated such that release of fission products to the environment is controlled. The inboard 42 inch purge supply and exhaust valves may be unable to close in the environment following a LOCA. Therefore, each of the purge valves is required to remain sealed closed during MODES 1 , 2, and 3. The outboard MSIVs must have a safety related air source available for use following an accident in order for leakage to be within limits. Therefore, anytime that this air source from the "8" train of P57 Safety Related Air System is not available, the outboard MSIVs may not be able to maintain valve leakage within the specified limits, PCIVs satisfy Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) in MODES 1,2, and 3. During MODES 4 and 5, there are no accident analyses that credit the primary containment. However, it was determined that Specifications shouldremaininplaceperCriterion4toaddress@ uel handling accidents. Criterion 3 would apply if dose calculations are revised to credit the primary containment during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours). (continued) PERRY - UNIT 1 B 3.6-18 Revision No. 7

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY PCIVs B 3.6.1.3 BASES (continued) LCO PCIVs form a part of the primary containment boundary and some also form a part of the RCPB. The PCIV safety function is related to minimizing the loss of reactor coolant inventory, and establishing primary containment boundary during a DBA. The power operated isolation valves are required to have isolation times within limits. Additionally, power operated automatic valves are required to actuate on an automatic isolation signal. Primary containment purge supply and exhaust valves are not qualified to close under accident conditions and therefore must be sealed closed (inboard) or blocked to prevent full opening (outboard valves) to be OPERABLE. The normally closed PCIVs or blind flanges are considered OPEHABLE when, as applicable, manual valves are closed or opened in accordance with applicable administrative controls, automatic valves are de-activated and secured in their closed position, check valves with flow through the valve secured, or blind flanges are in place. The valves covered by this LCO with their associated stroke times, if applicable, are listed in Reference 3. Primary containment purge valves with resilient seals, secondary containment bypass valves, MSlVs, and hydrostatically tested valves must meet additional leakage rate requirements. Other PCIV leakage rates are addressed by LCO 3.6.1.1, "Primary Containment-Operating," as Type B or C testing. This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory, and establish the prirnary containment boundary during accidents. APPLICABILITY ln IVODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. ln MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, ffiost PCIVs are not required to be OPERABLE and the primary containment purge valves are not required to be sealed closed in MODES 4 and 5. Certain valves are required to be OPERABLE-, when the in+a

                   @ing (continued)

PERRY _ UNIT 1 B 3.6-19 Revision No. 4 I

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY PCIVs B 3.6.1 .3 BASES APPLICABILITY (continued) associated instrumentation is required to be OPEHABLE according to LCO 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation." (This does not include the valves that isolate the associated instrumentation.) Due to radioactive decay, handling of fuel only requires containment isolation valve OPERABILITY when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critica! reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 5). ACTIONS The ACTIONS are modified by a Note allowing penetration flow path(s) except for the inboard 42 inch (1M14-F045 and 1M14-F085) inch primary containment purge supply and exhaust isolation valve flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated. Due to the size of the containment purge supply and exhaust (continued) PEFIFIY _ UNIT 1 B 3.6-19a Revision No, 4 I

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY PCIVs B 3.6.1 .3 BASES ACTIONS E.1 and E.2 (continued) 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. F.1 lf any Required Action and associated Completion Time cannot be met, action is required to suspend activities that represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a condition that minimizes risk. lf applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended (Required Action F.1). Suspension of these activities shall not preclude completion of movement of a component to a safe condition. (ePDRVs) are engei OPDRVS (Required Aetien G,1)te minimize the prebability ef a vessel iens weuld result in iselating a required residual heat remeval (RHP) shutdewn eesling pathwey G,l' te immediately initiate aetien te restere the valves te OPERABTE ing taken te restere the valves, ln additien; in MODE 4 er 5; ene ef the RHFI shutdewn eeeling suetien OPERABTE sinee tCO 3=3,6,1' "Primary eenHinment and Drywell lselatien lnstrumentatien" requires ene trip syslem ef the Reaeter Vessel Water tevel tew tevel 3 instrurnentatien te be OPERABtE, Therefere, if 1E12 F008 er 1E'l2 F009 valve; Required Aetien G,2 must be taken; in Funetien 5,b, PERRY _ UNIT 1 B 3.6-25 Revision No. 1S I

TS BASES MARK-UP - PROVIDED PCIVs FOR INFORMATION ONLY B 3.6.1 .3 BASES SURVEILLANCE SR 3.6.1 .3.1 (continued) REQUIREMENT is restricted to one valve in a penetration flow path at a given time (refer to discussion for Note 1 of the ACTIONS) in order to effect repairs to that valve. This allows one purge valve to be opened without resulting in a failure of the Surveillance and resultant entry into the ACTIONS for this purge valve, provided the stated restrictions are met. Condition D must be entered during this allowance, and the valve opened only as necessary for effecting repairs. Each purge valve in the penetration flow path may be alternately opened, provided one remains sealed closed, if necessary, to complete repairs on the penetration. The SR is modified by a Note stating that the inboard 42 inch primary containment purge supply and exhaust isolation valves are only required to be sealed closed in MODES 1, 2, and 3. lf a LOCA inside primary containment occurs in these MODES, the purge valves may not be capable of closing before the pressure pulse affects systems downstream of the purge valves and the subsequent release of radioactive material will exceed limits prior to the closing of the purge valves. At other times when the purge valves are required to be capable of closing te+-3u+ing gPERVs), pressurization concerns are not present and the purge valves are allowed to be open. sR 3.6.1 ,3.2 This SR verifies that the 18 inch (1M14-F190, 1MI14-F195, 1IVI14-F200, and 1 M 14-F205) and outboard 42 inch (1 M 14-F040 and 1 tU 14-F090) primary containment purge supply and exhaust isolation valves are closed as required or, if open, open for an allowable reason. lf a purge valve is open in violation of this SR, the valve is considered inoperable. lf the inoperable valve is not othenryise known to have excessive leakage when closed, it is not considered to have purge valve leakage outside the limits (Condition D). The SR is also modified by a Note (Note 1) stating that primary containment purge valves are only required to be closed in MODES 1, 2, and 3. At times other than MODE 1,2, or 3 when the purge valves are requiredtobecapableofclosingpressurization concerns are not present and the purge valves are allowed to be open (automatic isolation capability would be required by SR 3.6.1.3.5, SR 3.6.1.3.7 , and SR 3.6.1 .3.8), (continued) PERRY - UNIT 1 B 3.6-26 Revision No. 4

TS BASES MARK.UP. PROVIDED PCIVs FOR INFORMATION ONLY B 3.6.1 .3 BASES SURVEILLANCE sR 3.6.1.3.5 (continued) REQUIRETUENT full closure isolation time is demonstrated by SR 3.6.1.3.7. The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the safety analysis. The isolation time and Frequency of this SR are in accordance with the lnservice Testing Program. sR 3.6.1.3.6 For primary containment purge valves with resilient seals, additional Ieakage rate testing beyond the test requirements of 10 CFR 50, Appendix J (Ref. 4), is required to ensure OPERABILITY. Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Additionally, this SR must be performed within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation (beyond that which occurs to a valve that has not been opened). Thus, decreasing the interval (from 184 days) is a prudent measure after a valve has been opened. A purge valve leak rate acceptance criterion of 0.05 L, has been assigned to these valves. Note that purge valve leakage is a contributor to secondary containment bypass leakage, which has a separate acceptance criterion. The SR is modified by a Note stating that the primary containment purge valves are only required to meet leakage rate testing requirements in IVIODES 1, 2, and 3. lf a LOCA inside primary containment occurs in these MODES, purge valve leakage must be minimized to ensure offsite radiological release is within limits. At other times when the purge valves are required to be capable of closing , pressurization concerns are not present and the purge valves are not required to meet any specific leakage criteria. sR 3.6.1 .3.7 Verifying that the full closure isolation time of each MSIV is within the specified limits is required to demonstrate (continued) PERRY - UNIT 1 B 3.6-29 Revision No. {4

TS BASES MARK-UP . PROVIDED PCIVs FOR INFORMATION ONLY B 3.6.1.3 BASES SURVEILLANCE SR 3.6,1 .3,1 1 (continued) REQUIREIUENTS demonstrated at the frequency of the leakage test requirements of the Primary Containment Leakage Rate Testing Program. This SR is modified by a Note that states these valves are only required to meet the combined leakage rate in MODES 1, 2, and 3 since this is when the Reactor Coolant System is pressurized and primary containment is required. In some instances, the valves are required to be capable of automatically closing during TVIODES other than MODES 1, 2, and 3. However, specific leakage rate limits are not applicable in these other MODES or conditions. A second Note states that the Feedwater lines are excluded from this particular hydrostatic (water) testing program. This is because water leakage from the stem, bonnet and seat of the third, high integrity valves in the feedwater lines (the gate valves) is controlled by the Primary Coolant Sou rces O utside Contai nment Prog ram (Tech nical Specification 5.5.2). The acceptance criteria for the Primary Coolant Sources Outside Containment Program is 7.5 gallons per hour. sR 3.6.1 .3.12 Verifying that each outboard 42 inch (1M14-F040 and 1M14-F090) primary containment purge supply and exhaust isolation valve is blocked to restrict opening to < 50" is required to ensure that the valves can close under DBA conditions within the time limits assumed in the analyses of References 2 and 3. The SR is modified by a Note stating that this SR is only required to be met in MODES 1, 2, and 3. If a LOCA inside primary containment occurs in these IVIODES, the purge valves must close to maintain containment Ieakage within the values assumed in the accident analysis. At other times when the purge valves are required to be capable of closing{+g=,

                 @,pressurizationconcernsarenotpresent,thusthepurge valves can be fully open. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued) PERRY - UNIT 1 B 3.6-32 Revision No. {4 I

TS BASES MARK.UP - PROVIDED Prima ry Contai n ment-Shutdown FOR INFORMATION ONLY B 3.6.1 .10 BASES BACKGROUND This Specification ensures that the performance of the primary (continued) containment, in the event of a fuel handling accident involving handling of recently irradiated fuel, provides an acceptable leakage barrier to contain fission products, thereby minimizing offsite doses. APPLICABLE The safety design basis for the primary containment is that it contain SAFETY fission products to limit doses at the site boundary to within limits. The ANALYSES primary containment OPERABILITY in conjunction with the automatic closure of selected OPERABLE containment isolation valves (LCO 3.6.1.3, "Primary Containment lsolation Valves (PClVs)," and LCO 3.3.6.1, "Primary Containment and Drywell lsolation lnstrumentation"), assures a leak tight fission product barrier. The fuel handling accident calculations do not credit the primary or secondary containment; all gaseous fission products released from the water pool over the damaged fuel bundles are assumed to be immediately discharged directly to the environment (Ref. 2). During lvlODES 4 and 5, there are no accident analyses that credit the primary containment. However, it was determined that Specifications should remain in place per Criterion 4 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) to address an#fuel handling accidents. Criterion 3 of the NRC Policy Statement would apply if dose calculations are revised to credit the primary containment during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours). LCO Primary containment OPERABILITY is maintained by providing a contained volume to limit fission product escape following a fuel handling accident involving handling of recently irradiated fuel, or an unanticipated water level excursion. Compliance with this LCO will ensure a primary containment configuration, including the equipment (continued) PERRY - UNIT 1 B 3.6-56 Revision No. 7

TS BASES MARK.UP - PROVIDED Prim ary Contai n ment-Shutdown FOR INFORMATION ONLY B 3.6.1 .1 0 BASES LCO hatch, that is structurally sound and that will limit leakage to those (continued) leakage rates assumed in the safety analysis. Since offsite dose analyses conservatively assume LOCA leakage pathways and rates, the isolation and closure times of automatic containment isolation valves supports an OPERABLE primary containment during shutdown conditions. Furthermore, normal operation of the inclined fuel transfer system (IFTS) without the IFTS blind flange installed is considered acceptable for meeting Primary Containment-Shutdown OPERABILITY. Leakage rates specified for the primary containment and air locks, addressed in LCO 3.6.1.1 and LCO 3.6.1.2 are not directly applicable during the shutdown conditions addressed in this LCO. APPLICABILITY ln I\4ODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these tUODES. Therefore, maintaining an OPERABLE primary containment in MODE 4 or 5 to ensure a control volume, is only required during situations for which significant releases of radioactive material can be postulated; such as during movement of recently irradiated fuel assemblies in the primary containment

                                                                           . Due to radioactive decay, handling of fuel only requires OPERABILITY of Primary Containment when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 2). OPDRVs assume that ene er mere fuel assemblies are leaded inte the eere, Therefere, if the fuel is fully eff be EPERABtE, (continued)

PERRY - UNIT 1 B 3.6-57 Revision No. 7

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY Primary Containment-Shutdown B 3.6.1.10 BASES (continued) ACTIONS A.1+Hd+hE In the event that primary containment is inoperable, action is required to immediately suspend activities that represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a Condition that minimizes risk. lf applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. 4Jsqif {Gn+inued+ (nnnlinr red\ t PERRY _ UNIT 1 B 3.6-57a Revision No. 4

TS BASES MARK.UP - PROVIDED Primary Contain ment-Shutdown FOR INFORMATION ONLY B 3.6.1 .10 BASES (esntrued) immeAiatetV ;nitl

                  *etien must eentl SURVEILLANCE     sR 3.6.1 .10.1 REQUIRETUENTS This SR verifies that each primary containment penetration that could communicate gaseous fission products to the environment during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive gases outside of the primary containment boundary is within design limits. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. lsolation barriers that meet this criterion are a closed manual valve, a closed and de-activated automatic valve, and a blind flange. This SR does not require any testing or isolation device manipulation. Rather, it involves verification that these isolation devices capable of being mispositioned are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by three Notes. The first Note does not require this SR to be met for pathways capable of being isolated by OPERABLE primary containment automatic isolation valves. The second Note permits the Fire Protection System manual hose reel containment isolation valves (1P54-F726 and 1P54-F727) to be open during shutdown conditions to supply fire mains. The third Note is included to clarify that manual valves opened under administrative controls are not required to meet the SR during the time the manual valves are open. REFERENCES 1. Deleted.

2. USAR, Section 15.7.6 PERRY - UNIT 1 B 3.6-58 Revision No. 41

TS BASES MARK-UP - PROVIDED Containment Vacuum Breakers FOR INFORMATION ONLY B 3.6.1.1 1 BASES APPLICABLE b. lnadvertent actuation of both primary RHR containment spray SAFETY subsystems during normal operation; ANALYSES (continued) The results of these two cases show that the containment vacuum breakers, with an opening setpoint of 0.1 psid, are capable of maintaining the differential pres$ure within design Iimits. The containment vacuum breakers satisfy Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) in MODES 1, 2, and 3. During TVIODES 4 and 5, there are no accident analyses that credit the containment. However, it was determined that Specifications should remain in place per Criterion 4 to address P+R\k)+HC fuel handling accidents. Criterion 3 would apply if dose calculations are revised to credit the containment during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. LCO Only 3 of the 4 vacuum breakers must be OPERABLE for opening. All containment vacuum breakers, however, are required to be closed (except during testing or when the vacuum breakers are performing their intended design function). The vacuum breaker OPERABILITY requirement provides assurance that the containment negative differential pressure remains below the design value. The requirement that the vacuum breakers be closed ensures that there is no excessive bypass Ieakage should a LOCA occur. APPLICABILITY ln MODES 1, 2, and 3, the RHR Containment Spray System is required to be OPERABLE to mitigate the effects of a DBA. Excessive negative pressure inside the containment could occur due to inadvertent actuation of this system. The vacuum breakers, therefore, are required to be OPERABLE in MODES 1, 2, and 3, to mitigate the effects of inadvertent actuation of the RHR Containment Spray System. ln MODES 4 and 5, the probability and consequences of these events are reduced by the pressure and temperature limitations in these MODES; therefore, maintaining containment vacuum breakers OPERABLE is not required in MODE 4 or 5. (continued) PERRY - UNIT 1 B 3.6-60 Revision No. 7 I

T$ BASES MARK.UP. PROVIDED Containment Vacuum Breakers FOR INFORMATION ONLY B 3.6.1.1 1 BASES APPLICABILITY When handling recently irradiated fuel in the primary containmenhend (continued) PDRV+the primary containment is required to be OPERABLE. Containment vacuum breakers are therefore required to be OPERABLE during these evolutions to protect the primary containment against an inadvertent initiation of the Containment Spray System. Due to radioactive decay, handling of fuel only requires OPERABILITY of Containment Vacuum Breakers when the fuel being handled is recently irradiated, i.8., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 2). Sinee OPDRVS assume that ene er mere fuel assemblies are leaded inte the eere; this tCO weuld net be (continued) PERRY - UNIT 1 B 3.6-60a Revision No. 4

TS BASES MARK.UP - PROVIDED Containment Vacuum Breakers FOR INFORMATION ONLY B 3.6.1,1 1 BASES ACTIONS A.1 and A.2 (continued) A Note has been added to provide clarification that separate Condition entry is allowed for each containment vacuum breaker. B.1 and 8.2 If the Required Action of Condition A cannot be met, or if there are three or more containment vacuum breakers not closed, or if there are two or three required vacuum breakers inoperable for other reasons, the plant must be brought to a IVIODE or condition in which the LCO does not apply. To achieve this status, if the plant is operating, ACTION 8.1 requires that the plant be brought to at least MODE 3 within 12 hours and that the plant be brought to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. A Note has been added to stipulate that these Required Actions are only applicable if the plant is in MODE 1, 2, or 3. If the Condition occurs during movement of recently irradiated fuel in the primary containment, hen ACTION B.2 requires that action be taken to immediately suspend activities that represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a Condition that minimizes risk. lf applicable, movement of recently irradiated fuel in the primary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. @ien must be taken te suspend OPDRVS te minimize the prebability ef a vessel draindewn and subsequent petential fer fissien preduet release, Aetien Note has been added to the Required Actions to stipulate that these requirements are only applicable while moving recently irradiated fuel assemblies in the primary containment@. (continued) PERRY - UNIT 1 B 3.6-62 Revision No. 2

TS BASES MARK.UP . PROVIDED Containment Humidity Control FOR INFORMATION ONLY B 3.6.1 .12 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.12 Containment Humidity Control BASES BACKGROUND Primary containment temperature and humidity are initial condition inputs into the analysis that evaluates the initiation of RHR containment spray during normal plant operation. A curve was determined of initial primary containment average temperature and humidity which would maintain peak vacuum inside containment 5 A.72 psi (design is < 0.80 psi) during the spray initiation event. This curve then determines the containment average tem peratu re-to-humid ity com bi nations that are acceptable whenever the conditions exist for the inadvertent containment spray initiation event (whenever the primary containment leak tight barrier has been established). APPLICABLE Reference 1 contains the results of analyses that predict the primary SAFETY containment pressure response for the inadvertent initiation of the RHR ANALYSES Containment Spray System. The initial containment average temperature and relative humidity have an effect on the results of this analyses. As long as the average temperature and relative humidity is maintained within the limits of Figure B 3.6.1.12-1, the design can adequately perform in the inadvertent containment spray event. There is no need to monitor the containment average temperature-to-relative humidity when the primary containment is not OPERABLE (i.e., has large enough openings such that a vacuum would not be created during an RHR containment spray event). The containment relative humidity satisfies Criterion 3 of the NRC Final Policy Statement on Technical Specification Improvements (58 FR 39132) in TMODES 1, 2, and 3. During MODES 4 and 5, there are no accldent analyses that credit the containment. However, it was determined that Specifications should remain in place per Criterion 4 to address ePDRVs-and-fuel handling accidents. Criterion 3 would apply if dose calculations are revised to credit the containment during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. (continued) PERRY - UNIT 1 B 3.6-65 Revision No. 7

TS BASES MARK.UP . PROVIDED Containment Hum idity Control FOR INFORMATION ONLY B 3.6.1.12 BASES (continued) APPLICABILITY ln MODES 1, 2, and 3, the RHR Containment Spray System is required to be OPERABLE to mitigate the effects of a DBA. Excessive negative pressure inside the containment could occur due to inadvertent actuation of this system. The containment average temperature relationship with relative humidity, therefore, is required to be within limits in MODES 1, 2, and 3, to mitigate the effects of inadvertent actuation of the RHR Containment Spray System. In MODES 4 and 5, the probability and consequences of these events are reduced by the pressure and temperature limitations in these MODES. Therefore, maintaining limits on containment relative humidity and temperature is not required in MODE 4 or 5. When handling recently irradiated fuel in the primary containment, aRC during eperatiens with a petentialfer draining the reaeter vessel PDEVslthe primary containment is required to be OPERABLE. Therefore, the proper relationship between containment average temperature and relative humidity must exist during these evolutions. Due to radioactive decay, handling of fuel only requires control over Containment humidity when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 2). ACTIONS A.1 With the primary containment average temperature and relative humidity not within the established limits, actions must be taken to restore the primary containment relative humidity and temperature to within limits. Required Action A.1 stipulates that restoration must occur within I hours. The eight hour Completion Time is based on the time required to restore the relative humidity and temperature Iimits, and the low probability of an event occurring during this time period. (continued) PERRY - UNIT 1 B 3.6-66 Revision No. 4

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY Contain ment Humidity Control B 3.6.1 .12 BASES ACTIONS 8.1 and 8.2 (continued) If the primary containment relative humidity and temperature cannot be restored to within limits within the required Completion Time of Condition A, actions must be taken to place the plant in a MODE or condition in which the LCO does not apply. Required Action 8.1 requires that the plant be brought to at least MODE 3 within 12 hours and Required Action 8.2 requires that the plant be brought to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems C.1+nd42 lf the primary containment relative humidity and temperature cannot be restored to within Iimits within the required completion time of Condition A during movement of recently irradiated fuel in the primary containment, er

                   @actionisrequiredtoplacethepIantinaMoDEor condition in which the LCO does not apply.

Required Actions C.1 a++d4}requires that actions be taken to immediately suspend activities that represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a condition that minimizes rlsk. lf applicable, movement of recently irradiated fuel in the primary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. OPDRVs te minirnize the prebability ef a vessel draindewn and inue (continued) PERRY - UNIT 1 B 3.6-67 Revision No. 3

TS BASES MARK.UP - PROVIDED Suppression Pool Water Level FOR INFORMATION ONLY B 3.6.2.2 BASES LCO The limits on suppression pool water level (> 17 ft 9.5 inches and < 1B ft 6 inches) are required to assure that the primary containment conditions assumed for the safety analyses are met. Either high or low water level limits were used in the analyses, depending upon which is conservative for a particular calculation. The required suppression pool water level readings depend upon the drywell-to-containment differential pressure. The levels correspond to I 17 ft 9.5 inches and < 18 ft 6 inches for a 0 psid drywell-to-containment differential pressure. Adjusted levels are calcu lated for positive d rywel l-to-containment d ifferential pressu res to assure a proper suppression pool volume. When the reactor well to steam dryer storage pool gate is installed, the limits on the suppression pool water level are modif ied to > 18 ft 3.2 inches and < 18 ft 6 inches to assure that the primary containment conditions for the safety analyses are met (Reference 2). APPLICABILITY ln lt4ODES 1 , 2, and 3, a DBA could cause significant loads on the primary containment. ln tt/ODES 4 and 5, the probability and consequences of these events are reduced because of the pressure and temperature limitations in these NilODES. Requirements for suppression pool level in MODE 4 or 5 are addressed in LCO 3.5.2, "Reactor Pressure Vessel (RPV) Water ". ACTIONS 4.1 With suppression pool water level outside the limits, the conditions assumed for the safety analysis are not met. lf water level is below the minimum level, the pressure suppression function still exists as long as horizontal vents are covered, RCIC turbine exhaust is covered, and S/HV quenchers are covered. lf suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and due to OPERABLE containment sprays. Prompt action to restore the suppression pool water level to within the normal range is prudent, however, to retain the margin to weir wall overflow from an inadvertent upper pool dump and reduce the risks of increased pool swell and dynamic loading. Therefore, continued operation for a limited time is allowed. The 2 hour Completion Time is sufficient to restore suppression pool water level to within specified limits. Also, it takes into account the Iow probability of an event impacting the suppression pool water level occurring during this interval. (continued) PERRY _ UNIT 1 B 3.6-77 Revision No. l-t

TS BASES MARK.UP . PROVIDED Secondary Containment FOR INFORMATION ONLY B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4,1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). ln conjunction with operation of the Annulus Exhaust Gas Treatment (AEGT) System and manual closure of certain valves whose Iines penetrate the secondary containment, the secondary containment is designed to reduce the activity Ievel of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, such as during movement of recently irradiated fuel assemblies in the primary containment.pr4uring The secondary containment is a structure that completely encloses the primary containment. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the external pressure. To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment lsolation Valves (SClVs)," and LCO 3.6.4.3, "Annulus Exhaust Gas Treatment (AEGT) System." The isolation devices for the penetrations in the secondary containment boundary are a part of the secondary containment barrier. To maintain this barrier:

a. AII penetrations terminating in the secondary containment required to be closed during accident conditions are closed by at least one manual valve or blind flange, as applicable, secured in its closed position, except as provided in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SClVs)";

(continued) PERRY - UNIT 1 B 3.6-106 Revision No. 4 I

TS BASES MARK.UP . PROVIDED Secondary Containment FOR INFORMATION ONLY B 3.6.4.1 BASES BACKGROUND b. The containment equipment hatch is closed and sealed and the (continued) shield blocks are installed adjacent to the shield building; c The door in each access to the secondary containment is closed, except for entry and exit,

d. The sealing mechanism associated with each shield building penetration, e.g. welds, bellows, or O-rings, is functional; e The pressure within the secondary containment is less than or equal to the value required by Surveillance Requirement SR 3.6.4.1.1, except for entry and exit to the annulus; and

f. The Annulus Exhaust Gas Treatment System is OPERABLE APPLICABLE There is one accident for which credit is taken for secondary containment SAFETY OPERABILITY. This is a LOCA (Ref. 1). The secondary containment ANALYSES performs no active function in response to this Iimiting event; however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis, and that fission products entrapped within the secondary containment structure will be treated by the AEGT System prior to discharge to the environment.

Secondary containment satisfies Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) in MODES 1, 2, and 3. During MODES 4 and 5, there are no accident analyses that credit secondary containment. However, it was determined that Specifications should remain in place per Criterion 4 to address PBzusnd-fuel handling accidents. Criterion 3 would apply if dose calculations are revised to credit secondary containment during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. (continued) PERRY - UNIT 1 B 3.6-107 Revision No. 7

TS BASES MARK.UP. PROVIDED Secondary Containment FOR INFORMATION ONLY B 3.6.4.1 BASES (continued) APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY. ln MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during movement of recently irradiated fuel assemblies in the primary containment petential ter Aral . Due to radioactive decay, handling of fuel only requires OPERABILITY of Secondary Containment when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABLITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 2). lesded frem the reaeter vesselr the seeendary eentainment is net required te be OPERABtE, ACTIONS 4,1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours. The 4 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during tt/ODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal. (continued) PERRY _ UNIT 1 B 3.6-108 Revision No. 4 I

TS BASES MARK-UP . PROVIDED Secondary Containment FOR INFORMATION ONLY B 3.6.4.1 BASES ACTIONS C.1-enC42 (continued) lvlovement of recently irradiated fuel assemblies in the primary containment and ePDRVs can be postulated to cause significant fission product releases. ln such cases, the secondary containment is one of the barriers to release of fission products to the environment. lf applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended if the secondary containment is inoperable. Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. SURVEILLANCE sR 3.6.4.1 .1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.6.4.1.2 and SR 3.6.4.1.3 Verifying that the primary containment equipment hatch is closed and the shield blocks are installed adjacent to the shield building, and secondary containment access doors are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. In this application, the term "sealed" has no connotation of leak tightness. Verifying that all such openings are closed provides adequate (conti n rred) PERRY - UNIT 1 B 3.6-109 Revision No. 44

TS BASES MARK-UP - PROVIDED SCIVs FOR INFORMATION ONLY B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SClVs) BASES BACKGROUND The function of the SClVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1). The OPERABILITY requirements for SCIVs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. Isolation barrier(s) for the penetration are discussed in Reference 2. The isolation devices addressed by this LCO are passive, JVlanual valves and blind flanges are considered passive devices. Penetrations are isolated by the use of manual valves in the closed position or blind flanges. APPLICABLE The SCIVs must be OPERABLE to ensure the secondary containment SAFETY barrier to fission product releases is established. The principal accident ANALYSES for which the secondary containment boundary is required is a loss of coolant accident (Ref. 1). The secondary containment performs no active function in response to this Iimiting event, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Annulus Exhaust Gas Treatment (AEGT) System before being released to the environment. lt,tlaintaining SCIVs OPERABLE ensures that fission products will remain trapped inside secondary containment so that they can be treated by the AEGT System prior to discharge to the environment. SCIVs satisfy Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) in IVIODES 1,2, and 3. During MODES 4 and 5, there are no accident analyses that credit the secondary containment. However, it was determined that Specifications should remain in place per Criterion 4 to address OPBRI/s-nd-fusl handling accidents. Criterion 3 would apply if dose calculations are revised to credit the secondary containment during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. (continued) PERRY - UNIT 1 B 3.6-11 1 Revision No. 7

TS BASES MARK.UP. PROVIDED SCIVs FOR INFORMATION ONLY B 3.6.4.2 BASES APPLICABLE SCIVs form a paft of the secondary containment boundary. The SCIV SAFETY safety function is related to control of offsite radiation releases resulting ANALYSES from DBAs. (continued) The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed, or open in accordance with appropriate administrative controls, or blind flanges are in place. The valves covered by this LCO are included in Table B 3.6.4.2-1. APPLICABILITY In MODES 1,2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, OPERABILITY of SCIVs is required. ln I\IODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during movernent of recently irradiated fuel assemblies in the primary containment.;+rdurlng Due to radioactive decay, handling of fuel only requires OPERABILITY of secondary containment isolation valves when the fuel being handled is recently irradiated, i.8., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24hoursafterthereactorcoreissub-critical(Ref.3).fu that ene er mere fuel as+emblies are leaded inte the eere, Thereferer if the fuel is fully eff Ieaded frem the reaeter vessel' the SCIVs are net required te be OPERABtE. ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. ln this way, the (continued) PERRY _ UNIT 1 B 3.6-1 12 Revision No. 4

TS BASES MARK.UP . PROVIDED SCIVs FOR INFORMATION ONLY B 3.6.4.2 BASES ACTIONS A.1 and A.2 (continued) Required Action A.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or othenruise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low. 8.1 With two SCIVs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. lsolation barriers that meet this criterion are a closed manual valve, and a blind flange. The 4 hour Completion Time is reasonable, considering the time required to isolate the penetration and the low probability of a DBA occurring during this short time. The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves. This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations. C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B cannot be met in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be broughtto at least MODE 3 within 12 hours and to TVIODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. D.1+nd# lf any Required Action and associated Completion Time of Condition A or B cannot be met during movement of recently irradiated fuel assemblies in the primary containment, (continued) PERRY - UNIT 1 B 3.6-114 Revision No. 7 I

TS BASES MARK.UP . PROVIDED SCIVS FOR INFORMATION ONLY B 3.6.4.2 BASES ACTIONS D.1+nd-E# (continued)

                 @hepIantmustbepIacedinaconditioninwhichthe LCO does not apply. If applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. immediately initiated te suspend OPBRVS in erder te minimize the issre+ preguet release, A6 SURVEILLANCE sR 3.6.4.2.1 REQUIRETUENTS This SR verifies that each secondary containment isolation manual valve and blind flange that is not locked, sealed, or othenruise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the secondary containment boundary is within design limits. This SR does not require any testing or isolation device manipulation. Rather, it involves verification that those isolation devices in secondary containment that are capable of being mispositioned are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. Two Notes have been added to this SR. The first Note applies to valves and blind flanges Iocated in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these isolation devices once they have been verified to be in the proper position, is low. A second Note has been included to clarify that (continued) PERRY - UNIT 1 B 3.6-115 Revision No. 44

T$ BASES MARK.UP . PROVIDED AEGT System FOR INFORIUIATION ONLY B 3.6.4.3 BASES BACKGROUND humidity of the airstream to less than 71olo (Ref. 2). The roughing filter (continued) removes large particulate matter, while the HEPA filter is provided to remove fine particulate matter and protect the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter is provided to collect any carbon fines exhausted from the charcoal adsorber. The AEGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. AEGT System flows are controlled by two motor operated control dampers installed in branch ducts. One duct exhausts air to the unit vent, (AEGT Subsystem A exhausts to the Unit 1 plant vent; AEGT Subsystem B exhausts to the Unit 2 plant vent), while the other recirculates air back to the annulus. APPLICABLE The design basis for the AEGT System is to mitigate the consequences of SAFETY a loss of coolant accident. For all events analyzed, the AEGT System is ANALYSES shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment. The AEGT System satisfies Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) in MODES 1 , 2, and 3. During MODES 4 and 5, there are no accident analyses that credit the AEGT System. However, it was determined that Specifications should remain in place per Criterion 4 to address PDRVS-flC-fuel handling accidents. Criterion 3 would apply if dose calculations are revised to credit the AEGT System during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. LCO Following a DBA, a minimum of one AEGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. tVeeting the LCO requirements for two independent operable subsystems ensures operation of at Ieast one AEGT subsystem in the event of a single active failure. (continued) PERRY - UNIT 1 B 3.6-1 19 Revision No. 7

TS BASES MARK.UP . PROVIDED AEGT System FOR INFORMATION ONLY B 3.6.4.3 BASES APPLICABILITY other situations under which significant releases of radioactive material (continued) can be postulated, such as during rnovement of recently irradiated fuel assemblies in the primary containment f'etential ter dra . Due to radioactive decay, handling of fuel only requires OPERABILITY of the AEGT System when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 5). OPDRVs assume that ene er mere fuel assemblies are leaded inte the eere, Therefere-if the fuel is fully eff leaded frem the reaeter veesel; the ACTIONS A.1 With one AEGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. ln this Condition, the remaining OPERABLE AEGT subsystem is adequate to peform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant AEGT subsystem and the low probability of a DBA occurring during this period. 8.1 and 8.2 lf the AEGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the (continued) PERRY - UNIT 1 B 3.6-120 Revision No. 4

TS BASES MARK-UP - PROVIDED AEGT System FOR INFORMATION ONLY B 3.6.4.3 BASES ACTIONS 8.1 and B.2 (continued) plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least IVIODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. C,1 and, C.2,+,+nd# During movement of recently irradiated fuel assemblies in the primary containment,@henRequiredActionA.1cannotbe completed within the required Completion Time, the OPERABLE AEGT subsystem should be immediately placed in operation. This Required Action ensures that the remaining subsystem is OPERABLE, that no (continued) PERRY _ UNIT 1 B 3.6-120a Revision No. 4

TS BASES MARK.UP . PROVIDED AEGT System FOR INFORMATION ONLY B 3.6.4.3 BASES ACTIONS C.1 and' C.24,+nd# (continued) failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected. An alternative to Required Action C.1 is to immediately suspend activities that represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a Condition that minimizes risk, If applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. AJs'q{T minimize the prebability ef a vessel draindewn and subsequent petential fer fissien preduet release= Aetiens must eentinue until ePDRVs are supended. D.1 lf both AEGT subsystems are inoperable in MODE 1, 2, or 3, the AEGT System may not be capable of supporting the required radioactivity release control function. Therefore, LCO 3.0.3 must be entered immediately. E.1+ndE2 When two AEGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. AJs+if fer fissien preduet release, Aetiens must eentjnue until OPDRVs are s'rJspefiCe+ SURVEILLANCE sR 3.6.4.3.1 REQUIRETUENTS Operating each AEGT subsystem from the control room for

                > 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for > 10 continuous hours eliminates moisture on the adsorbers and HEPA filters.

{continued) PERRY - UNIT 1 B 3.6-1 21 Revision No. 44

T$ BASES MARK-UP . PROVIDED CRER System FOR INFORMATION ONLY B 3.7.3 BASES (continued) APPLICABLE The ability of the CRER System to maintain the habitability of the CRE is SAFETY an explicit assumption for the safety analyses presented in the USAR, ANALYSES Chapters 6 and 15 (Refs. 3 and 4, respectively). The emergency recirculation mode of the CRER System is assumed to operate following a DBA. The radiological doses to CRE occupants as a result of the various DBAs are summarized in Reference 4. No single active or passive failure will cause the loss of ability to recirculate air in the CRE. The CRER can provide protection from smoke and hazardous chemicals to CRE occupants. However, an evaluation of chemical hazards from onsite, offsite, and transportation sources has determined that the probability of a hazardous chemical spill resulting in unacceptable exposures is Iess than NRC licensing basis criteria. As a result, the plant licensing basis does not postulate hazardous chemical release events (Refs. 2 and 5). Therefore, no quantitative limits on inleakage of hazardous chemicals into the CRE have been established. A smoke assessment consistent with the guidance in Regulatory Guide 1.196 (Ref.7) and NEI 99-03 Rev. 0 (Ref. 10) determined that reactor control capability can be maintained from either the Control Room or the remote shutdown controls during a smoke event (Ref, 6), Therefore, no quantitative limits on inleakage of smoke into the CRE have been established. Because inleakage limits for hazardous chemicals and smoke are not necessary to protect CRE occupants, the limit established for radiological events is the Iimiting value for CRE inleakage. The CRER System satisfies Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) in MODES 1 , 2, or 3. During MODES 4 and 5, there are no accident analyses that credit the CRER System. However, it was determined that Specifications should remain in place per Criterion 4 to address OPDRVs1ql-fLrsl handling accidents. Criterion 3 would apply if dose calculations are revised to credit the CRER System during handling of recently irradiated fuel, i.e., fuel that has occupied paft of a critical reactor core within the previous 24 hours. LCO Two redundant subsystems of the CRER System are required to be OPERABLE to ensure that at Ieast one is available if a single active failure disables the other subsystem. Total system failure, such as from a loss of both ventilation subsystems or from an inoperable CRE (continued) PERRY _ UNIT 1 B 3.7-1 1 Revision No. 7

TS BASES MARK,UP. PROVIDED FOR INFORMATION ONLY CRER System B 3.7.3 BASES APPLICABILITY ln tt/ODES 4 and 5, the probability and consequences of a DBA are (continued) reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the CRER System OPERABLE is not required in MODE 4 or 5, except s+

                   +     g[During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building=;+nd

{opsElkt Due to radioactive decay, handling of fuel only requires OPERABILITY of the Control Room Emergency Recirculation System when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. ). CREFI System is net required te be OPERABtE, ACTIONS 4.1 With one CRER subsystem inoperable for reasons other than an inoperable CRE boundary, the inoperable CRER subsystem must be restored to OPERABLE status within 7 days. With the unit in this condition, the remaining OPERABLE CRER subsystem is adequate to perform the CRE occupant protection function. However, the overall reliability is reduced because a failure in the OPERABLE CRER subsystem could result in loss of CRER System function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and that the remaining subsystem can provide the requ ired capabi lities. (continued) PERRY - UNIT 1 B 3.7-12 Revision No. 7

TS BASES MARK.UP. PROVIDED CRER System FOR INFORMATION ONLY B 3.7.3 BASES ACTIONS D.1-and, D.2*+nCJ=Z (continued) The Required Actions of Condition D are modified by a Note indicating that LCO 3,0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building, +d{+Fing OPDRVS' if the inoperable CRER subsystem cannot be restored to OPERABLE status within the required Completion Time of Condition A, the OPERABLE CRER subsystem may be placed in the emergency recirculation mode. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected. An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing significant amounts of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. lf applicable, movement of recently irradiated fuel assemblies in the primary containment and fuel handling building must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. @ inimize the prebability ef a vessel draindewn and subsequent petential fer fissien supend+ E.1 If both CRER subsystems are inoperable in IVIODE 1, 2, or 3 for reasons other than an inoperable CRE boundary (i.e., Condition B), the CRER System may not be capable of performing the intended function and the unit is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately. (continued) PERRY - UNIT 1 B 3,7-13 Revision No. 7

TS BASES MARK-UP. PROVIDED CRER System FOR INFORMATION ONLY B 3.7.3 BASES ACTIONS F.1 and F=2 (continued) During movement of recently irradiated fuel assemblies in the primary containmentorfuelhandlingbuilding,@ithtwoCRER subsystems inoperable or with one or more CRER subsystems inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that present a potential for releasing significant amounts of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. lf applicable, movement of recently irradiated fuel assemblies in the primary containment and fuel handling building must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. @ inimize the prebability ef a vessel draindewn and subsequent petential fer fissien suspeflCe+ SURVEILLANCE sR 3.7.3.1 REQUIREMENTS Operating each CRER subsystem for > 10 continuous hours after initiating from the control room and ensuring flow through the HEPA filters and charcoal adsorbers ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. lt also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for > 10 continuous hours eliminates moisture on the adsorbers and HEPA filters. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. sR 3.7.3.2 This SR verifies that the required CRER testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter efficiency, charcoal adsorber efficiency and bypass leakage, system flow rate, and general operating parameters of the filtration system. (Note: Values identified in the VFTP are Surveillance Requirement values.). Specific test Frequencies and additional information are discussed in detail in the VFTP. (continued) PERRY _ UNIT 1 B 3.7-14 Revision No. 44

TS BASES MARK-UP . PROVIDED Control Room HVAC System FOR INFORMATION ONLY B 3.7.4 BASES APPLICABLE The Control Room HVAC System satisfies Criterion 3 of the NRC Final SAFETY Policy Statement on Technical Specification lmprovements ANALYSES (58 FR 39132) in IVIODES 1,2, and 3. During MODES 4 and 5, there are (continued) no accident analyses that credit the Control Room HVAC System. However, it was determined that Specifications should remain in place per Criterion 4 to address OPDRllsnl-fuel handling accidents. Criterion 3 would apply if dose calculations are revised to credit the Control Room HVAC during handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours. (continued) PERRY - UNIT 1 B 3.7-17a Revision No. 7

TS BASES MARK-UP . PROVIDED Control Room HVAC System FOR INFORMATION ONLY B 3.7.4 BASES (continued) LCO Two independent and redundant subsystems of the Control Room HVAC System are required to be OPERABLE to ensure that at Ieast one is available, assuming a single failure disables the other subsystem. Total system failure could result in the equipment operating temperature exceeding limits. The Control Room HVAC System is considered OPERABLE when the individual components necessary to maintain the control room temperature are OPERABLE in both subsystems. These components include the cooling coils, fans, chillers with compressors, ductwork, dampers, and associated instrumentation and controls. The heating coils are not required for control room HVAC OPERABILITY. APPLICABILITY ln lvlODE 1 ,2, or 3, the Control Room HVAC System must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits. ln MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room HVAC System OPERABLE is not required in MODE 4 or 5, except +e++ne+et+ewing

                   +    E[During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building=;+nd Due to radioactive decay, handling of fuel only requires OPERABILITY of the Control Room HVAC System when the fuel being handled is recently irradiated, i.8., fuel that has occupied part of a critical reactor core within the previous 24 hours. Although this Function retains APPLICABILITY during "movement of recently irradiated fuel", which could be interpreted to permit fuel handling before 24 hours of radiological decay if certain buildings and filtration systems are OPERABLE, this is not the case. Fuel handling during that period is prohibited since no dose calculations exist to address a fuel handling accident within the first 24 hours after the reactor core is sub-critical (Ref. 3).

(continued) PERRY _ UNIT 1 B 3.7-18 Flevision No.4

TS BASES MARK-UP . PROVIDED Control Room HVAC System FOR INFORMATION ONLY B 3.7.4 BASES ACTIONS D.1_ond' D.24,+nd*2 (continued) The Required Actions of Condition D are modified by a Note indicating that LCO 3,0.3 does not apply. lf moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assernblies is not sufficient reason to require a reactor shutdown. During movement of recently irradiated fuel assemblies in the primary containmentorfuelhandIingbuiIding,@ifthe inoperable control room HVAC subsystem cannot be restored to OPERABLE status within the required Completion Time of Condition A, the OPERABLE control room HVAC subsystem may be placed immediately in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected, An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing significant amounts of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. lf applicable, movement of recently irradiated fuel assemblies in the primary containment and fuel handling building must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. @ i+im+=e issien supende+ (continued) PERRY - UNIT 1 B 3.7-20 Revision No, 2 I

TS BASES MARK.UP. PROVIDED Control Room HVAC System FOR INFORMATION ONLY B 3.7.4 BASES ACTIONS E.1+nd# (continued) The Required Actions of Condition E.1 are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. During movement of recently irradiated fuel assemblies in the primary containmentorfueIhandlingbuilding,@iftheRequired Action and associated Completion Time of Condition B is not met, action must be taken to immediately suspend activities that present a potential for releasing significant amounts of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. If applicable, handling of recently irradiated fuel in the primary containment or fuel handling building must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. initiated inrmediately te suspend OPDRVS te minimize the prebability ef a vessel draindewn and subsequent petential fer fissien preduet release,

                   ,+etiens must centl SURVEILLANCE      sR    3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat Ioad assumed in the safety analysis.

The SR consists of a combination of testing and calculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. REFERENCES 1. USAR, Section 6.4.

2. USAR, Section 9.4.1.

3. USAR, Section 15.7.6.

PERRY - UNIT 1 B 3.7-21 Revision No. 44

TS BASES MARK.UP - PROVIDED AC Sources - Shutdown FOR INFORMATION ONLY B 3.8.2 B 3,8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources - Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources - Operating." APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 and 5 SAFETY and during movement of recently irradiated fuel assemblies in the primary ANALYSES containment or fuel handling building ensures that: a The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadveftent draindewn ef the

                           \resse}+Fa fuel handling accident involving handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours.

ln general, when the unit is shut down the Technical Specifications (TS) requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, a$$uming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst-case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCOs for required systems. (continued) PERRY _ UNIT 1 B 3.8-34 Revision No. 4

TS BASES MARK.UP. PROVIDED AC Sources - Shutdown FOR INFORMATION ONLY B 3.8.2 BASES LCO powered from offsite power. An OPERABLE DG, associated with a (continued) Division 1 or Division 2 Distribution System Engineered Safety Feature (ESF) 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, a separate offsite circuit to the Division 3 Class 1E onsite electrical power distribution subsystem, or an OPERABLE Division 3 DG, ensure an additional source of powerfor the HPCS. This additional source for Division 3 is not necessarily required to be connected to be OPERABLE. Either the circuit required by LCO Item a, or a circuit required to meet LCO ltem c may be connected, with the second source available for connection. Together, OPERABILITY of the required offsite circuit(s) and DG(s) ensure the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.9., fuel handling accidentsinvoIvinghandIingofrecentIyirradiatedfueI@l d{+iHC\An+). Automatic initiation of the required DG during shutdown conditions is specified in LCO 3.3,8.1, "LOP lnstrumentation." The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective ESF bus(es), and accepting required loads during an accident. Qualified offsite circuits are those that are described in the USAR and are part of the licensing basis for the plant. One offsite circuit consists of the Unit 1 startup transformer through the Unit 1 interbus transformer, to the Class 1E 4.16 kV ESF buses through source feeder breakers for each required division. A second acceptable offsite circuit consists of the Unit 2 startup transformer through the Unit 2 interbus transformer, to the Class 1E 4.16 kV ESF buses through source feeder breakers for each required division. Additional path(s) are available, as described in the USAR and the "AC Sources - Operating" Bases. The required DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 10 seconds for Division 1 and 2 and 13 seconds for Division 3. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions such as: DG in standby with the engine hot and DG in standby (continued) PERRY - UNIT 1 B 3.8-36 Revision No. g

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY AC Sources - Shutdown B 3.8.2 BASES LCO with the engine at ambient conditions. Additional DG capabilities must be (continued) demonstrated to meet required Surveillances, e.9., capability of the DG to revert to standby status on an ECCS signal while operating in parallel test mode. Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. ln addition, proper Ioad sequence operation is an integral part of offsite circuit and DG OPERABILITY since its inoperability impacts the ability to start and maintain energized loads required OPERABLE by LCO 3.8.8. lt is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply atl required AC electrical power d istribution subsystems. As described in Applicable Safety Analyses, in the event of an accident during shutdown, the TS are designed to maintain the plant in a condition such that, even with a single failure, the plant will not be in immediate difficulty. APPLICABILITY The AC sources required to be OPERABLE in IVIODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel handling building provide assurance that: a Systems that-le providecorecoolng mal+rpare available fer the irradiated fuel in the eere in ease ef i;

b. Systems used to mitigate a fuel handling accident involving handling of recently irradiated fuel are available (due to radioactive decay, handling of fuel only requires OPERABILITY of the AC Sources when the fuel being handled is recently irradiated, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours);

c Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. lnstrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.1. (continued) PERRY

UNIT 1 B 3.8-37 Revision No. 4 I

TS BASES MARK.UP. PROVIDED AC Sources - Shutdown FOR INFORMATION ONLY B 3.8.2 BASES (continued) ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply. lf moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require reactor shutdown. 4.1 A required offsite circuit is considered inoperable if no qualified circuit is supplying power to one required ESF division. If two or more ESF 4.16 kV buses are required per LCO 3.8.8, division(s) with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONSTand-movement of recently irradiated fuel.; Yesel" By allowing the option to declare required features inoperable which are not powered from offsite power, appropriate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS Required features remaining powered from offsite power (even though that circuit may be inoperable due to failing to power other features) are not declared inoperable by this Required Action. A.2.1. A.2.2, A.2.3. A#8.1, 8.2, aM 8.3+ndB4 With the offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With the required DG inoperable, the minimum required diversity of AC power sources is not available. lt is, therefore, required to suspend CORE ALTERATIONST anclmovement of recently irradiated fuel assemblies in the primary containment and fuel handling building. vessL Additionally, crane operations over the spent fuel storage pool shall be suspended when fuel assemblies are stored there. Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. lt is further required to initiate (continued) PERRY - UNIT 1 B 3.8-38 Revision No. 2 I

TS BASES MARK-UP - PROVIDED AC Sources - Shutdown FOR INFORMATION ONLY B 3.8.2 BASES ACTIONS A.2.1, 4.2.2, 4.2.3, A#4 8.1, 8.2, and B.3r-and+ (continued) action immediately to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power. Pursuant to LCO 3.0.6, the Actions for LCO 3.8.8 are not entered even if all AC sources to it are inoperable, resulting in de-energization. Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any required ESF bus, ACTIONS for LCO 3.8.8 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized. LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division. c.1 When the HPCS System is required to be OPERABLE, and the additional required Division 3 AC source is inoperable, the required diversity of AC power sources to the HPCS System is not available. Since these sources only affect the HPCS System, the HPCS System is declared inoperable and the Required Actions of the affected Emergency Core Cooling Systems LCO entered. ln the event all sources of power to Division 3 are lost, Condition A will also be entered and direct that the ACTIONS of LCO 3.8.8 be taken. lf only the Division 3 additional required AC source is inoperable, and power is still supplied to the HPCS System by the circuit meeting the LCO Item a requirement, 72 hours is allowed to restore the additional required AC source to OPERABLE. This is reasonable considering the HPCS System will still perform its function, absent an additional single failure. (continued) PERRY _ UNIT 1 B 3.8-39 Revision No. 4

TS BASES MARK-UP. PROVIDED AC Sources - Shutdown FOR INFORMATION ONLY B 3.8.2 BASES (continued) SURVEILLANCE sR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPEHABILITY of the AC sources in other than IflODES 1, 2, and 3. SR 3.8.1.8 is not required to be met since only one offsite circuit is required to be OPERABLE. SR 3.8.1.17 is not required to be met because the required OPERABLE DG(s) is not required to undergo periods of being synchronized to the offsite circuit. SR 3.8.1.20 is not required to be met because starting independence is not required with the DG(s) that is not required to be OPERABLE. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR. This SR is modified by two Notes. The reason for Note 'l is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and preclude de-energizing a required 4160 V ESF bus or disconnecting a required offsite circuit during performance of Surveillances. With limited AC sources available, a single event could compromise both the required circuit and the DG. lt is the intent that these SFls must still be capable of being met, but actual performance is not required during periods when the DG is required to be OPERABLE. Hence the NOTE provides an exception to SR 3.0.i during the period when only one diesel generator is OPEFIABLE. Note 2 states that SRs 3.8.1.12 and 3.8.1.19 are not required to be met when the associated ECCS subsystem(s) are not required to be OPERABLE. These SRs demonstrate the DG response to an ECCS signal (either alone or in conjunction with a loss of offsite power signal). This is consistent with the ECCS instrumentation requirements that do not require ECCS signals when the associated ECCS system is not required to be OPERABLE per LCO 3.5.2, Reactor Pressure V lnventorvControlffi REFERENCES None. PERRY _ UNIT 1 B 3.8-40 Revision No. g

TS BASES MARK.UP - PROVIDED FOR INFORMATION ONLY DC Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources - Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources - Operating." APPLICABLE The initial conditions of Design Basis Accident and transient analyses SAFETY in the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume ANALYSES that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control and switching during all tt/ODES of operation. The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' O P ERABI LITY. The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel handling building 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 Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindewn ef the yesel+Fa fuel handling accident involving handling of recently irradiated fuel, i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours.

The DC sources satisfy Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132). LCO One DC electrical power subsystem (consisting of either the Unit 1 or 2 battery, either the normal or reserve battery charger, and all the associated control equipment and interconnecting cabling supplying power to the associated (continued) PERRY - UNIT 1 B 3.8-60 Revision No. 7

TS BASES MARK.UP. PROVIDED FOR INFORMATION ONLY DC Sources - Shutdown B 3.8.5 BASES LCO bus), associated with the Division 1 or Division 2 onsite Class 1 E DC (continued) electrical power distribution subsystem(s) required OPERABLE by LCO 3.8.8, "Distribution Systems - Shutdowfl," is required to be OPERABLE. Similarly, when the High Pressure Core Spray (HPCS) System is required to be OPERABLE, the Division 3 DC electrical power subsystem associated with the Division 3 onsite Class 1E DC electrical power distribution subsystem required OPEHABLE by LCO 3.8.8 is required to be OPEFTABLE. ln addition to the preceding subsystems required to be OPERABLE, a Class 1 E battery or battery charger and the associated control equipment and interconnecting cabling capable of supplying power to the remaining Division 1 or Division 2 onsite Class 1E DC electrical power distribution subsystem, when portions of both Division 1 and Division 2 DC electrical power distribution subsystems are required to be OPERABLE by LCO 3.8.8. This ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.9., fuel handling accidents involving handling of recently irradiated fuel+nd

                                                  +inC{MR).

Division 1 consists of :

1. 125 volt battery 1 R42-S002 or 2R42-S002.

2. 125 volt f ull capacity charger 1 R42-S006 or 0R42-S007 Division 2 consists of:

1 . 125 volt battery 1 R42-S003 or 2R42-S003.

2. 125 volt f ull capacity charger 1 R42-S008 or 0R42-S009.

Division 3 consists of: 1 . 125 volt battery 1 E22-S005 or 2E22-S005.

2. 125 volt full capacity charger 1 E22-S006 or 0R42-S01 1 APPLICABILITY The DC electrical power sources required to be OPEFIABLE in hIODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment and fuel handling building provide assurance that:

a Required features to provide core cooling makeupare available fer the irradiated fuel assemblies in the eere in (continued) PERRY _ UNIT 1 B 3.8-61 Revision No. 2

TS BASES MARK-UP - PROVIDED FOR INFORMATION ONLY DC Sources - Shutdown B 3.8.5 BASES APPLICABILITY c. Required features necessary to mitigate the effects of events that can (continued) lead to core damage during shutdown are available; and

d. lnstrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The DC electrical power requirements for hilODES 1, 2, and 3 are covered in LCO 3.8.4. ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply. lf moving recently irradiated fuel assemblies while in [I/ODE 1 , 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require reactor shutdown. A.1. A.2.1. A.2.2. and A.2,3-and-4#;4 lf more than one DC distribution subsystem is required according to LCO 3.8.8, the DC subsystems remaining OPERABLE with one or more DC power sources inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS;d movement of recently irradiated fuel-,- By allowing the option to declare required features associated with an inoperable DC power source(s) inoperable, appropriate restrictions are implemented in accordance with the Required Actions of the LCOs for these associated required features. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative alternate actions (i.e., to suspend CORE ALTERATIONS;@ movement of recently irradiated fuel assemblies in theprimarycontainmentandfuelhandlingbuiIding@

                                                             ) is made.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. lt is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems. (continued) PERRY _ UNIT 1 B s.8-62 Bevision No. 2

TS BASES MARK.UP . PROVIDED FOR INFORMATION ONLY DC Sources - Shutdown B 3.8.5 BASES APPLICABILITY A.1. 4.2.1. 4.2.2. and A.2.3#nd+\*+ (continued) The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power. SURVEILLANCE sR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of each SR. This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE DC sources from being discharged below their capability to provide the required power supply or othenruise rendered inoperable during the performance of SRs. It is the intent that these SRs must still be capable of being met, but actual performance is not required REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 15.

PERRY - UNIT 1 B 3.8-63 Revision No. 4

TS BASES MARK-UP - PROVIDED FOR INFORMATION ONLY 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 transient analyses SAFETY in the USAR, ChapterG (Ref. 1) and Chapter 15 (Ref.2), assume ANALYSES Engineered Safety Feature (ESF) systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. The OPERABILITY of the AC and DC electrical power distribution systems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel handling building 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 draindewn ef the vessel or a fuel handling accident involving handling of recently irradiated fuel, i.e.,

fuel that has occupied part of a critical reactor core within the previous 24 hours. The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 3s132). (continued) PERRY - UNIT 1 B 3.8-81 Revision No. + I

TS BASES MARK-UP - PROVIDED Distribution Systems - Shutdown FOR INFORMATION ONLY 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. lmplicit in those requirements is the required OPERABILITY of necessary support required features. This LCO explicitly requires energization of the portions of the AC and DC electrical power distribution systems necessary to support OPERABILITY of Technical Specifications' required systems, equipment, and components-both specifically addressed by their own LCOs, and implicitly required by the definition of OPERABILITY. Maintaining these portions of the AC and DC electrical power distribution systems 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.9., fuel handling accidents involving handling of recently irradiated fuel and inadvertent reaeter vessel draindewn). APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel handling building provide assurance that: a Required features needed that orovide core coolino te+r++iefie are available fer the irradiated fuel in the eere in ease ef an inadvertent draindewn ef the reaeter vessel;

b. Required features used to mitigate a fuel handling accident involving handling of recently irradiated fuel are available (due to radioactive decay, handling of fuel only requires OPERABILITY of the Distribution Systems when the fuel being handled is recently irradiated, i.8., fuel that has occupied part of a critical reactor core within the previous 24 hours);

c Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and (continued) PERRY - UNIT 1 B 3.8-82 Revision No. 4

TS BASES MARK.UP. PROVIDED Distribution Systems - Shutdown FOR INFORMATION ONLY B 3.8.8 BASES (continued) ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply. lf moving recently irradiated fuel assemblies while in [/ODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require reactor shutdown. A.1, A.2.1, A.2.2. A.2.3, andA.2.4; ond 4,2,5 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS

                  @, movement of recently irradiated fuel.@

By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the Required Actions of the LCOs for these associated required features. Since this option may involve undesired administrative effotts, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS;d movement of recently irradiated fuel assemblies in the primary containment and fuel handling building+nC Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. lt is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems. Notwithstanding pedormance of the above conservative Required Actions, a required residual heat removal - shutdown cooling (RHR-SDC) subsystem may be inoperable. ln this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS (continued) PERRY - UNIT 1 B 3.8-83 Revision No. 2

TS BASES MARK-UP. PROVIDED Distribution Systems - Shutdown FOR INFORMATION ONLY B 3.8.8 BASES ACTIONS A.1, A.2.1, A.2.2. A.2.3" and A,2.4'+nC-A.2,5 (continued) would not be entered. Therefore, Required Action A.2.{5 is provided to direct declaring the associated required shutdown cooling subsystems inoperable, and not in operation, 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 AC and DC electrical power 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 REQUIREIVIENTS This Surveillance verifies that the required AC and DC electrical power distribution subsystems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures that the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper voltage availability on the required buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses. Additionally, when the Fuel Handling Building Ventilation Exhaust System is not required to be OPERABLE per LCO 3.7.9, "Fuel Handling Building Ventilation Exhaust System," 480 MCC EF-2-D-11 is not required to be energized to satisfy the requirements of this Surveillance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. REFERENCES 1 USAR, Chapter 6. 2 USAR, Chapter 15. PERRY - UNIT 1 B 3.8-84 Revision No. {4 I

TS BASES MARK-UP . PROVIDED lnservice Leak and Hydrostatic Testing Operation FOR INFORMATION ONLY B 3.10.1 BASES APPLICABLE coolant activity above the Iimits of LCO 3.4.8, "Reactor Coolant System SAFETY (RCS) Specific Activity," are minimized. ln addition, the primary ANALYSES containment will be OPERABLE, in accordance with this Special (continued) Operations LCO, and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated loss of coolant accidents inside of primary containment described in Reference 2. Therefore, these requirements will conservatively limit radiation releases to the environment. ln the event of a large primary system leak, the reactor vessel would rapidly depressurize, allowing the low pressure core cooling systems to operate. The Sakes.o_capability ef the Iew Bressure eeelant injeetien and required in MODE 4 by LCO 3.5.2, "Reactor Pressur Controlffi," would be more than adequate to keep the EPV water level above t under this low decay heat load condition. Small system leaks would be detected by Ieakage inspections before significant inventory loss occurred. For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the primary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and during postulated accident conditions. As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Final Policy Statement on Technical Specification lmprovements (58 FR 39132) apply. Special Operations LCOs provide flexibility to perform ceilain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. (continued) PERRY - UNIT 1 B 3.1 0-2 Revision No. 7}}

v t e Letter Sequence Request
EPID:L-2017-LLA-0405, TSTF-542 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers 24 July 2018 Results Approval Other: ML19072A012
MONTHYEAR2018-12-12 [Table View]

L-17-045, Application to Revise Technical Specifications to Adopt TSTF-542. Reactor Pressure Vessel Water Inventory Control.

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SUBJECT:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

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L-17-045, Application to Revise Technical Specifications to Adopt TSTF-542. Reactor Pressure Vessel Water Inventory Control.

Text

SUBJECT:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

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