Text

Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control
	ML17319A898
Person / Time
Site:	River Bend
Issue date:	11/15/2017
From:	Maguire W; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RBG-47801
	Download: ML17319A898 (149)
	v • d • e

C-:-U Entergx RBG-47801 November 15, 2017 U. S. Nuclear Regulatory Commission AnN: Document Control Desk Washington, DC 20555

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47

Dear Sir or Madam:

Entergy Operations, Inc.

River Bend Stallon 5485 U.S. Highway 61N

51. Francisville, LA 70775 Tel 225-381-4374 William F. Maguire Site Vice President Pursuant to 10 CFR 50.90, Entergy Operations, Inc. is submitting a request for an amendment to the Technical Specifications (TS) for River Bend Station, Unit l. The proposed change replaces existing TS requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.l.l.3. Safety Limit 2.l.l.3 requires reactor vessel water level to be greater than the top of active irradiated fuel. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides, for information only, the existing TS Bases pages marked to show the proposed changes.

Approval of the proposed amendment is requested by September 15, 2018. Once approved, the amendment shall be implemented within 120 days. In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated louisiana Official.

If you have any questions, please contact Mr. Tim Schenk at 225-381-4177. It is recognized that NRC Enforcement Guidance Memorandum 11-003, Rev. 3 wi" remain in effect until the requested amendment is implemented. This document contains no commitments.

I declare under penalty of perjury that the foregoing is true and correct, Executed on November 15, 2017.

Sincerely, WFMjdhw Attachments: l. DeSCription and Assessment of the Proposed Changes

2. Proposed Technical Specification Changes (Mark-up)

3. Proposed Technical Specification Bases Changes (Mark-up)

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DescriptionandAssessmentoftheProposedChanges

DESCRIPTIONandASSESSMENT

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 Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

2.0

Assessment

2.1

ApplicabilityofPublishedSafetyEvaluation

Entergy Operations, Inc. (EOI) has reviewed the safety evaluation provided to the Technical Specification Task Force (TSTF) on December 20, 2016, as well as the information provided in TSTF-542. EOI has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the NRC staff is applicable to River Bend Station - Unit 1 (RBS) and justify this amendment for the incorporation of the changes to the RBS TS.

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

3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation 3.3.7.1 Control Room Fresh Air (CRFA) System Instrumentation 3.5.2 ECCS - Shutdown 3.6.1.2 Primary Containment Air Locks 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 3.6.1.10 Primary Containment-Shutdown 3.7.2 Control Room Fresh Air (CRFA) System 3.7.3 Control Room Air Conditioning (AC) System 3.8.2 AC Sources - Shutdown 3.8.5 DC Sources - Shutdown 3.8.8 Inverters - Shutdown 3.8.10 Distribution Systems - Shutdown

2.2

Variations

EOI is proposing variations from the TS changes described in the TSTF-542 or the applicable parts of the NRC staffs safety evaluation. These variations do not affect the applicability of TSTF-542 or the NRC staff's safety evaluation to the proposed license amendment.

RBS is a BWR/6 plant. The proposed variations are based on the TSTF-542 markup of NUREG-1434 without a Setpoint Control Program.

The RBS TS do not contain a Surveillance Frequency Control Program. Therefore, the references to a Surveillance Frequency Control Program for Specification 3.5.2 are not included in the updated TS.

2.2.1

AdministrativeVariations

2.2.1.1 The RBS TS utilize different numbering and titles than the NUREG-1434 Improved Standard Technical Specifications (ISTS) on which TSTF-542 was based.

The following table relates the administrative differences between the TS described in the TSTF-542 and the RBS TS:

TSTF-542 TS Number TSTF-542 TS Title RBS TS Number RBS TS Title 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation New TS New TS 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation 3.3.5.2

[Changed to 3.3.5.3]

Same 3.3.6.1 Primary Containment Isolation Instrumentation Same Primary Containment and Drywell Isolation Instrumentation 3.3.6.2 Secondary Containment Isolation Instrumentation Same Secondary Containment and Fuel Building Isolation Instrumentation 3.6.4.1

[Secondary]

Containment Same Secondary Containment -

Operating 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

Same Secondary Containment Isolation Dampers (SCIDs) and Fuel Building Isolation Dampers (FBIDs) 3.7.3 Control Room Fresh Air (CRFA) System 3.7.2 Same 3.7.4

[ Control Room Air Conditioning (AC) ]

System 3.7.3 Same

2.2.1.2 The RBS TS contain administrative variations to the NUREG-1434 ISTS related to the numbering/labeling of sections and notes within the TS. The following table will relate the pertinent RBS TS sections and notes to the corresponding NUREG-1434 ISTS sections:

TSTF-542 TS RBS TS Description 3.3.5.1 3.3.5.1 Function 1.d of the RBS TS Table 3.3.5.1-1 corresponds with Function 1.c of the NUREG-1434 ISTS. This function is changed in accordance with the TSTF-542 markup.

3.3.5.1 3.3.5.1 Footnote (c) of the RBS TS Table 3.3.5.1-1 corresponds with footnote (e) of the NUREG-1434 ISTS. Footnote (c) of the RBS TS is removed in accordance with the TSTF-542 markup.

3.3.5.1 3.3.5.1 Footnote (d) of the RBS TS Table 3.3.5.1-1 corresponds with footnote (f) of the NUREG-1434 ISTS. The RBS footnote is updated similar to the TSTF-542 ISTS markup to update the footnote numbering.

3.3.5.1 3.3.5.1 Functions 1.c and 2.d listed in the RBS TS Table 3.3.5.1-1 are not included in the ISTS. The inclusion of these functions in the RBS TS offset the numbering of the remaining functions in this table from the ISTS.

3.3.5.1 &

3.3.5.2 3.3.5.1 &

3.3.5.2 Functions 1.e, Reactor Vessel Pressure-Low (Injection Permissive) and Function 2.e, Reactor Vessel Pressure-Low (Injection Permissive) of Table 3.3.5.1-1 of the RBS TS correspond with Function 1.d, Reactor Steam Dome Pressure -

Low (Injection Permissive) and Function 2.d, Reactor Steam Dome Pressure - Low (Injection Permissive) of the ISTS. The functions of these instruments are identical to the corresponding instruments in the ISTS and differ in name only; therefore, the existing RBS nomenclature is retained in RBS TS Function 1.a and 2.a of the proposed TS Table 3.3.5.2-1.

3.3.6.1 3.3.6.1 Footnote (b) of the NUREG-1434 Table 3.3.6.1-1 does not have a corresponding footnote in RBS TS Table 3.3.6.1-1. Therefore the change to this footnote does not apply to the RBS TS.

3.3.6.1 3.3.6.1 Footnote (e) of RBS TS Table 3.3.6.1-1 corresponds with footnote (c) of the NUREG-1434 ISTS. Footnote (e) of the RBS TS is removed in accordance with the TSTF-542 markup.

3.3.6.1 3.3.6.1 Function 5.b of the RBS TS Table 3.3.6.1-1 corresponds with Function 5.c of the NUREG-1434 ISTS. The RBS function is updated similar to the TSTF-542 function.

3.6.1.3 3.6.1.3 RBS TS 3.6.1.3 Condition F corresponds with NUREG-1434 ISTS Condition H.

3.8.5 3.8.5 RBS TS 3.8.5 Condition A corresponds with NUREG-1434 ISTS Condition B. This condition is changed in accordance with the TSTF-542 markup.

2.2.1.3 Note: Footnote (a) of Table 3.3.5.1-1 of the RBS TS contains different wording than the corresponding footnote (a) in the ISTS Table 3.3.5.1-1. This footnote is removed in alignment with the TSTF-542 markup.

2.2.1.4 The Allowable Values listed in the new RBS TS Table 3.3.5.2-1 for functions 1.b, 1.c, 2.b, and 3d are different from what is shown in the TSTF-542 TS markup. The TSTF-542 document format has a maximum and minimum value for pump flow while the RBS version has only a minimum flow rate. According to the TSTF-542 document, the allowable values for 3.3.5.2 Functions 1.b, 1.c, 2.b, and 3.d are unchanged from the existing TS requirements. The existing RBS TS requirements in table 3.3.5.1-1 only have a minimum flow rate for these functions. This is not a deviation but is specified here for clarification.

2.2.1.5 RBS TS 3.3.6.1 required action J corresponds to the TSTF-542 required action K. RBS TS 3.3.6.1 does not contain an equivalent to the NUREG-1434 required action K.2.2 to initiate action to suspend OPDRVs, so this change is not implemented in the RBS TS. Because this action does not exist in the RBS TS, this is an acceptable administrative variance from the TSTF-542 markup.

2.2.1.6 No changes were required to RBS TS 3.3.6.2, Secondary Containment and Fuel Building Isolation Instrumentation, because there are no existing OPDRV related requirements.

This deviation is necessary because none of the items changed in the ISTS TS 3.3.6.2 exist in the current RBS TS 3.3.6.2. Secondary Containment Isolation Instrumentation is not required during OPDRV activities at RBS because [Secondary Containment] as described in the ISTS refers to the RBS Primary Containment (Refer to note 1 on page 6 of TSTF-542 revision 1).

RBS TS 3.6.1.10 required that Primary Containment be operable during OPDRVs rather than Secondary Containment.

2.2.1.6 Note (a) of RBS TS Table 3.3.7.1-1 does not apply to Function 1, Reactor Vessel Water Level - Low. The change proposed by the TSTF-542 markup of the NUREG-1434 ISTS to this function to remove this note from the applicable modes column is not necessary.

2.2.1.7 The RBS TS 3.5.2 SR 3.5.2.5 (updated to be SR 3.5.2.6) will not include the wording, through the recirculation line that is provided in the TSTF-542 markup of SR 3.5.2.6.

This wording is not implemented to avoid confusion that the ECCS systems must be operated through the Reactor Recirculation lines rather than the ECCS test return lines.

2.2.1.8 RBS TS 3.5.2 does not contain the Note (LCO 3.0.4.b is not applicable to RCIC) below the ACTIONS of LCO 3.5.2 that is removed in the ISTS markup. This variation is administrative as this note is removed in the ISTS markup but is not included in the current RBS TS, so this change is not necessary.

2.2.1.9 The RBS TS contains a Note in TS 3.5.2 Surveillance Requirement 3.5.2.4 regarding realignment to the Low Pressure Coolant Injection mode. However, this note is located in the LCO statement of ISTS 3.5.2. The note is updated in the RBS TS to reflect the changes made by the TSTF-542 ISTS markup; however the location of the note is maintained in the Surveillance Requirement. This is not a deviation from the TSTF-542 document but is specified for clarification.

2.2.1.10 The change made by TSTF-542 to TS Surveillance Requirement 3.5.2.2 includes changing the word the to the word a referring to the required High Pressure Core Spray (HPCS) system. The RBS TS will not include this wording change as RBS has only one HPCS system.

2.2.2 Technical Variations 2.2.2.1 Table 3.3.5.1-1 of the RBS TS currently contains requirements for Function 1.c, LPCS Pump Start-Time Delay Relay and for Function 2.d, LPCI Pump C Start-Time Delay Relay that are not included in the ISTS. In accordance with the justification included in TSTF-542, applicability to modes 4 and 5 are removed in accordance with the evaluation provided in TSTF-542. These functions are no longer required in modes 4 and 5 due to the relatively slow transient of unexpected drain events. Sufficient time is permitted for operators to mitigate such a transient.

2.2.2.2 Function 2.e listed in table 3.3.5.1-1, Reactor Vessel Pressure - Low (Injection Permissive), is listed as applicable in modes 4 and 5 while the corresponding ISTS Function, 2.d, Reactor Steam Dome Pressure - Low (Injection Permissive) is not. In accordance with the justification included in TSTF-542, applicability to modes 4 and 5 are deleted because the instrumentation requirements during shutdown are being consolidated into the new TS 3.3.5.2.

2.2.2.3 Table 3.3.5.2-1 is revised to reflect the RBS 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-542 are not included in the proposed Technical Specifications. This corrects an error in TSTF-542 that affects the BWR/5 and BWR/6 ECCS instrumentation requirements.

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 8 signal prevents overfilling of the reactor vessel into the main steam lines by closing the HPCS injection valves when the water level is above the Level 8 setpoint. Therefore, if HPCS is the required ECCS subsystem and the water level is above Level 8, manually actuating Function 3.e will not inject inventory into the reactor vessel. This is not the desired response. If the Level 8 function is retained in Table 3.3.5.2-1, the function would need to be rendered inoperable in order to inject water when above the Level 8 water level. This would not be consistent with including the function in Table 3.3.5.2-1.

RBS 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. If 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. If the water level is above Level 8, then manual override of the Level 8 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 HPCS Level 8 isolation is already part of the BWR 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 8 isolation override feature) is tested.

2.2.2.4 Required Actions C.3 and D.4 listed in the ISTS LCO 3.5.2 are not applicable to RBS. The Standby Gas Treatment Subsystem is external to the primary containment and is not required for the operability of primary containment. Therefore, Required Actions C.3 and D.4 will not be incorporated into the RBS TS.

2.2.2.5 Required Action D.3 listed in the current RBS TS LCO 3.5.2 will be removed because the ISTS action D.2 to establish primary containment will require that the primary containment air lock boundary be secured for completion. This action is no longer necessary.

2.2.2.6 RBS TS LCO 3.6.1.2 Primary Containment Air Locks is currently applicable during OPDRV activities. This Applicability statement is removed along with the Condition E Required Action E.2 to suspend OPDRVs. The new drain time requirements of TS LCO 3.5.2 will preclude RPV Water Inventory Control activities with Drain Time <36 hours when at least one door in each Primary Containment Air Lock cannot be closed to restore Primary Containment. No additional Primary Containment Air Lock requirements are required.

2.2.2.7 RBS TS 3.6.4.1, Secondary Containment-Operating, is not affected by this TS change because there are no current OPDRV requirements for secondary containment.

[Secondary Containment] in the ISTS refers to the RBS Primary Containment. Rather, RBS TS 3.6.1.10 Primary Containment-Shutdown is revised to remove OPDRV LCO Applicability requirement similar to the ISTS markup of TS 3.6.4.1.

2.2.2.8 RBS TS 3.6.4.2, Secondary Containment Isolation Dampers (SCIDs) and Fuel Building Isolation Dampers (FBIDs), like TS 3.6.4.1 above are not affected by this TS change because there are no current OPDRV requirements. The current RBS Secondary Containment Isolation TS does not contain OPDRV requirements because [Secondary Containment] as described in the ISTS refers to the RBS Primary Containment. The OPDRV requirements similar to those listed in the current ISTS for TS 3.6.4.2 are correlated with the current RBS TS requirements in TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs). RBS TS 3.6.1.3 is revised to remove applicability to modes 4 and 5 because the applicability of LCO 3.3.6.1 Function 5.b during modes 4 and 5 has been eliminated in accordance with the TSTF-542 markup. Condition F of TS 3.6.1.3 has also been removed in its entirety as there are no longer any required actions for PCIVs to be operable during Mode 4 or 5. Primary containment operability requirements during modes 4 and 5 are established by new RBS TS 3.5.2.

2.2.2.9 RBS TS 3.6.4.3, Standby Gas Treatment (SGT) System, is not affected by this TS change because there are no current OPDRV requirements related to the SGT system. The SGT System is not required during RPV Water Inventory Control activities because the Standby Gas Treatment Subsystem is external to the Primary Containment and is not required for Primary Containment Operability.

3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Analysis Entergy Operations, Incorporated (EOI) requests adoption of TSTF-542 "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications (ISTS), into the River Bend Station (RBS) Technical Specifications (TS). The proposed amendment replaces the existing requirements in the TS related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

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

1.

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

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

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

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

The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

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

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

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

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed change will not alter the design function of the equipment involved. Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

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

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

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

Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

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

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

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

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TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (continued)

RIVER BEND ii Revision No. 6-5 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)......... B 3.2-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)......................... B 3.2-5 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR).......................... B 3.2-9 B 3.2.4 FRACTION OF CORE BOILING BOUNDARY (FCBB).................... B 3.2-12 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation............. 3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation.................. 3.3-10 3.3.1.3 Period Based Detection System (PBDS)........................ 3.3-14a 3.3.2.1 Control Rod Block Instrumentation........................... 3.3-15 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation.............. 3.3-19 3.3.3.2 Remote Shutdown System...................................... 3.3-23 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation........................................ 3.3-25 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation................... 3.3-29 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation........ 3.3-32 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation........................................ 3.3-44 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation........................................ 3.3-48 3.3.6.2 Secondary Containment and Fuel Building Isolation Instrumentation......................................... 3.3-58 3.3.6.3 Containment Unit Cooler System Instrumentation.............. 3.3-62 3.3.6.4 Relief and Low-Low Set (LLS) Instrumentation................ 3.3-66 3.3.7.1 Control Room Fresh Air (CRFA) System Instrumentation........ 3.3-68 3.3.8.1 Loss of Power (LOP) Instrumentation......................... 3.3-72 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring............................................. 3.3-75 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation................. B 3.3-1 B 3.3.1.2 Source Range Monitor (SRM) Instrumentation...................... B 3.3-32 B 3.3.1.3 Period Based Detection System (PBDS)............................ B 3.3-39a B 3.3.2.1 Control Rod Block Instrumentation............................... B 3.3-40 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation.................. B 3.3-49 B 3.3.3.2 Remote Shutdown System.......................................... B 3.3-60 B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation........................................ B 3.3-65 B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation................... B 3.3-76 B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation............ B 3.3-85 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation........................................ B 3.3-122 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation....................................................................3.3-43a 3

B 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Instrumentation............................................................................. B 3.3 121a 3

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS RIVER BEND iv Revision No. 167 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating................................................................................ 3.5-1 3.5.2 ECCS - Shutdown................................................................................ 3.5-6 3.5.3 RCIC System....................................................................................... 3.5-10 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating................................................................................ B 3.5-1 B 3.5.2 ECCS - Shutdown................................................................................ B 3.5-15 B 3.5.3 RCIC System....................................................................................... B 3.5-20 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 Isolation Valves (PCIVs)................................... 3.6-9 3.6.1.4 Primary Containment Pressure........................................................... 3.6-21 3.6.1.5 Primary Containment Air Temperature................................................ 3.6-22 3.6.1.6 Low - Low Set (LLS) Valves................................................................. 3.6-23 3.6.1.7 Primary Containment Unit Coolers...................................................... 3.6-25 3.6.1.8 DELETED............................................................................................ 3.6-27 3.6.1.9 Main Steam - Positive Leakage Control System (MS - PLCS)............. 3.6-29 3.6.1.10 Primary Containment - Shutdown........................................................ 3.6-31 3.6.2.1 Suppression Pool Average Temperature............................................ 3.6-33 3.6.2.2 Suppression Pool Water Level............................................................ 3.6-36 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling................. 3.6-37 3.6.3.1 DELETED............................................................................................3.6-39 3.6.3.2 Primary Containment and Drywell Hydrogen Igniters.......................... 3.6-41 3.6.3.3 Primary Containment/Drywell Hydrogen Mixing System..................... 3.6-44 3.6.4.1 Secondary Containment - Operating.................................................... 3.6-46 3.6.4.2 Secondary Containment Isolation Dampers (SCIDs) and Fuel Building Isolation Dampers (FBIDs)............................................ 3.6-48 3.6.4.3 Standby Gas Treatment (SGT) System............................................... 3.6-51 3.6.4.4 DELETED............................................................................................ 3.6-53 3.6.4.5 Fuel Building........................................................................................ 3.6-55 3.6.4.6 DELETED............................................................................................ 3.6-56 3.6.4.7 Fuel Building Ventilation System - Fuel Handling................................ 3.6-58 3.6.5.1 Drywell................................................................................................. 3.6-60 3.6.5.2 Drywell Air Lock................................................................................... 3.6-63 3.6.5.3 Drywell Isolation Valves....................................................................... 3.6-67 3.6.5.4 Drywell Pressure................................................................................. 3.6-71 3.6.5.5 Drywell Air Temperature...................................................................... 3.6-72 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment - Operating........................................................ B 3.6-1 B 3.6.1.2 Primary Containment Air Locks........................................................... B 3.6-5 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs)................................... B 3.6-15 B 3.6.1.4 Primary Containment Pressure........................................................... B 3.6-30 (continued)

, REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL Reactor Pressure Vessel (RPV) Water Inventory Control

Definitions 1.1 1.1 Definitions (continued)

RIVER BEND 1.0-2 Amendment No. 81, 132 CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, display, and trip functions, and channel failure trips. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.

The following exceptions are not considered to be CORE ALTERATIONS:

a.

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

b.

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

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

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

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR) 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 1989.

(continued)

Insert "A"

1.02a (InsertA)

TS1.1 DRAIN TIME The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming:

a)

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

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

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

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

3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation 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; d)

No additional draining events occur; and e)

Realistic cross-sectional areas and drain rates are used.

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

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-32 Amendment No. 81 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.5.1-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more channels inoperable.

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

Immediately B.

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

B.1

NOTES----------

1.

Only applicable in MODES 1, 2, and 3.

2.

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

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

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

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-33 Amendment No. 81 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.2

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

1.

Only applicable in MODES 1, 2, and 3.

2.

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

Declare High Pressure Core Spray (HPCS)

System inoperable.

AND B.3 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCS initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.

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

C.1

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

1.

Only applicable in MODES 1, 2, and 3.

2.

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

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

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

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-35 Amendment No. 81 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E.

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

E.1

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

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

2. Only applicable for Functions 1.f, 1.g, and 2.f.

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

AND E.2 Restore channel to OPERABLE status.

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

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

F.1 Declare Automatic Depressurization System (ADS) valves inoperable.

AND 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of ADS initiation capability in both trip systems (continued)

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-39 Amendment No. 81 Table 3.3.5.1-1 (page 1 of 5)

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

1.

Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems

a.

Reactor Vessel Water Level

Low Low Low, Level 1 1,2,3, 4(a),5(a) 2(b)

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t -147 inches

b.

Drywell PressureHigh 1,2,3 2(b)

B SR 3.3.5.1.1 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 1.88 psid

c.

LPCS Pump StartTime Delay Relay 1,2,3, 4(a),5(a) 1 C

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 1.8 seconds and d 2.2 seconds

d.

LPCI Pump A StartTime Delay Relay 1,2,3, 4(a),5(a) 1 C

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 6.3 seconds and d 7.7 seconds

e.

Reactor Vessel PressureLow (Injection Permissive) 1,2,3 4

C SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 472 psig and d 502 psig 4(a),5(a) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 472 psig and d 502 psig

f.

LPCS Pump Discharge FlowLow (Bypass) 1,2,3, 4(a),5(a) 1 E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 750 gpm

g.

LPCI Pump A Discharge FlowLow (Bypass) 1,2,3, 4(a),5(a) 1 E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 900 gpm

h.

Manual Initiation 1,2,3, 4(a),5(a) 1 per system C

SR 3.3.5.1.6 NA (continued)

(a)

When associated subsystem(s) are required to be OPERABLE.

(b)

Also required to initiate the associated diesel generator.

a a

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-40 Amendment No. 81 Table 3.3.5.1-1 (page 2 of 5)

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

2.

LPCI B and LPCI C Subsystems

a.

Reactor Vessel Water LevelLow Low Low, Level 1 1,2,3, 4(a),5(a) 2(b)

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t -147 Inches

b.

Drywell PressureHigh 1,2,3 2(b)

B SR 3.3.5.1.1 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 1.88 psid

c.

LPCI Pump B StartTime Delay Relay 1,2,3, 4(a),5(a) 1 C

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 6.3 seconds and d 7.7 seconds

d.

LPCI Pump C StartTime Delay Relay 1,2,3, 4(a),5(a) 1 C

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 1.8 seconds and d 2.2 seconds

e.

Reactor Vessel PressureLow (Injection Permissive) 1,2,3 4

C SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 472 psig and d 502 psig 4(a),5(a) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 472 psig and d 502 psig

f.

LPCI Pump B and LPCI Pump C Discharge FlowLow (Bypass) 1,2,3, 4(a),5(a) 1 per pump E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 900 gpm

g.

Manual Initiation 1,2,3, 4(a),5(a) 1 C

SR 3.3.5.1.6 NA (continued)

(a)

When associated subsystem(s) are required to be OPERABLE.

(b)

Also required to initiate the associated diesel generator.

a a

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-41 Amendment No. 81 Table 3.3.5.1-1 (page 3 of 5)

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

3.

High Pressure Core Spray (HPCS) System

a.

Reactor Vessel Water LevelLow Low, Level 2 1,2,3, 4(a),5(a) 4(b)

B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t -47 inches

b.

Drywell Pressure High 1,2,3 4(b)

B SR 3.3.5.1.1 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 1.88 psid

c.

Reactor Vessel Water LevelHigh, Level 8 1,2,3, 4(a),5(a) 2 C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 d 55 inches

d.

Condensate Storage Tank LevelLow 1,2,3, 4(c),5(c) 2 D

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t -4.5 inches

e.

Suppression Pool Water LevelHigh 1,2,3 2

D SR 3.3.5.1.1 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 8 inches

f.

HPCS Pump Discharge PressureHigh (Bypass) 1,2,3, 4(a),5(a) 1 E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 275 psig

g.

HPCS System Flow RateLow (Bypass) 1,2,3, 4(a),5(a) 1 E

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 710 gpm

h.

Manual Initiation 1,2,3, 4(a),5(a) 1 C

SR 3.3.5.1.6 NA (continued)

(a)

When associated subsystem(s) are required to be OPERABLE.

(b)

Also required to initiate the associated diesel generator.

(c)

When HPCS is OPERABLE for compliance with LCO 3.5.2, "ECCSShutdown," and aligned to the condensate storage tank while tank water level is not within the limit of SR 3.5.2.2.

a a

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-42 Amendment No. 81 Table 3.3.5.1-1 (page 4 of 5)

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

4.

Automatic Depressurization System (ADS) Trip System A

a.

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

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t -147 inches

b.

Drywell PressureHigh 1,2(d),3(d) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 d 1.88 psid

c.

ADS Initiation Timer 1,2(d),3(d) 1 G

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 117 seconds

d.

Reactor Vessel Water LevelLow, Level 3 (Confirmatory) 1,2(d),3(d) 1 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 8.7 inches

e.

LPCS Pump Discharge Pressure High 1,2(d),3(d) 2 G

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 130 psig

f.

LPCI Pump A Discharge PressureHigh 1,2(d),3(d) 2 G

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 120 psig

g.

ADS Bypass Timer (High Drywell Pressure) 1,2(d),3(d) 2 G

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 5.5 minutes

h.

Manual Initiation 1,2(d),3(d) 2 per system G

SR 3.3.5.1.6 NA (continued)

(d)

With reactor steam dome pressure > 100 psig.

(Re-number all Notes on this page from "d" to "b".

ECCS Instrumentation 3.3.5.1 RIVER BEND 3.3-43 Amendment No. 81 Table 3.3.5.1-1 (page 5 of 5)

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

5.

ADS Trip System B

a.

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

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t -147 Inches

b.

Drywell PressureHigh 1,2(d),3(d) 2 F

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 d 1.88 psid

c.

ADS Initiation Timer 1,2(d),3(d) 1 G

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 117 seconds

d.

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

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 8.7 Inches

e.

LPCI Pumps B & C Discharge PressureHigh 1,2(d),3(d) 2 per pump G

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 t 120 psig

f.

ADS Bypass Timer (High Drywell Pressure) 1,2(d),3(d) 2 G

SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 5.5 minutes

g.

Manual Initiation 1,2(d),3(d) 2 per system G

SR 3.3.5.1.6 NA (continued)

(d)

With reactor steam dome pressure > 100 psig.

(Re-number all Notes on this page from "d" to "b".

RPV Water Inventory Control Instrumentation 3.3.5.2 RIVER BEND 3.3-43a Amendment No. xxx 3.3 INSTRUMENTATION 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.2 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.5.2-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more channels inoperable.

A.1 Enter the Condition referenced in Table 3.3.5.2-1 for the channel.

Immediately B.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

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

AND B.2 Calculate DRAIN TIME.

Immediately Immediately C.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

C.1 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (continued)

RPV Water Inventory Control Instrumentation 3.3.5.2 RIVER BEND 3.3-43b Amendment No. xxx ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

D.1 Declare HPCS system inoperable OR D.2 Align the HPCS pump suction to the suppression pool.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour E.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

E.1 Declare HPCS system inoperable AND E.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 24 hours F.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

F.1 Restore channel to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months G. Required Action and associated Completion Time of Condition C, D, E, or F not met.

G.1 Declare associated ECCS injection/spray subsystem inoperable.

Immediately

RPV Water Inventory Control Instrumentation 3.3.5.2 RIVER BEND 3.3-43c Amendment No.xxx SURVEILLANCE REQUIREMENTS

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

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

SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.5.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST.

24 months

RIVER BEND 3.3-43d Amendment No.xxx Table 3.3.5.2-1 (page 1 of 2)

RPV Water Inventory Control Instrumentation

FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCE FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

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

a.

Reactor Vessel Pressure - Low (Injection Permissive) 4, 5 4(a)

C SR 3.3.5.2.1 SR 3.3.5.2.2 502 psig

b.

LPCS Pump Discharge Flow -

Low (Bypass) 4, 5 1(a)

F SR 3.3.5.2.1 SR 3.3.5.2.2 750 gpm

c.

LPCI Pump A Discharge Flow -

Low (Bypass) 4, 5 1(a)

F SR 3.3.5.2.1 SR 3.3.5.2.2 900 gpm

d.

Manual Initiation 4, 5 1 per system (a)

F SR 3.3.5.2.3 NA

2.

LPCI B and LPCI C Subsytems

a.

Reactor Vessel Pressure - Low (Injection Permissive) 4, 5 4(a)

C SR 3.3.5.2.1 SR 3.3.5.2.2 502 psig

b.

LPCI Pump B and LPCI Pump C Discharge Flow -

Low (Bypass) 4, 5 1 per pump (a)

F SR 3.3.5.2.1 SR 3.3.5.2.2 900 gpm

c.

Manual Initiation 4, 5 1 per system (a)

F SR 3.3.5.2.3 NA (a)

Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, Reactor Pressure Vessel Water Inventory Control.

RIVER BEND 3.3-43e Amendment No.xxx Table 3.3.5.2-1 (page 2 of 2)

RPV Water Inventory Control Instrumentation

FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCE FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

3.

High Pressure Core Spray (HPCS) System

a.

Condensate Storage Tank Level - Low 4(b), 5(b) 2(a)

D SR 3.3.5.2.1 SR 3.3.5.2.2

-4.5 inches

b.

HPCS Pump Discharge Pressure

- High (Bypass) 4, 5 1(a)

F SR 3.3.5.2.1 SR 3.3.5.2.2 275 psig

c.

HPCS System Flow Rate - Low (Bypass) 4, 5 1(a)

F SR 3.3.5.2.1 SR 3.3.5.2.2 710 gpm

4.

RHR System Isolation Reactor Vessel Water Level - Low (Level 3)

(c) 2 in one trip system B

SR 3.3.5.2.1 SR 3.3.5.2.2 8.7 inches

5.

Reactor Water Cleanup (RWCU) System Isolation

a.

Reactor Water Level

- Low, Low (Level

2)

(c) 2 in one trip system B

SR 3.3.5.2.1 SR 3.3.5.2.2

-47 inches (a)

Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, Reactor Pressure Vessel Water Inventory Control.

(b)

When HPCS is OPERABLE for compliance with LCO 3.5.2, RPV Water Inventory 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.

RCIC System Instrumentation 3.3.5.2 RIVER BEND 3.3-44 Amendment No. 81 3.3 INSTRUMENTATION 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.2 The RCIC System instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.

APPLICABILITY:

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

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more channels inoperable.

A.1 Enter the Condition referenced in Table 3.3.5.2-1 for the channel.

Immediately B.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

B.1 Declare RCIC System inoperable.

AND B.2 Place channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

C.1 Restore channel to OPERABLE status.

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

(replace all deleted characters on this page with "3")

RCIC System Instrumentation 3.3.5.2 RIVER BEND 3.3-45 Amendment No. 81 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

D.1

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

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

Declare RCIC System inoperable.

AND D.2.1 Place channel in trip.

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

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours E.

Required Action and associated Completion Time of Condition B, C, or D not met.

E.1 Declare RCIC System inoperable.

Immediately (replace all deleted characters on this page with "3")

RCIC System Instrumentation 3.3.5.2 RIVER BEND 3.3-46 Amendment No. 81, 168 SURVEILLANCE REQUIREMENTS

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

1.

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

2.

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

SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.5.2.3 Calibrate the trip units.

92 days SR 3.3.5.2.4 Perform CHANNEL CALIBRATION.

24 months SR 3.3.5.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST.

24 months (replace all deleted characters on this page with "3")

RCIC System Instrumentation 3.3.5.2 RIVER BEND 3.3-47 Amendment No. 81 Table 3.3.5.2-1 (page 1 of 1)

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

1.

Reactor Vessel Water LevelLow Low, Level 2 4

B SR 3.3.5.2.1 SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5 t -47 inches

2.

Reactor Vessel Water LevelHigh, Level 8 2

C SR 3.3.5.2.1 SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5 d 52 inches

3.

Condensate Storage Tank LevelLow 2

D SR 3.3.5.2.1 SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5 t -4.5 inches

4.

Suppression Pool Water LevelHigh 2

D SR 3.3.5.2.1 SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5 d 8 inches

5.

Manual Initiation 1

C SR 3.3.5.2.5 NA (replace all deleted characters on this page with "3")

Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 RIVER BEND 3.3-57 Amendment No. 81 Table 3.3.6.1-1 (page 5 of 5)

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

5.

RHR System Isolation

a.

RHR Equipment Room Ambient TemperatureHigh 1,2,3 2

F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6 d 121.1qF

b.

Reactor Vessel Water LevelLow, Level 3 1,2,3(c) 2 F

SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 t 8.7 Inches 3(d),4,5 2(e)

J SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 t 8.7 inches

c.

Reactor Vessel Water Level

Low Low Low, Level 1 1,2,3 2

F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 t -147 Inches

d.

Reactor Steam Dome Pressure

High 1,2,3 2

F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 d 150 psig

e.

Drywell Pressure High 1,2,3 2

F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 d 1.88 psid

f.

Manual Initiation 1,2,3 2

G SR 3.3.6.1.6 NA (c)

With reactor steam dome pressure greater than or equal to the RHR cut-in permissive pressure.

(d)

With reactor steam dome pressure less than the RHR cut-in permissive pressure.

(e)

Only one trip system required in MODES 4 and 5 with RHR Shutdown Cooling System integrity maintained.

CRFA System Instrumentation 3.3.7.1 RIVER BEND 3.3-71 Amendment No. 81 95 119, 132 Table 3.3.7.1-1 (page 1 of 1)

Control Room Fresh Air System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Reactor Vessel Water Level -

Low Low, Level 2 1,2,3 2

B SR 3.3.7.1.1 SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

-47 inches

2.

Drywell Pressure - High 1,2,3 2

C SR 3.3.7.1.1 SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5 1.88 psid

3.

Control Room Local Intake Ventilation Radiation Monitors 1,2,3 (a),(b) 1 D

SR 3.3.7.1.1 SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.5 0.97 x 10-5

µCi/cc (a)

During operations with a potential for draining the reactor vessel.

(ba)

During movement of recently irradiated fuel assemblies in the primary containment or fuel building.

ECCS Operating 3.5.1 RIVER BEND 3.5-1 Amendment No. 81, 156 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 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 seven 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 100 psig.

ACTIONS

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

LCO 3.0.4.b is not applicable to HPCS.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One low pressure ECCS injection/spray subsystem inoperable.

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

7 days B.

High Pressure Core Spray (HPCS) System inoperable.

B.1 Verify by administrative means RCIC System is OPERABLE when RCIC is required to be OPERABLE.

AND B.2 Restore HPCS System to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 14 days (continued)

, RPV WATER INVENTORY CONTROL,

ECCS Shutdown 3.5.2 RIVER BEND 3.5-6 Amendment No. 81 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.2 ECCS-Shutdown LCO 3.5.2 Two ECCS injection/spray subsystems shall be OPERABLE.

APPLICABILITY:

MODE 4, MODE 5 except with the upper containment fuel pool gate opened and water level 23 ft over the top of the reactor pressure vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One required ECCS injection/spray subsystem inoperable.

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

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

Required Action and associated Completion Time of Condition A not met.

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

Immediately C.

Two required ECCS injection/spray subsystems inoperable.

C.1 Initiate action to suspend OPDRVs.

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

Immediately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

, RPV WATER INVENTORY CONTROL, RPV Water Inventory Control Initiate action to establish a method of water injection capable of operating without offsite electrical power.

R MODES 4 and 5 Reactor Pressure Vessel (RPV) Water Inventory Control DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AND (Insert "B" here)

One

3.56a (Insert B)

TS 3.5.2 C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and > 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C.1 Verify primary containment boundary is capable of being established in less than the DRAIN TIME.

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

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. DRAIN TIME < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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

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

D.1 Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Immediately AND D.2 Initiate action to establish primary containment boundary.

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

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

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

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

Immediately

ECCS Shutdown 3.5.2 RIVER BEND 3.5-7 Amendment No. 81 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

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

D.1 Initiate action to restore primary containment to OPERABLE status.

AND D.2 Initiate action to isolate required primary containment penetration flow paths.

AND D.3

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

Entry and exit is permissible under administrative control.

Initiate action to close one door in each primary containment air lock.

Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure ECCS injection/spray subsystem, the suppression pool water level is 13 ft 3 inches.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued) a 2

RPV Water Inventory Control (Insert "C" here)

(InsertC)

TS3.5.2

SR3.5.2.1 VerifyDRAINTIMEis36hours.

12hours

ECCS Shutdown 3.5.2 RIVER BEND 3.5-8 Amendment No. 81, 188 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for the required High Pressure Core Spray (HPCS) System, the:

a.

Suppression pool water level is 13 ft 3 inches; or b.

Condensate storage tank water level is 11 ft 1 inch.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.5.2.3 Verify, for each required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

31 days SR 3.5.2.4

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

1.

One low pressure coolant injection (LPCI) subsystem may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned and not otherwise inoperable.

2.

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

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

31 days (continued) 3 4

5 the

, for the

, each RPV Water Inventory Control A

ECCS Shutdown 3.5.2 RIVER BEND 3.5-9 Amendment No. 81, 168 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the specified flow rate with the specified pump differential pressure.

PUMP DIFFERENTIAL SYSTEM FLOW RATE PRESSURE LPCS 5010 gpm 282 psid LPCI 5050 gpm 102 psid HPCS 5010 gpm 415 psid In accordance with the Inservice Testing Program SR 3.5.2.6

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

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.

24 months RPV Water Inventory Control 6

8 Operate the required ECCS injection/

spray subsystem for 10 minutes.

92 days (Insert "D")

Verify the required LPCI or LPCS subsystem actuates on a manual initiation signal, or the required HPCS subsystem can be manually operated.

(InsertD)

TS3.5.2

SR3.5.2.7 Verifyeachvalvecreditedforautomaticallyisolatinga penetrationflowpathactuatestotheisolationpositionon anactualorsimulatedactuationsignal.

24months

RCIC System 3.5.3 RIVER BEND 3.5-10 Amendment No. 81, 156 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY:

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

ACTIONS

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

LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A.

RCIC System inoperable.

A.1 Verify by administrative means High Pressure Core Spray System is OPERABLE.

AND A.2 Restore RCIC System to OPERABLE status.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 14 days B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

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

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

, RPV WATER INVENTORY CONTROL,

Primary Containment Air Locks 3.6.1.2 RIVER BEND 3.6-3 Amendment No. 81, 85 3.6 CONTAINMENT 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,.

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

ACTIONS

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

1.

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

2.

Separate Condition entry is allowed for each air lock.

2.

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 containment air locks with one primary containment air lock door inoperable.

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

1.

Required Actions A.1, A.2, and A.3 are not applicable if both 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)

Primary Containment Air Locks 3.6.1.2 RIVER BEND 3.6-6 Amendment No. 81, 85 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

(continued)

C.3 Restore air lock to OPERABLE status.

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

Required Action and associated Completion Time of Condition A, B, or C not met in MODE 1, 2, or

3.

D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

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

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

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

AND E.2 Initiate action to suspend OPDRVs.

Immediately Immediately

PCIVs 3.6.1.3 RIVER BEND 3.6-9 Amendment No. 81 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, MODES 4 and 5 for RHR Shutdown Cooling System suction from the reactor vessel isolation valves when associated isolation instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation,"

Function 5.b.

ACTIONS

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

1.

Penetration flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

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

4.

Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria, in MODES 1, 2, and 3.

(continued)

PCIVs 3.6.1.3 RIVER BEND 3.6-14 Amendment No. 81 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued)

D.3 Perform SR 3.6.1.3.5 for the resilient seal purge valves closed to comply with Required Action D.1.

Once per 92 days E.

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

E.1 Be in MODE 3.

AND E.2 Be in MODE 4.

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

Required Action and associated Completion Time of Condition A, B, C, or D not met for PCIV(s) required to be OPERABLE during MODE 4 or 5.

F.1 Initiate action to suspend OPDRVs.

OR F.2 Initiate action to restore valve(s) to OPERABLE status.

Immediately Immediately

Primary Containment-Shutdown 3.6.1.10 RIVER BEND 3.6-31 Amendment No. 81, 119 3.6 CONTAINMENT SYSTEMS 3.6.1.10 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.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Primary containment inoperable.

A.1 Suspend movement of recently irradiated fuel in the primary containment.

AND A.2 Initiate action to suspend OPDRVs.

Immediately Immediately

CRFA System 3.7.2 RIVER BEND 3.7-5 Amendment No. 81 119 132, 154 3.7 PLANT SYSTEM 3.7.2 Control Room Fresh Air (CRFA) System LCO 3.7.2 Two CRFA subsystems shall be OPERABLE.

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

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

APPLICABILITY:

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

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One CRFA subsystem inoperable for reasons other than Condition B.

A.1 Restore CRFA subsystem to OPERABLE status.

7 days B.

One or more CRFA subsystems inoperable due to inoperable CRE boundary in MODE 1, 2, or 3.

B.1 Initiate action to implement mitigating actions.

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

AND B.3 Restore CRE boundary to OPERABLE status.

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

CRFA System 3.7.2 RIVER BEND 3.7-6 Amendment No. 81 119 132 154 185 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

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

C.1

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

LCO 3.0.4.a is not applicable when entering MODE 3.

Be in MODE 3.

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

Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the primary containment or fuel building or during OPDRVs.

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

LCO 3.0.3 is not applicable.

D.1 Place OPERABLE CRFA subsystem in emergency mode.

OR D.2.1 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

AND D.2.2 Initiate action to suspend OPDRVs.

Immediately Immediately Immediately E.

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

E.1

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

LCO 3.0.4.a is not applicable when entering MODE 3.

Be in MODE 3.

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

CRFA System 3.7.2 RIVER BEND 3.7-7 Amendment No. 81 119 132 154 165 168, 183 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F.

Two CRFA subsystems inoperable during movement of recently irradiated fuel assemblies in the primary containment or fuel building, or during OPDRVs.

OR One or more CRFA subsystems inoperable due to inoperable CRE boundary during movement of recently irradiated fuel assemblies in the primary containment or fuel building, or during OPDRVs.

F.1 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

AND F.2 Initiate action to suspend OPDRVs.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Operate each CRFA subsystem for 15 continuous minutes.

31 days SR 3.7.2.2 Perform required CRFA filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.2.3 Verify each CRFA subsystem actuates on an actual or simulated initiation signal.

24 months (continued)

Control Room AC System 3.7.3 RIVER BEND 3.7-9 Amendment No. 81 119 132 185 3.7 PLANT SYSTEMS 3.7.3 Control Room Air Conditioning (AC) System LCO 3.7.3 Two control room AC subsystems shall be OPERABLE.

APPLICABILITY:

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

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One control room AC subsystem inoperable.

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

30 days B.

Two control room AC subsystems inoperable.

B.1 Verify control room area temperature 104°F.

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

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 7 days C.

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

C.1

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

LCO 3.0.4.a is not applicable when entering MODE 3.

Be in MODE 3.

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

Control Room AC System 3.7.3 RIVER BEND 3.7-10 Amendment No. 81 119, 132 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the primary containment or fuel building, or during OPDRVs.

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

LCO 3.0.3 is not applicable.

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

OR D.2.1 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

AND D.2.2 Initiate action to suspend OPDRVs.

Immediately Immediately Immediately (continued)

Control Room AC System 3.7.3 RIVER BEND 3.7-11 Amendment No. 81 119 132, 168 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E.

Required Action and associated Completion Time of Condition B not met during movement of recently irradiated fuel assemblies in the primary containment or fuel building, or during OPDRVs.

E.1 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

AND E.2 Initiate action to suspend OPDRVs.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify each control room AC subsystem has the capability to remove the assumed heat load.

24 months

AC SourcesShutdown 3.8.2 RIVER BEND 3.8-18 Amendment No. 81, 132 ACTIONS

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

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A.

LCO Item a not met.

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

Enter applicable Condition and Required Actions of LCO 3.8.10, when any required division is de-energized as a result of Condition A.

A.1 Declare affected required feature(s) with no offsite power available from a required circuit inoperable.

OR A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

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

AND Immediately Immediately Immediately Immediately (continued)

AC SourcesShutdown 3.8.2 RIVER BEND 3.8-19 Amendment No. 81, 132 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2.4 Initiate action to restore required offsite power circuit to OPERABLE status.

Immediately B.

LCO Item b not met.

B.1 Suspend CORE ALTERATIONS.

AND B.2 Suspend movement of recently irradiated fuel assemblies in primary containment and fuel building.

AND B.3 Initiate action to suspend OPDRVs.

AND B.4 Initiate action to restore required DG to OPERABLE status.

Immediately Immediately Immediately Immediately C.

LCO Item c not met.

C.1 Declare High Pressure Core Spray System and Standby Service Water System pump 2C inoperable.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 3

3

DC SourcesShutdown 3.8.5 RIVER BEND 3.8-29 Amendment No. 81, 132 ACTIONS

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

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required DC electrical power subsystems inoperable.

A.1 Declare affected required feature(s) inoperable.

OR A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

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

AND A.2.4 Initiate action to restore required DC electrical power subsystems to OPERABLE status.

Immediately Immediately Immediately Immediately Immediately 3

InvertersShutdown 3.8.8 RIVER BEND 3.8-36 Amendment No. 81, 132 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 InvertersShutdown LCO 3.8.8 One Divisional inverter shall be OPERABLE capable of supplying one division of the Division I or II onsite Class 1E uninterruptible AC vital bus electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems-Shutdown".

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in the primary containment or fuel building.

ACTIONS

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

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required inverters inoperable.

A.1 Declare affected required feature(s) inoperable.

OR A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend handling of recently irradiated fuel assemblies in the primary containment or fuel building.

AND Immediately Immediately Immediately (continued)

InvertersShutdown 3.8.8 RIVER BEND 3.8-37 Amendment No. 81 ACTIONS CONDITIONS REQUIRED ACTION COMPLETION TIME A.

(continued)

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

AND A.2.4 Initiate action to restore required inverters to OPERABLE status.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage, frequency, and alignments to required AC vital buses.

7 days 3

Distribution SystemsShutdown 3.8.10 RIVER BEND 3.8-41 Amendment No. 81, 132 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution SystemsShutdown LCO 3.8.10 The necessary portions of the Division I, Division II, and Division III AC, DC, and Division I and II AC vital bus electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 4 and 5, During movement of recently irradiated fuel assemblies in the primary containment or fuel building.

ACTIONS

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

LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required AC, DC, or AC vital bus electrical power distribution subsystems inoperable.

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

OR A.2.1 Suspend CORE ALTERATIONS.

AND A.2.2 Suspend movement of recently irradiated fuel assemblies in the primary containment and fuel building.

AND Immediately Immediately Immediately (continued)

Distribution SystemsShutdown 3.8.10 RIVER BEND 3.8-42 Amendment No. 81 ACTIONS CONDITIONS REQUIRED ACTIONS COMPLETION TIME A.

(continued)

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

AND A.2.4 Initiate actions to restore required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status.

AND A.2.5 Declare associated required shutdown cooling subsystem(s) inoperable and not in operation.

Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC, and AC vital bus electrical power distribution subsystems.

7 days 3

4

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ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-91 Revision No. 0 BACKGROUND Diesel Generators (continued)

Feature (ESF) buses if a loss of offsite power occurs.

(Refer to Bases for LCO 3.3.8.1.)

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

ECCS instrumentation satisfies Criterion 3 of the NRC Policy Statement.

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

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

Table 3.3.5.1-1, footnote (ba), is added to shows that certain ECCS instrumentation Functions are also required to be OPERABLE to perform DG initiation.

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

The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived (continued)

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-93 Revision No. 0 APPLICABLE 1.a, 2.a Reactor Vessel Water Level-Low Low Low, Level 1 SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Reactor Vessel Water Level-Low Low Low, Level 1 signals are initiated from four 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 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 only required to be OPERABLE when the associated ECCS or DG is required to be OPERABLE, to ensure that no single instrument failure can preclude ECCS initiation.

(Two channels input to LPCS and LPCI A, while the other two channels input to LPCI B and LPCI C.) Refer to LCO 3.5.1 and LCO 3.5.2, "ECCS-Shutdown," for Applicability Bases for the low pressure ECCS subsystems; LCO 3.8.1, "AC Sources-Operating"; and LCO 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs.

1.b, 2.b. Drywell Pressure-High High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell Pressure-High Function in order to minimize the possibility of fuel damage. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment.

Negative barometric fluctuations are accounted for in the Allowable Value.

The Drywell Pressure-High Function is required to be OPERABLE when the associated ECCS and DGs are required to be OPERABLE in conjunction with times when the primary (continued)

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-95 Revision No. 0 APPLICABLE 1.c, 1.d, 2.c, and 2.d. Low Pressure Coolant Injection Pump A, B, and C SAFETY ANALYSES, and LPCS Pump Start-Time Delay Relay (continued)

LCO, and APPLICABILITY 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 LPCS/LPCI Pump Start-Time Delay Relay Function is only required to be OPERABLE when the associated LPCS/LPCI subsystem is required to be OPERABLE. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the LPCI subsystems.

1.e, 2.e. Reactor Vessel Pressure-Low (Injection 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 is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. In addition, the Reactor Vessel Pressure-Low Function is directly assumed in the analysis of the recirculation line break (Ref. 2). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

The Reactor Vessel Pressure-Low signals are initiated from four pressure transmitters that sense the reactor dome pressure. The four pressure transmitters each drive a master and two slave trip units (for a total of eight slave trip units).

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.

Four channels of Reactor Vessel Pressure-Low Function per associated Division are required to be OPERABLE when the (continued)

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-96 Revision No. 6-12 APPLICABLE 1.e, 2.e. Reactor Vessel Pressure-Low (Injection Permissive)

SAFETY ANALYSES, (continued)

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

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

The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The LPCI and LPCS Pump Discharge Flow-Low Functions are assumed to be OPERABLE 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. 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 immediately open after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the RHR shutdown cooling mode (for RHR A and RHR B). The Pump Discharge Flow-Low Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.

Each channel of Pump Discharge Flow-Low Function (one LPCS channel and three LPCI channels) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE, to ensure that no single instrument failure can preclude the ECCS function. Refer to LCO 3.5.1 and (continued)

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-97 Revision No. 0 APPLICABLE 1.f, 1.g, 2.f. Low Pressure Coolant Injection and Low Pressure Core SAFETY ANALYSES, Spray Pump Discharge Flow-Low (Bypass) (continued)

LCO, and APPLICABILITY LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

1.h, 2.g. Manual Initiation The Manual 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 Initiation Function is not assumed in any accident or transient analyses in the USAR. However, the Function is retained for 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 Manual Initiation Function (one channel per Division) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

High Pressure Core Spray System 3.a. Reactor Vessel Water Level-Low Low, Level 2 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCS System and associated DG are 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 References 1 and 3. The Reactor Vessel Water Level-Low Low, Level 2 Function associated with HPCS is directly assumed in the analysis of the recirculation line break (Ref. 2). The core cooling (continued)

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-109 Revision No. 0 ACTIONS B.1, B.2, and B.3 (continued)

For Required Action B.2, redundant automatic initiation capability is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system. In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.3 is not appropriate and the feature(s) associated with the inoperable, untripped channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

As noted (Note 1 to Required Action B.1 and Required Action B.2), the two Required Actions are only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action B.3) is allowed during MODES 4 and 5. Notes are also provided (Note 2 to Required Action B.1 and Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable. This ensures that the proper loss of initiation capability check is performed.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action B.1, the Completion Time only begins upon discovery that a redundant feature in both Divisions (e.g., any Division 1 ECCS and Division 2 ECCS) cannot be automatically initiated due to inoperable, untripped channels within the same variable as described in the paragraph above. For Required Action B.2, the Completion Time only begins upon discovery that the HPCS System cannot be automatically initiated due to two inoperable, untripped channels for the associated Function in the same trip system. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable (continued)

A Note is also provided to

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-111 Revision No. 0 ACTIONS C.1 and C.2 (continued)

In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1), the Required Action is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of automatic initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Actions B.3 and C.2) is allowed during MODES 4 and 5.

Note 2 states that Required Action C.1 is only applicable for Functions 1.c, 1.d, 1.e, 2.c, 2.d, and 2.e. The Required Action is not applicable to Functions 1.h, 2.g, and 3.h (which also require entry into this Condition if a channel in these Functions is inoperable), since they are the Manual Initiation Functions and are not assumed in any accident or transient analysis. Thus, a total loss of manual initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action C.2) is allowed. Required Action C.1 is also not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic).

This loss was considered during the development of Reference 4 and considered acceptable for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed by Required Action C.2.

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

Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 4) to permit restoration of any (continued)

A

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-113 Revision No. 0 ACTIONS D.1, D.2.1, and D.2.2 (continued) allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per Required Action D.2.2. Placing the inoperable channel in trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to continue. If Required Action D.2.1 or Required Action D.2.2 is performed, measures should be taken to ensure that the HPCS System piping remains filled with water.

Alternately, if it is not desired to perform Required Actions D.2.1 and D.2.2, Condition H must be entered and its Required Action taken.

E.1 and E.2 Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped 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.g, and 2.f (e.g., low pressure ECCS). Redundant automatic initiation capability is lost if three of the four channels associated with Functions 1.f, 1.g, and 2.f are inoperable. Since each inoperable channel would have Required Action E.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected low pressure ECCS pump to be declared inoperable. However, since channels for more than one low pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable.

In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is not appropriate and the feature(s) associated with each inoperable channel must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after discovery of loss of initiation capability for feature(s) in both Divisions. As noted (Note 1 to Required Action E.1),

Required Action E.1 is only applicable in (continued)

ECCS Instrumentation B 3.3.5.1 BASES RIVER BEND B 3.3-114 Revision No. 0 ACTIONS E.1 and E.2 (continued)

MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 7 days (as allowed by Required Action E.2) is allowed during MODES 4 and 5. A Note is also provided (Note 2 to Required Action E.1) to delineate that Required Action E.1 is only applicable to low pressure ECCS Functions. Required Action E.1 is not applicable to HPCS Functions 3.f and 3.g since the loss of one channel results in a loss of the Function (one-out-of-one logic).

This loss was considered during the development of Reference 4 and considered acceptable for the 7 days allowed by Required Action E.2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "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 variable (Pump Discharge Flow-Low) cannot be automatically initiated due to inoperable channels. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.

If the instrumentation that controls the pump minimum flow valve is inoperable such that the valve will not automatically open, extended pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation such that the valve would not automatically close, a portion of the pump flow could be diverted from the reactor injection path, causing insufficient core cooling.

These consequences can be averted by the operator's manual control of the valve, which would be adequate to maintain ECCS pump protection and required flow. Furthermore, other ECCS pumps would be sufficient to complete the assumed safety function if no additional single failure were to occur. The 7 day Completion Time of Required Action E.2 to restore the inoperable channel to OPERABLE status is reasonable based on the remaining capability of the associated ECCS subsystems, the redundancy available in the ECCS design, and the low probability of a DBA occurring during the allowed out of (continued)

RPV Water Inventory Control Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION B 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation BASES RIVER BEND B 3.3-121a Revision No. 0 BACKGROUND The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF.

If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.3 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.

Technical Specifications are required by 10 CFR 50.36 to include limiting safety system settings (LSSS) for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded." The Analytical Limit is the limit of the process variable at which a safety action is initiated to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur. The actual settings for the automatic isolation channels are the same as those established for the same functions in MODES 1, 2, and 3 in LCO 3.3.5.1, "Emergency Core Cooling System (ECCS)

Instrumentation," or LCO 3.3.6.1, "Primary Containment Isolation instrumentation".

With the unit in MODE 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses.

RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur. Under the definition of DRAIN TIME, some penetration flow paths may be excluded from the DRAIN TIME calculation if they will be isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation.

(continued) new section

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121b Revision No. 4-1 BACKGROUND The purpose of the RPV Water Inventory Control Instrumentation is to support the requirements of LCO 3.5.2, Reactor Pressure Vessel (RPV)

Water Inventory Control, and the definition of DRAIN TIME. There are functions that are required for manual initiation or operation of the ECCS injection/spray subsystem required to be OPERABLE by LCO 3.5.2 and other functions that support automatic isolation of Residual Heat Removal subsystem and Reactor Water Cleanup system penetration flow path(s) on low RPV water level.

The RPV Water Inventory Control Instrumentation supports operation of low pressure core spray (LPCS), low pressure coolant injection (LPCI),

and high pressure core spray (HPCS). The equipment involved with each of these systems is described in the Bases for LCO 3.5.2.

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is not required SAFETY to mitigate any events or accidents evaluated in the safety analysis. RPV ANALYSES, LCO, water inventory control is required in MODES 4 and 5 to protect Safety and APPLICABILITY Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur.

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

seismic event, loss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one ECCS injection/spray subsystem can be manually initiated to maintain adequate reactor vessel water level.

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

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

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

(continued)

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121c Revision No. 0 APPLICABLE The specific Applicable Safety Analysis, LCO, and Applicablity SAFETY discussions are listed below on a Function by Function basis.

ANALYSES, LCO, and APPLICABILITY Low Pressure Coolant Injection and Low Pressure Core Spray 1.a, 2.a, Reactor Vessel Pressure Low - (Injection Permissive)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. While it is assured during Modes 4 and 5 that the reactor steam dome pressure will be below the ECCS maximum design pressure, the Reactor Steam Dome Pressure - Low signals are assumed to be operable and capable of permitting initiation of the ECCS.

The Reactor Steam Dome Pressure - Low signals are initiated from four pressure transmitters that sense the reactor dome pressure. The four pressure transmitters each drive a master and slave trip unit. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each Division.

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

Four channels of Reactor Steam Dome Pressure - Low Function per associated ECCS Division are only required to be OPERABLE in MODES 4 and 5 when ECCS Manual Initiation is required to be OPERABLE, since these channels support the manual initiation Function. In addition, the channels are only required when the associated ECCS subsystem is required to be OPERABLE by LCO 3.5.2.

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

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

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121d Revision No. 0 One flow transmitter per ECCS pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the Residual Heat Removal (RHR) shutdown cooling mode (for RHR A and RHR B).

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

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

1.d, 2.c. Manual Initiation The Manual Initiation push button channels introduce signals into the appropriate ECCS logic to provide manual initiation capability. 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 only the manual initiation function required to be OPERABLE is that associated with the ECCS subsystem required to be OPERABLE by LCO 3.5.2.

There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.

High Pressure Core Spray System 3.a. Reactor Vessel Water Level High - Level 8 The high RPV water level Level 8 signal is used to close the HPCS injection valve to prevent overflow into the main steam lines (MSLs).

Reactor Vessel Water Level - High, Level 8 signals for HPCS are initiated from two level transmitters from the narrow range water level measurement instrumentation. Two channels associated with the HPCS System required to be OPERABLE by LCO 3.5.2 is required to be OPERABLE.

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121e Revision No. 0 The Reactor Vessel Water Level - High, Level 8 Allowable Value is chosen to isolate flow from the HPCS System prior to water overflowing into the MSLs.

Two channels of Reactor Vessel Water Level - High, Level 8 Function are required to be OPERABLE in MODES 4 and 5 when the associated HPCS is required to be OPERABLE by LCO 3.5.2 to ensure to the HPCS is capable of injecting into the Reactor Pressure Vessel when manually initiated.

3.b. Condensate Storage Tank Level - Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source. Normally the suction valves between HPCS and the CST are open and 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.

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 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 fulfill the requirements of LCO 3.5.2 and HPCS is aligned to the CST.

(continued)

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121f Revision No. 2-6 3.c, 3.d. HPCS Pump Discharge Pressure - High (Bypass) and HPCS System Flow Rate - Low (Bypass)

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

One flow transmitter 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 and HPCS Pump Discharge Pressure - High 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 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 associated with one pump is required to be OPERABLE when HPCS is required to be OPERABLE by LCO 3.5.2 in MODES 4 and 5.

3.e. Manual Initiation The Manual Initiation push button channel introduces a signal into the HPCS logic to provide manual initiation capability. There is one push button for the HPCS System. One channel of the Manual Initiation Function is only required to be OPERABLE in MODES 4 and 5 when the associated ECCS subsystem is required to be OPERABLE per LCO 3.5.2.

There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button.

(continued)

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121g Revision No. 0 RHR System Isolation 4.a. Reactor Vessel Water Level - Low, Level 3-The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being automatically isolated by RPV water level isolation instrumentation prior to the RPV water level being equal to the TAF. The Reactor Vessel Water Level - Low, Level 3 Function is only required to be OPERABLE when automatic isolation of the associated RHR penetration flow path is credited in calculating DRAIN TIME.

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

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

This Function isolates the Groups 5 and 14 valves.

Reactor Water Cleanup (RWCU) System Isolation 5.a. Reactor Vessel Water Level - Low Low, Level 2 The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being automatically isolated by RPV water level isolation instrumentation prior to the RPV water level being equal to the TAF. The Reactor Vessel Water Level - Low Low, Level 2 Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.

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

(continued)

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121h Revision No. 0 The Reactor Vessel Water Level - Low Low, Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level - Low Low, Level 2 Allowable Value (LCO 3.3.5.1), since the capability to cool the fuel may be threatened.

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

This Function isolates the Group 8 valves.

ACTIONS A Note has been provided to modify the ACTIONS related to RPV Water Inventory Control instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPV Water Inventory Control instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable RPV Water Inventory Control instrumentation channel.

A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is Function dependent. Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.

B.1 and B.2 RHR System Isolation, Reactor Vessel Water Level - Low Level 3, and Reactor Water Cleanup System, Reactor Vessel Water Level - Low Low, Level 2 functions are applicable when automatic isolation of the associated penetration flow path is credited in calculating Drain Time. If the instrumentation is inoperable, Required Action B.1 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation. Required Action B.2 directs calculation of DRAIN TIME. The calculation cannot credit automatic isolation of the affected penetrations flow paths.

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121i Revision No. 0 C.1 Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual initiation functions. If this permissive is inoperable, manual initiation of ECCS is prohibited. Therefore, the permissive must be placed in the trip condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the permissive in the trip condition, manual initiation may be performed. Prior to placing the permissive in the tripped condition, the operator can take manual control of the pump and the injection valve to inject water into the RPV.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is intended to allow the operator time to evaluate any discovered inoperabilities and to place the channel in trip.

D.1 and D.2 Required Actions D.1 and D.2 are intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function result in a loss of automatic suction swap for the HPCS system from the condensate storage tank to the suppression pool. The HPCS system must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or the HPCS pump suction must be aligned to the suppression pool, since, if aligned, the function is already performed.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes the risk of HPCS being needed without an adequate water source while allowing time for restoration or alignment of HPCS pump suction to the suppression pool.

E.1 and E.2 Required Actions E.1 and E.2 apply when the HPCS Reactor Vessel Water Level - High, Level 8 function is inoperable. If the function is inoperable and the channel is tripped, the HPCS pump discharge valve will not open and HPCS injection is prevented. The HPCS system must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and the function must be restored to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is provided to declare the HPCS System inoperable. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities. The Completion Time is appropriate given the ability to manually start the HPCS and to locally open the discharge valve.

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121j Revision No. 0 F.1 If an LPCI or LPCS Discharge Flow - Low bypass function or HPCS System Discharge Pressure - High or Flow Rate - Low bypass function is inoperable, there is a risk that the associated ECCS pump could overheat when the pump is operating and the associated injection valve is not fully open. In this condition, the operator can take manual control of the pump and the injection valve to ensure the pump does not overheat. If a manual initiation function is inoperable, the ECCS subsystem pumps can be started manually and the valves can be opened manually, but this is not the preferred condition.

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

G.1 With the Required Action and associated Completion Time of Conditions C, D, E, or f not met, the associated ECCS injection/spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately.

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

SR 3.3.5.2.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST.

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

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-121k Revision No. 0 that the instrument has drifted outside its limit.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based upon operating experience that demonstrates channel failure is rare.

The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.

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

The Frequency of 92 days is based upon operating experience that demonstrates channel failure is rare.

SR 3.3.5.2.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.2 overlaps this Surveillance to complete testing of the assumed safety function.

The 24 month Frequency is based on operating experience that has shown that these components usually pass the Surveillance when performed at the 24 month Frequency.

RPV Water Inventory Control Instrumentation B 3.3.5.2 RIVER BEND B 3.3-121l Revision No. 0 REFERENCES

1. Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup,"

November 1984.

2. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.

3. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F), " August 1992.

4. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.

5. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.

RCIC System Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES RIVER BEND B 3.3-122 Revision No. 0 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, "RCIC 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 test line isolation valves close on a RCIC initiation signal to allow full system flow.

The RCIC 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. 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. 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. Two level transmitters are used to detect low water level in the CST. Either switch can cause the suppression pool suction valve to open and the CST suction valve to close. The suppression pool suction valve also automatically opens and the CST suction valve closes if high water level is detected in the suppression pool (continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-123 Revision No. 4-1 BACKGROUND (one-out-of-two logic similar to the CST water level logic). To prevent (continued) losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes.

The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) 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 RCIC 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 ANALYSES, reactor in response to transient events. The RCIC System is not an LCO, and Engineered Safety Feature System and no credit is taken in the safety 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 NRC Policy Statement.

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 3.3.5.2-1. Each Function must have a required number of OPERABLE 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 (continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-124 Revision No. 0 APPLICABLE conservative than the nominal trip setpoint, but within its Allowable Value, SAFETY ANALYSES, is acceptable. Each Allowable Value specified accounts for instrument LCO, and uncertainties appropriate to the Function. These uncertainties are APPLICABILITY described in the setpoint methodology.

(continued)

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

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

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

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-125 Revision No. 2-6 APPLICABLE

2. Reactor Vessel Water Level-High, Level 8 SAFETY ANALYSES, LCO, and High RPV water level indicates that sufficient cooling water inventory APPLICABILITY exists in the reactor vessel such that there is no danger to the fuel.

(continued)

Therefore, the Level 8 signal closes the RCIC steam supply valve to prevent overflow into the main steam lines (MSLs).

Reactor Vessel Water Level-High, Level 8 signals for RCIC are initiated from two level transmitters from the narrow range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

The Reactor Vessel Water Level-High, Level 8 Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.

Two channels of Reactor Vessel Water Level-High, Level 8 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.

3. Condensate Storage Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source. Normally the suction 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.

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-126 Revision No. 0 APPLICABLE

3. Condensate Storage Tank Level-Low (continued)

SAFETY ANALYSES, LCO, and Two level transmitters are used to detect low water level in the CST. The APPLICABILITY Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST.

Two channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source. Refer to LCO 3.5.3 for RCIC Applicability Bases.

4. Suppression Pool Water Level-High 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 Policy Statement. 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.

Suppression pool water level signals are initiated from two level transmitters. The Allowable Value for the Suppression Pool Water Level-High Function is set low enough to ensure that RCIC will be aligned to take suction from the suppression pool before the water level reaches the point at which suppression design loads would be exceeded.

Two channels of Suppression Pool Water Level-High Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source. Refer to LCO 3.5.3 for RCIC Applicability Bases.

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-127 Revision No. 0 APPLICABLE

5. Manual Initiation SAFETY ANALYSES, LCO, and The Manual Initiation push button switch introduces a signal into the RCIC APPLICABILITY 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 Manual Initiation Function is not assumed in any accident or transient analyses in the USAR. However, the Function is retained for the RCIC function as required by the NRC in the plant licensing basis.

There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button.

One channel of Manual Initiation is required to be OPERABLE when RCIC is required to be OPERABLE.

ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel.

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-128 Revision No. 0 ACTIONS A.1 (continued)

Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1 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.

B.1 and B.2 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System. In this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after discovery of loss of RCIC initiation capability.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two inoperable, untripped Reactor Vessel Water Level-Low Low, Level 2 channels in the same trip system. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the (continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-129 Revision No. 0 ACTIONS B.1 and B.2 (continued) tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g.,

as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.

C.1 A risk based analysis was performed and determined that an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1), limiting the allowable out of service time if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level-High, Level 8 Function, whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability. As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable. This Condition also applies to the Manual Initiation Function. Since this Function is not assumed in any accident or transient analysis, a total loss of manual initiation capability (Required Action C.1) for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed. The Required Action does not allow placing a channel in trip since this action would not 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. In this (continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-130 Revision No. 0 ACTIONS D.1, D.2.1, and D.2.2 (continued) situation (loss of automatic suction swap), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Actions D.2.1 and D.2.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action D.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1, which performs the intended function of the channel (shifting the suction source to the suppression pool).

Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water.

If it is not desired to perform Required Actions D.2.1 and D.2.2, Condition E must be entered and its Required Action taken.

With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES (continued)

RIVER BEND B 3.3-131 Revision No. 130 SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC System REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.5.2-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 5; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1, 3, and 4 provided the associated Function maintains initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 1) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.

SR 3.3.5.2.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or 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. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-132 Revision No. 0 SURVEILLANCE SR 3.3.5.2.1 (continued)

REQUIREMENTS The Frequency is based upon operating experience that demonstrates channel failure is rare. 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 3.3.5.2.2 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.

The Frequency of 92 days is based on the reliability analysis of Reference 1.

SR 3.3.5.2.3 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 3.3.5.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be re-adjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.

The Frequency of 92 days is based on the reliability analysis of Reference 1.

SR 3.3.5.2.4 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter with the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

(continued) 3

RCIC System Instrumentation B 3.3.5.2 BASES RIVER BEND B 3.3-133 Revision No. 143 SURVEILLANCE SR 3.3.5.2.4 (continued)

REQUIREMENTS The Frequency is based on the assumption of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.5.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

REFERENCES

1.

NEDE-770-06-2, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.

3

Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES RIVER BEND B 3.3-156 Revision No. 0 APPLICABLE 5.a. Ambient Temperature-High (continued)

SAFETY ANALYSES, LCO, and The Allowable Values are set low enough to detect a leak equivalent to APPLICABILITY 25 gpm.

The RHR Equipment Room Ambient Temperature-High Function is only required to be OPERABLE in MODES 1, 2, and 3. In MODES 4 and 5, insufficient pressure and temperature are available to develop a significant steam leak in this piping and significant water leakage is protected by the Reactor Vessel Water Level-Low, Level 3 Function.

This Function isolates the Group 5 and 14 valves.

5.b. Reactor Vessel Water Level-Low, Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water Level-Low, Level 3 Function associated with RHR Shutdown Cooling System isolation is not directly assumed in any transient or accident analysis, since bounding analyses are performed for large breaks such as MSLBs. The RHR Shutdown Cooling System isolation on Level 3 supports actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown event through the 1E12*F008 and 1E12*F009 valves caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System.

Reactor Vessel Water Level-Low, Level 3 signals are initiated from 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 (two channels per trip system) of the Reactor Vessel Water Level-Low, Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (d) to Table 3.3.6.1-1), only two channels of the Reactor Vessel Water Level-Low, Level 3 Function are required to be OPERABLE in MODES 4 and 5 (both channels must input into the same trip system) to provide an isolation signal to the RHR Shutdown Cooling System suction from the reactor vessel provided the RHR Shutdown Cooling System integrity is (continued)

Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES RIVER BEND B 3.3-157 Revision No. 115 APPLICABLE 5.b. Reactor Vessel Water Level-Low, Level 3 (continued)

SAFETY ANALYSES, LCO, and maintained. System integrity is maintained provided the piping is intact APPLICABILITY and no maintenance is being performed that has the potential for draining the reactor vessel through the system.

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

The Reactor Vessel Water Level-Low, Level 3 Function is required to be OPERABLE in MODES 1, 2, 3 with reactor pressure less than the RHR cut-in permissive pressure, 4, and 5 to prevent this potential flow path from lowering reactor vessel level to the top of the fuel. This instrumentation is required to be OPERABLE in MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut-in permissive pressure to support actions to ensure that offsite dose limits of 10 CFR 50.67 are not exceeded.

This Function isolates the Group 5 and 14 valves.

5.c. Reactor Vessel Water Level-Low Low Low, Level 1 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of the shutdown cooling portion of the RHR System occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level-Low Low Low, Level 1 Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. The Reactor Vessel Water Level-Low Low Low, Level 1 Function associated with isolation is implicitly assumed in the USAR analysis since these leakage paths are assumed to be isolated for a DBA.

Reactor vessel water level signals are initiated from 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 Reactor Vessel Water Level-Low Low Low, Level 1 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

(continued)

Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES RIVER BEND B 3.3-164 Revision No. 0 ACTIONS I.1 and I.2 (continued) this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated SLC subsystem inoperable or isolating the RWCU System.

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

J.1, J.2, J.3.1 and J.3.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the RHR Shutdown Cooling System suction from the reactor vessel should be isolated (i.e., closing 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 or to provide alternate decay heat removal capability and subsequently isolate the RHR Shutdown Cooling System or to provide means for control of potential radioactive releases. This is accomplished by ensuring primary containment is OPERABLE. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillances may need to be performed to restore the component to OPERABLE status. In addition, at least one door in each primary containment air lock must be closed. The closed air lock door completes the boundary for control of potential radioactive releases. With the appropriate administrative controls however, the closed air lock door can be opened intermittently for entry and exit. This allowance is acceptable due to the need for containment access and due to the slow progression of events which may result from a reactor vessel draindown event. Reactor vessel draindown events would not be expected to result in the immediate release of appreciable fission products to the containment atmosphere.

Actions must continue until all requirements of this Condition are satisfied.

(continued)

CRFA System Instrumentation B 3.3.7.1 BASES RIVER BEND B 3.3-200 Revision No. 0 APPLICABLE

1. Reactor Vessel Water Level-Low Low, Level 2 SAFETY ANALYSES, LCO, and Low reactor pressure vessel (RPV) water level indicates that the 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 CRFA System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

Reactor Vessel Water Level - Low Low, Level 2 signals are initiated from 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 Reactor Vessel Water Level - Low Low, Level 2 Function are available (two channels per trip system) and are required to be OPERABLE to ensure that no single instrument failure can preclude CRFA System initiation.

The Allowable Value for the Reactor Vessel Water Level - Low Low, Level 2 is chosen to be the same as the Secondary Containment Isolation Reactor Vessel Water Level - Low Low, Level 2 Allowable Value (LCO 3.3.6.2).

The Reactor Vessel Water Level - Low Low, Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 to ensure that the control room personnel are protected. In MODES 4 and 5, the probability of a vessel draindown event or of a LOCA, is minimal. Therefore this Function is not required. In addition, the Control Room Ventilation Radiation Monitor Function provides adequate protection.

2. Drywell Pressure - High 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 CRFA System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

(continued)

CRFA System Instrumentation B 3.3.7.1 BASES RIVER BEND B 3.3-201 Revision No. 115 APPLICABLE

2. Drywell Pressure - High (continued)

SAFETY ANALYSES, LCO, and Drywell Pressure-High signals are initiated from four pressure APPLICABILITY transmitters that sense drywell pressure. Four channels of Drywell Pressure-High Function are available (two channels per trip system) and are required to be OPERABLE to ensure that no single instrument failure can preclude CRFA System initiation.

The Drywell Pressure-High Allowable Value was chosen to be the same as the Secondary Containment Isolation Drywell Pressure-High Allowable Value (LCO 3.3.6.2).

The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 to ensure that control room personnel are protected during a LOCA. In MODES 4 and 5, the Drywell Pressure-High Function is not required since there is insufficient energy in the reactor to pressurize the drywell to the Drywell Pressure-High setpoint.

3. Control Room Local Intake Ventilation Radiation Monitors The Control Room Local Intake Ventilation Radiation Monitors measure radiation levels exterior to the inlet ducting of the MCR. A high radiation level may pose a threat to MCR personnel; thus, a detector indicating this condition automatically signals initiation of the CRFA System.

The Control Room Local Intake Ventilation Radiation Monitors Function consists of two independent monitors. Two channels of Control Room Local Intake Ventilation Radiation Monitors are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CRFA System initiation. The Allowable Value was selected to ensure protection of the control room personnel.

The Control Room Local Intake Ventilation Radiation Monitors Function is required to be OPERABLE in MODES 1, 2, and 3, and during operations with a potential for draining the reactor vessel (OPDRVs) and movement of recently irradiated fuel in the primary containment or fuel building to ensure that control room personnel are protected during a LOCA, fuel handling event, or a vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., OPDRVs), the probability of a LOCA or fuel damage is low; thus, the Function is not required.

(continued) or a

ECCSOperating B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCSOperating BASES RIVER BEND B 3.5-1 Revision No. 0 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 low 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 starts; 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 sparger above the core. If the break is small, HPCS will maintain coolant inventory, as well as vessel level, while the RCS is still pressurized. If 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 timed sequence would be allowed to 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. Alternately, procedures may direct this automatic function be inhibited until subsequently required. If the break is large, RCS pressure initially drops rapidly, and the LPCI and LPCS systems cool the core.

(continued)

, RPV WATER INVENTORY CONTROL,

ECCSOperating B 3.5.1 BASES (continued)

RIVER BEND B 3.5-5 Revision No. 163 LCO Each ECCS injection/spray subsystem and seven 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 further subdivided into the following groups. Management of gas voids is important to ECCS injection/spray subsystem OPERABILITY.

a)

The low 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 limiting design basis LOCA concurrent with the worst case single failure, the limits 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 otherwise 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 MODES 1, 2, and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, the ADS function is not required when pressure is 100 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 5 are specified in LCO 3.5.2, "ECCSShutdown."

(continued)

R RPV Water Inventory Control

ECCSShutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 ECCSShutdown BASES RIVER BEND B 3.5-15 Revision No. 163 BACKGROUND A description of the High Pressure Core Spray (HPCS) System, Low Pressure Core Spray (LPCS) System, and low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCSOperating."

APPLICABLE ECCS performance is evaluated for the entire spectrum of break sizes for SAFETY ANALYSES a postulated loss of coolant accident (LOCA). The long term cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one ECCS injection/spray subsystem is required, post LOCA, to maintain the peak cladding temperature below the allowable limit. It is reasonable to assume, based on engineering judgement, that while in MODES 4 and 5, one ECCS injection/spray subsystem can maintain adequate reactor vessel water level. To provide redundancy, a minimum of two ECCS subsystems are required to be OPERABLE in MODES 4 and 5.

The ECCS satisfy Criterion 3 of the NRC Policy Statement.

LCO Two ECCS injection/spray subsystems are required to be OPERABLE.

The ECCS injection/spray subsystems are defined as the three LPCI subsystems, the LPCS System, and the HPCS System. The LPCS System and each LPCI subsystem consist of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV.

The HPCS System consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV. Management of gas voids is important to ECCS injection/spray subsystem OPERABILITY.

One LPCI subsystem (A or B) may be aligned for decay heat removal in MODE 4 or 5 and considered OPERABLE for the ECCS function, if it can be manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Because of low pressure and low temperature conditions in MODES 4 (continued)

RPV Water Inventory Control Reactor Pressure Vessel (RPV) Water Inventory Control

, RPV WATER INVENTORY CONTROL, Insert "F"

ECCSShutdown B 3.5.2 BASES RIVER BEND B 3.5-16 Revision No. 0 LCO and 5, sufficient time will be available to manually align and initiate LPCI (continued) subsystem operation to provide core cooling prior to postulated fuel uncovery.

APPLICABILITY OPERABILITY of the ECCS injection/spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1. ECCS subsystems are not required to be OPERABLE during MODE 5 with the upper containment fuel pool to reactor cavity gate opened, and the water level maintained at 23 ft above the RPV flange. This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown.

The Automatic Depressurization System is not required to be OPERABLE during MODES 4 and 5 because the RPV pressure is < 150 psig, and the LPCS, HPCS, and LPCI subsystems can provide core cooling without any depressurization of the primary system.

ACTIONS A.1 and B.1 If any one required ECCS injection/spray subsystem is inoperable, the required inoperable ECCS injection/spray subsystem must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In this Condition, the remaining OPERABLE subsystem can provide sufficient RPV flooding capability to recover from an inadvertent vessel draindown. However, overall system reliability is reduced because a single failure in the remaining OPERABLE subsystem concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for restoring the required ECCS injection/spray subsystem to OPERABLE status is based on engineering judgment that considered the availability of one subsystem and the low probability of a vessel draindown event.

(continued)

RPV Water Inventory Control

ECCSShutdown B 3.5.2 BASES RIVER BEND B 3.5-17 Revision No. 0 ACTIONS A.1 and B.1 (continued)

With the inoperable subsystem not restored to OPERABLE status within the required Completion Time, action must be initiated immediately to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

C.1, C.2, D.1, D.2, and D.3 If both of the required ECCS injection/spray subsystems are inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must be initiated immediately to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. One ECCS injection/spray subsystem must also be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If at least one ECCS injection/spray subsystem is not restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes initiating immediate action to restore primary containment to OPERABLE status and isolate the penetrations that are assumed to be isolated to mitigate radioactivity releases. This may be performed by an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. Verification does not require performing the Surveillances needed to demonstrate OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillances may need to be performed to restore the component to OPERABLE status. In addition, at least one door in each primary containment air lock must be closed. The closed air lock door completes the boundary for control of potential radioactive releases. With the appropriate administrative controls, however, the closed door can be opened intermittently for entry and exit. This allowance is acceptable due to the need for containment access and due to the slow progression of events which may (continued)

RPV Water Inventory Control

ECCSShutdown B 3.5.2 BASES RIVER BEND B 3.5-18 Revision No. 0 ACTIONS C.1, C.2, D.1, D.2, and D.3 (continued) result from the identified conditions. The lack of available ECCS during shutdown conditions would not be expected to result in the immediate release of appreciable fission products to the containment atmosphere.

Actions must continue until all requirements of this Condition are satisfied.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time to restore at least one ECCS injection/spray subsystem to OPERABLE status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 REQUIREMENTS The minimum water level of 13 ft 3 inches required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the ECCS pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, all ECCS injection/spray subsystems are inoperable unless they are aligned to an OPERABLE CST.

When the suppression pool level is < 13 ft 3 inches, the HPCS System is considered OPERABLE only if it can take suction from the CST and the CST water level is sufficient to provide the required NPSH for the HPCS pump. Therefore, a verification that either the suppression pool water level is 13 ft 3 inches or the HPCS System is aligned to take suction from the CST and the CST contains 125,000 gallons of water, equivalent to 11 ft 1 inch, ensures that the HPCS System can supply makeup water to the RPV.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of these SRs was developed considering operating experience related to suppression pool and CST water level variations during the applicable MODES. Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.

(continued)

RPV Water Inventory Control

ECCSShutdown B 3.5.2 BASES RIVER BEND B 3.5-19 Revision No. 163 SURVEILLANCE SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6 REQUIREMENTS (continued)

The Bases provided for SR 3.5.1.1, SR 3.5.1.4, and SR 3.5.1.5 are applicable to SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6, respectively.

SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.

In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor.

Therefore, RHR valves that are required for LPCI subsystem operation may be aligned for decay heat removal. This SR is modified by a Note that allows one LPCI subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned (remote or local) to allow injection into the RPV and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent vessel draindown should occur.

The Surveillance is modified by a Note which exempts system vent flow paths opened under administrative control. The administrative control should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication with the operators in the control room. This individual will have a method to rapidly close the system vent flow path if directed.

REFERENCES

1.

USAR, Section 6.3.3.4.

RPV Water Inventory Control

(InsertFreplacesBases3.5.2initsentirety)

BACKGROUND TheRPVcontainspenetrationsbelowthetopoftheactivefuel(TAF)thathave thepotentialtodrainthereactorcoolantinventorytobelowtheTAF.Ifthe waterlevelshoulddropbelowtheTAF,theabilitytoremovedecayheatis reduced,whichcouldleadtoelevatedcladdingtemperaturesandclad perforation.SafetyLimit2.1.1.3requirestheRPVwaterleveltobeabovethe topoftheactiveirradiatedfuelatalltimestopreventsuchelevatedcladding temperatures.

APPLICABLE

SAFETY ANALYSIS

WiththeunitinMODE4or5,RPVwaterinventorycontrolisnotrequiredto mitigateanyeventsoraccidentsevaluatedinthesafetysizesforapostulated lossofcoolantaccident(LOCA).Thelongtermanalyses.RPVwaterinventory controlisrequiredinMODES4and5toprotectSafetyLimit2.1.1.3andthe fuelcladdingbarriertopreventthereleaseofradioactivematerialtothe environmentshouldanunexpecteddrainingeventoccur.

AdoubleendedguillotinebreakoftheReactorCoolantSystem(RCS)isnot postulatedinMODES4and5duetothereducedRCSpressure,reducedpiping stresses,andductilepipingsystems.Instead,aneventisconsideredinwhich singleoperatorerrororinitiatingeventallowsdrainingoftheRPVwater inventorythroughasinglepenetrationflowpathwiththehighestflowrate,or thesumofthedrainratesthroughmultiplepenetrationflowpathssusceptible toacommonmodefailure(e.g.,seismicevent,lossofnormalpower,single humanerror).Itisassumed,basedonengineeringjudgment,thatwhilein MODES4and5,onelowpressureECCSinjection/spraysubsystemcan maintainadequatereactorvesselwaterlevel.

AsdiscussedinReferences1,2,3,4,and5,operatingexperiencehasshown RPVwaterinventorytobesignificanttopublichealthandsafety.Therefore, RPVWaterInventoryControlsatisfiesCriterion4of10CFR50.36(c)(2)(ii).

LCO TheRPVwaterlevelmustbecontrolledinMODES4and5toensurethatifan unexpecteddrainingeventshouldoccur,thereactorcoolantwaterlevel remainsabovethetopoftheactiveirradiatedfuelasrequiredbySafetyLimit 2.1.1.3.

TheLimitingConditionforOperation(LCO)requirestheDRAINTIMEofRPV waterinventorytotheTAFtobe36hours.ADRAINTIMEof36hoursis consideredreasonabletoidentifyandinitiateactiontomitigateunexpected drainingofreactorcoolant.AneventthatcouldcauselossofRPVwater inventoryandresultintheRPVwaterlevelreachingtheTAFingreaterthan36 hoursdoesnotrepresentasignificantchallengetoSafetyLimit2.1.1.3andcan bemanagedaspartofnormalplantoperation.

OneECCSinjection/spraysubsystemisrequiredtobeOPERABLEandcapable ofbeingmanuallystartedtoprovidedefenseindepthshouldanunexpected drainingeventoccur.AnECCSinjection/spraysubsystemisdefinedaseither

oneofthethreeLowPressureCoolantInjection(LPCI)subsystems,theLow PressureCoreSpray(LPCS)System,ortheHighPressureCoreSpray(HPCS)

System.TheLPCIsubsystemandtheeachLPCSSystemconsistofonemotor drivenpump,piping,andvalvestotransferwaterfromthesuppressionpoolto theRPV.TheHPCSSystemconsistsofonemotordrivenpump,piping,and valvestotransferwaterfromthesuppressionpoolorcondensatestoragetank (CST)totheRPV.

TheLCOismodifiedbyaNotewhichallowsarequiredLPCIsubsystem(AorB) tomaybeconsideredOPERABLEduringalignmentandoperationfordecay heatremoval,ifcapableofbeingmanuallyrealigned(remoteorlocal)tothe LPCImodeandisnototherwiseinoperable.Alignmentandoperationfordecay heatremovalincludeswhentherequiredRHRpumpisnotoperatingorwhen thesystemisrealignedfromortotheRHRshutdowncoolingmode.This allowanceisnecessarysincetheRHRSystemmayberequiredtooperatein theshutdowncoolingmodetoremovedecayheatandsensibleheatfromthe reactor.BecauseoftherestrictionsonDRAINTIME,sufficienttimewillbe availablefollowinganunexpecteddrainingeventtomanuallyalignandinitiate LPCIsubsystemoperationtomaintainRPVwaterinventorypriortotheRPV waterlevelreachingtheTAF

APPLICABILITY RPVwaterinventorycontrolisrequiredinMODES4and5.Requirementson waterinventorycontrolinotherMODESarecontainedinLCOsinSection3.3, Instrumentation,andotherLCOsinSection3.5,ECCS,RCIC,andRPVWater InventoryControl.RPVwaterinventorycontrolisrequiredtoprotectSafety Limit2.1.1.3whichisapplicablewheneverirradiatedfuelisinthereactor vessel.

ACTIONS A.1andB.1

IftherequiredECCSinjection/spraysubsystemisinoperable,itmustbe restoredtoOPERABLEstatuswithin4hours.InthisCondition,theLCO controlsonDRAINTIMEminimizethepossibilitythatanunexpecteddraining eventcouldnecessitatetheuseoftheECCSinjection/spraysubsystem, howeverthedefenseindepthprovidedbytheECCSinjection/spraysubsystem islost.The4hourCompletionTimeforrestoringtherequiredECCS injection/spraysubsystemtoOPERABLEstatusisbasedonengineering judgmentthatconsiderstheLCOcontrolsonDRAINTIMEandthelow probabilityofanunexpecteddrainingeventthatwouldresultinlossofRPV waterinventory.

IftheinoperableECCSinjection/spraysubsystemisnotrestoredtoOPERABLE statuswithintherequiredCompletionTime,actionmustbeinitiated immediatelytoestablishamethodofwaterinjectioncapableofoperating withoutoffsiteelectricalpower.Themethodofwaterinjectionincludesthe necessaryinstrumentationandcontrols,watersources,andpumpsandvalves neededtoaddwatertotheRPVorrefuelingcavityshouldanunexpected drainingeventoccur.Themethodofwaterinjectionmaybemanually

initiatedandmayconsistofoneormoresystemsorsubsystems,andmustbe abletoaccesswaterinventorycapableofmaintainingtheRPVwaterlevel abovetheTAFfor36hours.Ifrecirculationofinjectedwaterwouldoccur,it maybecreditedindeterminingthenecessarywatervolume

C.1andC.2

WiththeDRAINTIMElessthan36hoursbutgreaterthanorequalto8hours, compensatorymeasuresshouldbetakentoensuretheabilitytoimplement mitigatingactionsshouldanunexpecteddrainingeventoccur.Shoulda drainingeventlowerthereactorcoolantleveltobelowtheTAF,thereis potentialfordamagetothereactorfuelcladdingandreleaseofradioactive material.Additionalactionsaretakentoensurethatradioactivematerialwill becontainedanddilutedpriortobeingreleasedtotheenvironment.

D.1,D.2,andD.3

WiththeDRAINTIMElessthan8hours,mitigatingactionsareimplementedin caseanunexpecteddrainingeventshouldoccur.NotethatiftheDRAINTIME islessthan1hour,RequiredActionE.1isalsoapplicable.

RequiredActionD.1requiresimmediateactiontoestablishanadditional methodofwaterinjectionaugmentingtheECCSinjection/spraysubsystem requiredbytheLCO.Theadditionalmethodofwaterinjectionincludesthe necessaryinstrumentationandcontrols,watersources,andpumpsandvalves neededtoaddwatertotheRPVorrefuelingcavityshouldanunexpected drainingeventoccur.TheNotetoRequiredActionD.1statesthateitherthe ECCSinjection/spraysubsystemortheadditionalmethodofwaterinjection mustbecapableofoperatingwithoutoffsiteelectricalpower.Theadditional methodofwaterinjectionmaybemanuallyinitiatedandmayconsistofoneor moresystemsorsubsystems.Theadditionalmethodofwaterinjectionmust beabletoaccesswaterinventorycapableofbeinginjectedtomaintainthe RPVwaterlevelabovetheTAFfor36hours.Theadditionalmethodofwater injectionandtheECCSinjection/spraysubsystemmayshareallorpartofthe samewatersources.Ifrecirculationofinjectedwaterwouldoccur,itmaybe creditedindeterminingtherequiredwatervolume.

ShouldadrainingeventlowerthereactorcoolantleveltobelowtheTAF,there ispotentialfordamagetothereactorfuelcladdingandreleaseofradioactive material.Additionalactionsaretakentoensurethatradioactivematerialwill becontainedanddilutedpriortobeingreleasedtotheenvironment.

E.1

IftheRequiredActionsandassociatedCompletiontimesofConditionsCorD arenotmetoriftheDRAINTIMEislessthan1hour,actionsmustbeinitiated immediatelytorestoretheDRAINTIMEto36hours.Inthiscondition,there maybeinsufficienttimetorespondtoanunexpecteddrainingeventto

preventtheRPVwaterinventoryfromreachingtheTAF.NotethatRequired ActionsD.1,D.2,D.3,andD.4arealsoapplicablewhenDRAINTIMEislessthan 1hour.

SURVEILLANCE REQUIREMENTS SR3.5.2.1

ThisSurveillanceverifiesthattheDRAINTIMEofRPVwaterinventorytothe TAFis36hours.Theperiodof36hoursisconsideredreasonabletoidentify andinitiateactiontomitigatedrainingofreactorcoolant.LossofRPVwater inventorythatwouldresultintheRPVwaterlevelreachingtheTAFingreater than36hoursdoesnotrepresentasignificantchallengetoSafetyLimit2.1.1.3 andcanbemanagedaspartofnormalplantoperation.

ThedefinitionofDRAINTIMEstatesthatrealisticcrosssectionalareasand drainratesareusedinthecalculation.Arealisticdrainratemaybe determinedusingasingle,stepwise,orintegratedcalculationconsideringthe changingRPVwaterlevelduringadrainingevent.ForaControlRodRPV penetrationflowpathwiththeControlRodDriveMechanismremovedand notreplacedwithablankflange,therealisticcrosssectionalareaisbasedon thecontrolrodbladeseatedinthecontrolrodguidetube.Ifthecontrolrod bladewillberaisedfromthepenetrationtoadjustorverifyseatingofthe blade,theexposedcrosssectionalareaoftheRPVpenetrationflowpathis used.

ThedefinitionofDRAINTIMEexcludesfromthecalculationthosepenetration flowpathsconnectedtoanintactclosedsystem,orisolatedbymanualor automaticvalvesthatarelocked,sealed,orotherwisesecuredintheclosed position,blankflanges,orotherdevicesthatpreventflowofreactorcoolant throughthepenetrationflowpaths.Ablankflangeorotherbolteddevice mustbeconnectedwithasufficientnumberofboltstopreventdraininginthe eventofanOperatingBasisEarthquake.Normalorexpectedleakagefrom closedsystemsorpastisolationdevicesispermitted.Determinationthata systemisintactandclosedorisolatedmustconsiderthestatusofbranchlines andongoingplantmaintenanceandtestingactivities.

TheResidualHeatRemoval(RHR)ShutdownCoolingSystemisonlyconsidered anintactclosedsystemwhenmisalignmentissues(Reference6)havebeen precludedbyfunctionalvalveinterlocksorbyisolationdevices,suchthat redirectionofRPVwateroutofanRHRsubsystemisprecluded.Further,RHR ShutdownCoolingSystemisonlyconsideredanintactclosedsystemifits controlshavenotbeentransferredtoRemoteShutdown,whichdisablesthe interlocksandisolationsignals.

TheexclusionofpenetrationflowpathsfromthedeterminationofDRAIN TIMEmustconsiderthepotentialeffectsofasingleoperatorerrororinitiating eventonitemssupportingmaintenanceandtesting(rigging,scaffolding, temporaryshielding,pipingplugs,snubberremoval,freezeseals,etc.).If

failureofsuchitemscouldresultandwouldcauseadrainingeventfroma closedsystemorbetweentheRPVandtheisolationdevice,thepenetration flowpathmaynotbeexcludedfromtheDRAINTIMEcalculation.

SurveillanceRequirement3.0.1requiresSRstobemetbetweenperformances.

Therefore,anychangesinplantconditionsthatwouldchangetheDRAINTIME requiresthatanewDRAINTIMEbedetermined.

TheFrequencyof12hoursissufficientinviewofindicationsofRPVwaterlevel availabletotheoperator.

SR3.5.2.2andSR3.5.2.3

Theminimumwaterlevelof13feet3inchesrequiredforthesuppressionpool isperiodicallyverifiedtoensurethatthesuppressionpoolwillprovide adequatenetpositivesuctionhead(NPSH)fortheECCSpumps,recirculation volume,andvortexprevention.Withthesuppressionpoolwaterlevelless thantherequiredlimit,therequiredECCSinjection/spraysubsystemis inoperableunlessalignedtoanOPERABLECST.

Whenthesuppressionpoollevelis<13feet3inches,theHPCSSystemis consideredOPERABLEonlyifitcantakesuctionfromtheCSTandtheCST waterlevelissufficienttoprovidetherequiredNPSHfortheHPCSpump.

Therefore,averificationthateitherthesuppressionpoolwaterlevelis13 feet3inchesortheHPCSSystemisalignedtotakesuctionfromtheCSTand theCSTlevelis11feet1inchensuresthattheHPCSSystemcansupply makeupwatertotheRPV.

The12hourFrequencyoftheseSRswasdevelopedconsideringoperating experiencerelatedtosuppressionpoolandCSTwaterlevelvariationsand instrumentdriftduringtheapplicableMODES.Furthermore,the12hour Frequencyisconsideredadequateinviewofotherindicationsavailableinthe controlroom,includingalarms,toalerttheoperatortoanabnormal suppressionpoolorCSTwaterlevelcondition.

SR3.5.2.4

Theflowpathpipinghasthepotentialtodevelopvoidsandpocketsof entrainedair.MaintainingthepumpdischargelinesoftherequiredECCS injection/spraysubsystemsfullofwaterensuresthattheECCSsubsystemwill performproperly.ThismayalsopreventawaterhammerfollowinganECCS initiationsignal.Oneacceptablemethodofensuringthatthelinesarefullisto ventatthehighpoints.The31dayFrequencyisbasedonthegradualnature ofvoidbuildupintheECCSpiping,theproceduralcontrolsgoverningsystem operation,andoperatingexperience.

SR3.5.2.5

Verifyingthecorrectalignmentformanual,poweroperated,andautomatic valvesintherequiredECCSsubsystemflowpathsprovidesassurancethatthe properflowpathswillbeavailableforECCSoperation.ThisSRdoesnotapply tovalvesthatarelocked,sealed,orotherwisesecuredinpositionsincethese valveswereverifiedtobeinthecorrectpositionpriortolocking,sealing,or securing.Avalvethatreceivesaninitiationsignalisallowedtobeinanon accidentpositionprovidedthevalvewillautomaticallyrepositionintheproper stroketime.ThisSRdoesnotrequireanytestingorvalvemanipulation; rather,itinvolvesverificationthatthosevalvescapableofpotentiallybeing mispositionedareinthecorrectposition.ThisSRdoesnotapplytovalvesthat cannotbeinadvertentlymisaligned,suchascheckvalves.The31day Frequencyisappropriatebecausethevalvesareoperatedunderprocedural controlandtheprobabilityoftheirbeingmispositionedduringthistimeperiod islow.

Note1isincludedtoclarifytheconditionofOPERABILITYofaLPCIsystem whenoperatinginadecayheatremovalmode.Note2providesfor administrativelycontrollingthesystemventflowpathswhilemaintaining compliancewiththisSR.

SR3.5.2.6

VerifyingthattherequiredECCSinjection/spraysubsystemcanbemanually startedandoperateforatleast10minutesdemonstratesthatthesubsystem isavailabletomitigateadrainingevent.Theminimumoperatingtimeof10 minuteswasbasedonengineeringjudgment.Theperformancefrequencyof 92daysisconsistentwithsimilaratpowertestingrequiredbySR3.5.1.7.

SR3.5.2.7

Verifyingthateachvalvecreditedforautomaticallyisolatingapenetration flowpathactuatestotheisolationpositiononanactualorsimulatedRPV waterlevelisolationsignalisrequiredtopreventRPVwaterinventoryfrom droppingbelowtheTAFshouldanunexpecteddrainingeventoccur.The24 monthFrequencyisbasedontheneedtoperformthisSurveillanceunderthe conditionsthatapplyduringaplantoutageandthepotentialforanunplanned transientiftheSurveillancewereperformedwiththereactoratpower.

Operatingexperiencehasshownthesecomponentsusuallypassthe Surveillancewhenperformedatthe24monthFrequency.Therefore,the Frequencywasconcludedtobeacceptablefromareliabilitystandpoint.

SR3.5.2.8

TherequiredECCSsubsystemisrequiredtohaveamanualstartcapability.

ThisSurveillanceverifiesthatamanualinitiationsignalwillcausetherequired LPCIsubsystemorLPCSSystemtostartandoperateasdesigned,including

pumpstartupandactuationofallautomaticvalvestotheirrequiredpositions.

TheHPCSsystemisverifiedtostartmanuallyfromastandbyconfiguration, andincludestheabilitytooverridetheRPVLevel8injectionvalveisolation.

The24monthFrequencyisbasedontheneedtoperformtheSurveillance undertheconditionsthatapplyduringaplantoutageandthepotentialforan unplannedtransientiftheSurveillancewereperformedwiththereactorat power.

OperatingexperiencehasshownthatthesecomponentsusuallypasstheSR whenperformedatthe24monthFrequency,whichisbasedontherefueling cycle.Therefore,theFrequencywasconcludedtobeacceptablefroma reliabilitystandpoint.

ThisSRismodifiedbyaNotethatexcludesvesselinjection/sprayduringthe Surveillance.Sinceallactivecomponentsaretestableandfullflowcanbe demonstratedbyrecirculationthroughthetestline,coolantinjectionintothe RPVisnotrequiredduringtheSurveillance.

REFERENCES

1. InformationNotice8481"InadvertentReductioninPrimaryCoolant InventoryinBoilingWaterReactorsDuringShutdownandStartup,"

November1984.

2. InformationNotice8674,"ReductionofReactorCoolantInventory BecauseofMisalignmentofRHRValves,"August1986.

3. GenericLetter9204,"ResolutionoftheIssuesRelatedtoReactorVessel WaterLevelInstrumentationinBWRsPursuantto10CFR50.54(F),"

August1992.

4. NRCBulletin9303,"ResolutionofIssuesRelatedtoReactorVesselWater LevelInstrumentationinBWRs,"May1993.

5. InformationNotice9452,"InadvertentContainmentSprayandReactor VesselDraindownatMillstone1,"July1994.

6. GeneralElectricServiceInformationLetterNo.388,"RHRValve MisalignmentDuringShutdownCoolingOperationforBWR3/4/5/6,"

February1983.

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.3 RCIC System BASES RIVER BEND B 3.5-20 Revision No. 6-14 BACKGROUND The RCIC System is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level. Under these conditions, the High Pressure Core Spray (HPCS) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference 1 are satisfied.

The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the Feedwater System piping. 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, if the CST water supply is low, or the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from main steam line A, upstream of the inboard main steam line isolation valve.

The RCIC System is designed to provide core cooling for a wide range of reactor pressures, 150 psig to 1231 psig. Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

(continued)

, RPV WATER INVENTORY CONTROL,

RCIC System B 3.5.3 BASES RIVER BEND B 3.5-21 Revision No. 163 BACKGROUND The RCIC pump is provided with a minimum flow, bypass line which (continued) discharges to the suppression pool. The valve in this line automatically opens to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge 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 ANALYSES providing makeup coolant to the reactor. The RCIC system is not an Engineered Safety Feature system, and the safety analysis does not consider RCIC to be a system needed to mitigate the consequences of a control rod drop accident. 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 Policy Statement.

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. Management of gas voids is important to RCIC system OPERABILITY.

APPLICABILITY The RCIC System is required to be OPERABLE in MODE 1, and MODES 2 and 3 with reactor steam dome pressure > 150 psig since RCIC is the primary nonECCS water source for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure 150 psig, and in MODES 4 and 5, RCIC is not required to be OPERABLE since the ECCS injection/spray subsystems can provide sufficient flow to the vessel.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable RCIC system. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable RCIC system and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

(continued) the ECCS injection/spray subsystems can provide sufficient flow to the vessel. In RPV water inventory control is required by LCO 3.5.2, "RPV Water Inventory Control."

Primary Containment Air Locks B 3.6.1.2 BASES RIVER BEND B 3.6-7 Revision No. 110 LCO The primary containment air locks are required to be OPERABLE. For (continued) 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 allows only one air lock door to be open at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE.

Closure of a single 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 In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, 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 during operations with a potential for draining the reactor vessel (OPDRVs) or during fuel movement of recently irradiated fuel assemblies in the primary containment. Due to radioactive decay, primary containment air locks are only required during fuel handling in the primary containment involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component. If the outer door is inoperable, then it may be easily accessed for most repairs. It is preferred that the air lock be accessed from inside primary containment by entering through the other OPERABLE air lock. However, if this is not practicable, or if repairs on either door must be performed from the barrel side of the door, then it is permissible to enter the air lock through the OPERABLE door, which means there is a short time during which the primary containment boundary is not intact (during access through the OPERABLE door). The ability to open the OPERABLE door, even if it means the primary containment boundary is (continued)

Primary Containment Air Locks B 3.6.1.2 BASES RIVER BEND B 3.6-11 Revision No. 2-3 ACTIONS C.1, C.2, and C.3 (continued) failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed) primary containment remains OPERABLE, yet only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (according to LCO 3.6.1.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits. Required Action C.2 requires that one door in the affected primary containment air locks must be verified closed. This Required Action must be completed within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time.

This specified time period is consistent with the ACTIONS of LCO 3.6.1.1, which require that primary containment be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Additionally, the air lock must be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable for restoring an inoperable air lock to OPERABLE status considering that at least one door is maintained closed in each affected air lock.

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

E.1, E.2, and E.3 If the inoperable primary containment airlock cannot be restored to OPERABLE status within the associated Completion Time during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the primary containment, action is required to immediately suspend activities that represent a potential for releasing radioactive material, (continued)

Primary Containment Air Locks B 3.6.1.2 BASES RIVER BEND B 3.6-12 Revision No. 128 ACTIONS E.1, E.2, and E.3 (continued) 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. Also, if applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended.

SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of the Primary Containment Leakage Rate Testing Program (Ref. 5). This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). Following the removal of the fuel building as a secondary containment boundary in accordance with License Amendment 113, the leakage from primary containment air lock 1JRB*DRA2 represents secondary bypass leakage limit. The secondary containment leakage limit of 580,000 cc/hr accounted for the potential leakage paths and was assumed in the Alternate Source Term (AST)

LOCA Analysis (Amendment 132). This provides assurance in MODES 1, 2, and 3 that the assumptions in the radiological evaluations are met.

The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (e.g., leakage through the air lock door with the highest leakage) unless the penetration is isolated by use of (for this Specification) one closed and locked air lock door. The leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation devices (e.g., air lock door). If both air lock doors are closed, the actual leakage rate is the lesser leakage rate of the two barriers (doors). This method of quantifying maximum pathway leakage is only to be used for this SR (i.e., Appendix J, Option B, maximum pathway leakage limits used to evaluate Type A, B and C limits are to be quantified in accordance with Appendix J, Option B).

During the operational conditions of moving irradiated fuel assemblies in the primary containment, CORE ALTERATIONS, or OPDRVS, (continued) or

PCIVs B 3.6.1.3 BASES RIVER BEND B 3.6-17 Revision No. 128 LCO are listed with their associated stroke times, if applicable, in the USAR (continued)

(Ref. 3). Purge valves with resilient seals, secondary containment bypass valves, MSIVs, and hydrostatically tested valves must meet additional leakage rate requirements. Other PCIV leakage rates are addressed by LCO 3.6.1.1, "Primary ContainmentOperating," as Type B or C testing.

This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory, and establish the primary containment boundary during accidents.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, most PCIVs are not required to be OPERABLE. Certain valves are required to be OPERABLE, however, to prevent a potential flow path (the RHR Shutdown Cooling System suction from the reactor vessel) from lowering reactor vessel level to the top of the fuel. These valves are those whose associated instrumentation is required to be OPERABLE according to LCO 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation," Function 5.b. (This does not include the valves that isolate the associated instrumentation.)

ACTIONS The ACTIONS are modified by a Note allowing penetration flow path(s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.

A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable PCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions.

(continued) when the

PCIVs B 3.6.1.3 BASES RIVER BEND B 3.6-21 Revision No. 129 ACTIONS D.1, D.2, and D.3 (continued) verification that those isolation devices outside primary containment and potentially capable of being mispositioned are in the correct position. For the isolation devices inside primary containment, the time period specified as "prior to entering MODE 2 or 3, from MODE 4 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of administrative controls that will ensure that isolation device misalignment is an unlikely possibility.

For a primary containment purge valve with a resilient seal that is isolated in accordance with Required Action D.1, SR 3.6.1.3.5 must be performed at least once every 92 days. This provides assurance that degradation of the resilient seal is detected and confirms that the leakage rate of the primary containment purge valve does not increase during the time the penetration is isolated. Since more reliance is placed on a single valve while in this Condition, it is prudent to perform the SR more often.

Therefore, a Frequency of once per 92 days was chosen and has been shown acceptable based on operating experience.

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

F.1 and F.2 If any Required Action and associated Completion Time cannot be met, the plant must be placed in a condition in which the LCO does not apply.

Action must be immediately initiated to suspend operations with a potential for draining the reactor (continued)

PCIVs B 3.6.1.3 BASES RIVER BEND B 3.6-22 Revision No. 110 ACTIONS F.1 and F.2 (continued) vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. If suspending the OPDRVs would result in closing the residual heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valves to OPERABLE status. This allows RHR to remain in service while actions are being taken to restore the valve.

SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR verifies that the 36 inch primary containment purge valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable. If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of the limits.

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 required to be capable of closing (e.g., during movement of recently irradiated fuel assemblies) pressurization concerns are not present and the purge valves are allowed to be open (automatic isolation capability would be required by SR 3.6.1.3.4 and SR 3.6.1.3.7).

The SR is modified by a Note (Note 2) stating that the SR is not required to be met when the purge valves are open for the stated reasons. The Note states that these valves may be opened for pressure control, ALARA, or air quality considerations for personnel entry or for Surveillances, or special testing on the purge system that require the valves to be open (e.g., testing of the containment purge radiation monitors). These primary containment purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements.

(continued)

Primary ContainmentShutdown B 3.6.1.10 BASES (continued)

RIVER BEND B 3.6-52 Revision No. 110 APPLICABILITY In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining 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 operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the primary containment. Due to radioactive decay, the primary containment is only required to be OPERABLE during fuel handling involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1.

ACTIONS A.1 and A.2 In the event that primary containment is inoperable, action is required to immediately suspend activities that represent a potential for releasing radioactive material, thus placing the unit in a Condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.

Also, if applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended.

SURVEILLANCE SR 3.6.1.10.1 REQUIREMENTS 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. Single isolation barriers that meet this criterion are a closed and de-activated power operated or automatic valve, a closed manual valve, a blind flange, or equivalent. This does not preclude the use of two active (ie, power operated and/or automatic) valves in the closed position for a given penetration. This SR does not require any testing or valve manipulation.

(continued)

Suppression Pool Water Level B 3.6.2.2 BASES RIVER BEND B 3.6-60 Revision No. 0 APPLICABLE Suppression pool water level satisfies Criteria 2 and 3 of Statement.

SAFETY ANALYSES (continued)

LCO A limit that suppression pool water level be 19 ft 6 inches and 20 ft 0 inches is required to ensure that the primary containment conditions assumed for the safety analysis are met. Either the high or low water level limits were used in the safety analysis, depending upon which is conservative for a particular calculation.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant loads on the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced because of the pressure and temperature limitations in these MODES. Requirements for suppression pool level in MODE 4 or 5 are addressed in LCO 3.5.2, "ECCSShutdown."

ACTIONS A.1 With suppression pool water level outside the limits, the conditions assumed for the safety analysis are not met. If 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/RV quenchers are covered. If suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and due to OPERABLE containment unit coolers. Prompt action to restore the suppression pool water level to within the normal range is prudent, however, to reduce the risks of increased pool swell and dynamic loading.

Therefore, continued operation for a limited time is allowed. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient to restore suppression pool water level to within specified limits. Also, it takes into account the low probability of an event impacting the suppression pool water level occurring during this interval.

(continued)

RPV Water Inventory Control

CRFA System B 3.7.2 BASES RIVER BEND B 3.7-12 Revision No. 132 APPLICABILITY In MODES 1, 2, and 3, the CRFA System must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA, since the DBA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the CRFA System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a.

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

b.

Dduring the movement of recently irradiated fuel assemblies in the primary containment or fuel building.

ACTIONS A.1 With one CRFA subsystem inoperable for reasons other than an inoperable CRE boundary, the inoperable CRFA subsystem must be restored to OPERABLE status within 7 days. With the unit in this condition, the remaining OPERABLE CRFA subsystem is adequate to perform the CRE occupant protection function. However, the overall reliability is reduced because a failure in the OPERABLE subsystem could result in loss of CRFA 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 required capabilities.

B.1, B.2, and B.3 If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated dose of the licensing basis analyses of DBA consequences (allowed to be up to 5 rem TEDE), or inadequate protection of CRE occupants from hazardous chemicals or smoke, the CRE boundary is inoperable. Actions must be taken to restore an OPERABLE CRE boundary within 90 days.

During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that in the event of a DBA, the mitigating actions will ensure that CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analyses of DBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke. These mitigating actions (i.e., actions that are taken to offset the consequences of the inoperable CRE boundary) should be preplanned for (continued)

CRFA System B 3.7.2 BASES RIVER BEND B 3.7-12a Revision No. 161 ACTIONS implementation upon entry into the condition, regardless of whether entry (continued) is intentional or unintentional. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions. The 90 day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.

C.1 In MODE 1, 2, or 3, if the inoperable CRFA subsystem or the CRE boundary cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE that minimizes overall plant risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 11) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action C.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

D.1, D.2.1, and D.2.2 The Required Actions of Condition D are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the (continued) and

CRFA System B 3.7.2 BASES RIVER BEND B 3.7-13 Revision No. 161 ACTIONS D.1, D.2.1, and D.2.2 (continued) 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 building or during OPDRVs, if the inoperable CRFA subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CRFA subsystem may be placed in the emergency mode. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk.

If applicable, movement of recently irradiated fuel assemblies in the primary containment or fuel building must be suspended immediately.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release.

Actions must continue until the OPDRVs are suspended.

E.1 If both CRFA subsystems are inoperable in MODE 1, 2, or 3, for reasons other than an inoperable CRE, the CRFA System may not be capable of performing the intended function and the unit is in a condition outside of the accident analyses.

Therefore, the plant must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 11) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

(continued) and

CRFA System B 3.7.2 BASES RIVER BEND B 3.7-14 Revision No. 159 ACTIONS F.1 and F.2 (continued)

During movement of recently irradiated fuel assemblies in the primary containment or fuel building or during OPDRVs, with two CRFA subsystems inoperable, or with one or more CRFA subsystems inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk.

If applicable, movement of recently irradiated fuel assemblies in the primary containment and fuel building must be suspended immediately.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. If applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand from the control room and continues to operate with flow through the HEPA filters and charcoal adsorbers. Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this system are not severe, testing each subsystem once every month provides an adequate check on this system. Furthermore, the 31 day Frequency is based on the known reliability of the equipment and the two subsystem redundancy available.

(continued)

Control Room AC System B 3.7.3 BASES (continued)

RIVER BEND B 3.7-18 Revision No. 110 LCO Two independent and redundant subsystems of the Control Room AC System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in the equipment operating temperature exceeding limits.

The Control Room AC 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, compressors, ductwork, dampers, and associated instrumentation and controls.

APPLICABILITY In MODE 1, 2, or 3, the Control Room AC System must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits.

In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a.

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

b.

During movement of recently irradiated fuel assemblies in the primary containment or fuel building.

ACTIONS A.1 With one control room AC subsystem inoperable, the inoperable control room AC subsystem must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE control room AC subsystem is adequate to perform the control room air conditioning function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of the control room air conditioning (continued) during

Control Room AC System B 3.7.3 BASES RIVER BEND B 3.7-19a Revision No. 161 ACTIONS The allowed Completion Time is reasonable, based on operating (continued) experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

D.1, D.2.1, and D.2.2 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply.

(continued) and

Control Room AC System B 3.7.3 BASES RIVER BEND B 3.7-20 Revision No. 115 ACTIONS D.1, D.2.1, and D.2.2 (continued)

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 building or during OPDRVs, if Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE control room AC subsystem may be placed immediately in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, movement of recently irradiated fuel assemblies in the primary containment and fuel building must be suspended immediately.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release.

Actions must continue until the OPDRVs are suspended.

E.1 and E.2 During movement of recently irradiated fuel assemblies in the primary containment or fuel building or during OPDRVs if the Required 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 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 building must be suspended immediately.

Suspension of these activities shall (continued) and

Control Room AC System B 3.7.3 BASES RIVER BEND B 3.7-21 Revision No. 161 ACTIONS E.1 and E.2 (continued) not preclude completion of movement of a component to a safe position.

Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analysis.

The SR consists of a combination of testing and calculation. The 24 month Frequency is appropriate since significant degradation of the Control Room AC System is not expected over this time period.

REFERENCES 1.

USAR, Section 6.4.

2.

USAR, Section 9.4.1.

3.

NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

AC Sources Shutdown B 3.8.2 RIVER BEND B 3.8-34 Revision No. 110 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC SourcesShutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC SourcesOperating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 and 5 SAFETY ANALYSES and during movement of recently irradiated fuel assemblies in the primary containment or fuel 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 inadvertent draindown of the vessel or a fuel handling accident.

In 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, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and 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.

During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS.

This allowance is in recognition that (continued)

AC Sources Shutdown B 3.8.2 RIVER BEND B 3.8-36 Revision No. 0 BASES LCO electrical power support, assuming a loss of the offsite circuit. Similarly, (continued) when the high pressure core spray (HPCS) is required to be OPERABLE, a separate offsite circuit to the Division III Class 1E onsite electrical power distribution subsystem, or an OPERABLE Division III DG, ensure an additional source of power for the HPCS. This additional source for Division III 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 Item 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.g., fuel handling accidents, reactor vessel draindown).

The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective 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. The offsite circuit consists of incoming breaker and disconnect to the respective preferred station service transformers 1C and 1D, the 1C and 1D preferred station service transformers, and the respective circuit path including feeder breakers to all 4.16 kV ESF buses required by LCO 3.8.10.

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 DG 1A and DG 1B and 13 seconds for DG 1C. 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 with the engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillance, e.g., 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 (continued)

AC Sources Shutdown B 3.8.2 RIVER BEND B 3.8-37 Revision No. 115 BASES LCO nonessential loads, is a required function for DG OPERABILITY. In (continued) addition, proper load sequence operation is an integral part of offsite circuit and DG OPERABILITY since its inoperability impacts the ability to start and maintain energized any loads required OPERABLE by LCO 3.8.10.

It is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply all required AC electrical power distribution 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 MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel building provide assurance that:

a.

Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b.

Systems needed to mitigate a fuel handling accident are available;

c.

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

d.

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

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

ACTIONS The ACTIONS 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 reactor shutdown.

(continued) that provide core cooling are available

AC Sources Shutdown B 3.8.2 RIVER BEND B 3.8-38 Revision No. 110 BASES ACTIONS A.1 (continued)

An offsite circuit is considered inoperable if it is not available to one required ESF division. If two or more ESF 4.16 kV buses are required per LCO 3.8.10, division(s) with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By the allowance of 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.2.4, B.1, B.2, B.3, and B.4 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. It is, therefore, required to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in the primary containment or fuel building, and activities that could potentially result in inadvertent draining of the reactor vessel.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize probability of the occurrence of postulated events. It is further required to initiate 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 (continued) and movement of recently irradiated fuel and

DC Sources Shutdown B 3.8.5 RIVER BEND B 3.8-59 Revision No. 110 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC SourcesShutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources Operating."

APPLICABLE The initial conditions of Design Basis Accident and transient analyses in SAFETY ANALYSES the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control and switching during all MODES of operation.

The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel 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 DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

The DC sources satisfy Criterion 3 of the NRC Policy Statement.

LCO One DC electrical power subsystem consisting of one battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus within the division, associated with Division (continued)

DC Sources Shutdown B 3.8.5 RIVER BEND B 3.8-60 Revision No. 110 BASES LCO I or II onsite Class 1E DC electrical power distribution subsystem(s)

(continued) required by LCO 3.8.10, "Distribution SystemsShutdown" is required to be OPERABLE. Similarly, when the High Pressure Core Spray (HPCS) system is required to be OPERABLE, the Division III DC electrical power subsystem associated with the Division III onsite Class 1E DC electrical power distribution subsystem required to be OPERABLE by LCO 3.8.10 is required to be OPERABLE. In addition to the preceding subsystems required to be OPERABLE, a Class 1E battery or battery charger and the associated control equipment and interconnecting cabling capable of supplying power to the remaining Division I or II onsite Class 1E DC electrical power distribution subsystem(s), when portions of both Division I and II DC electrical power distribution subsystems are required to be OPERABLE by LCO 3.8.10. 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.g.,

fuel handling accidents and inadvertent reactor vessel draindown).

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the primary containment or fuel building provide assurance that:

a.

Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b.

Required features needed to mitigate a fuel handling accident are available;

c.

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

d.

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

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.

(continued) core cooling

DC Sources Shutdown B 3.8.5 RIVER BEND B 3.8-61 Revision No. 115 BASES (continued)

ACTIONS The ACTIONS 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 reactor shutdown.

A.1, A.2.1, A.2.2, A.2.3, and A.2.4 If more than one DC distribution subsystem is required according to LCO 3.8.10, 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, fuel movement, and operations with a potential for draining the reactor vessel.

By allowing the option to declare required features inoperable with associated DC power source(s) inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. In many instances this option may involve undesired administrative efforts.

Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies, and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The 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 (continued) and

Inverters Shutdown B 3.8.8 RIVER BEND B 3.8-74 Revision No. 110 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Inverters Shutdown BASES BACKGROUND A description of the inverters is provided in the Bases for LCO 3.8.7, "Inverters Operating."

APPLICABLE The initial conditions of Design Basis Accident (DBA) and transient SAFETY ANALYSES accident analyses in the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature systems are OPERABLE.

The DC to AC inverters are designed to provide the required capacity, capability, redundancy, and reliability to ensure the availability of necessary power to portions of the ESF instrumentation and controls so that the fuel, Reactor Coolant System, and containment design limits are not exceeded.

The OPERABILITY of the inverters is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum inverters to each AC vital bus during MODES 4 and 5, and during movement of recently irradiated fuel assemblies in the primary containment or fuel building ensures that:

a.

The facility can be maintained in the shutdown or refueling condition for extended periods;

b.

Sufficient instrumentation and control capability are available for monitoring and maintaining the unit status; and

c.

Adequate power is available to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

The inverters were previously identified as part of the Distribution System and, as such, satisfy Criterion 3 of the NRC Policy Statement.

(continued)

Inverters Shutdown B 3.8.8 RIVER BEND B 3.8-75 Revision No. 115 BASES (continued)

LCO The inverters ensure the availability of electrical power for the instrumentation for systems required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or postulated DBA. The battery powered inverters provide uninterruptible supply of AC electrical power to the AC vital buses even if the 4.16 kV safety buses are de-energized. OPERABLE inverters require the associated AC vital bus be powered by the inverter through inverted DC voltage from the required Class 1E battery, or from an internal AC source via a rectifier with the battery available as backup. This ensures the availability of sufficient inverter power sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents and inadvertent reactor vessel draindown).

APPLICABILITY The inverters required to be OPERABLE in MODES 4 and 5 and also any time during movement of recently irradiated fuel assemblies in the primary containment or fuel building provide assurance that:

a.

Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b.

Systems needed to mitigate a fuel handling accident are available;

c.

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

d.

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

Inverter requirements for MODES 1, 2, and 3 are covered in LCO 3.8.7.

ACTIONS The ACTIONS 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 reactor shutdown.

(continued) that provide core cooling are available

Inverters Shutdown B 3.8.8 RIVER BEND B 3.8-76 Revision No. 110 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, and A.2.4 (continued)

If two divisions are required by LCO 3.8.10, "Distribution SystemsShutdown," the remaining OPERABLE inverters may be capable of supporting sufficient required feature(s) to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By the allowance of the option to declare required feature(s) inoperable with the associated inverter(s) inoperable, appropriate restrictions are implemented in accordance with the affected required feature(s) of the LCOs' ACTIONS. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in the primary containment and fuel building, and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required inverters and to continue this action until restoration is accomplished in order to provide the necessary inverter power to the plant safety systems.

Notwithstanding performance of the above conservative Required Actions, the unit is still without sufficient AC vital power sources to operate in a safe manner. Therefore, action must be initiated to restore the minimum required AC vital power sources and continue until the LCO requirements are restored.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required inverters should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power or powered from a constant voltage source transformer.

SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers (continued) and and movement of recently irradiated fuel

Distribution SystemsShutdown B 3.8.10 RIVER BEND B 3.8-89 Revision No. 110 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.10 Distribution SystemsShutdown BASES BACKGROUND A description of the AC, DC, and AC vital bus electrical power distribution systems is provided in the Bases for LCO 3.8.9, "Distribution SystemsOperating."

APPLICABLE The initial conditions of Design Basis Accident and transient analyses in SAFETY ANALYSES the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC, DC, and AC vital bus 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, DC, and AC vital bus electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC, DC, and AC vital bus 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 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 draindown of the vessel or a fuel handling accident.

The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement.

(continued)

Distribution SystemsShutdown B 3.8.10 RIVER BEND B 3.8-90 Revision No. 115 BASES (continued)

LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support required features. This LCO explicitly requires energization of the portions of the electrical distribution system necessary to support OPERABILITY of Technical Specifications' required systems, equipment, and components - both specifically addressed by their own LCOs, and implicitly required by the definition of OPERABILITY.

Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents and inadvertent reactor vessel draindown).

APPLICABILITY The AC, DC, and AC vital bus 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 building provide assurance that:

a.

Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b.

Systems needed to mitigate a fuel handling accident are available;

c.

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

d.

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

The AC, DC, and AC vital bus electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8.9.

(continued) that provide core cooling are available

Distribution SystemsShutdown B 3.8.10 RIVER BEND B 3.8-91 Revision No. 115 BASES (continued)

ACTIONS The ACTIONS 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 reactor shutdown.

A.1, A.2.1, A.2.2, A.2.3, A.2.4, and A.2.5 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in the primary containment and fuel building and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.

Notwithstanding performance of the above conservative Required Actions, a required residual heat removalshutdown cooling (RHR-SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS would not be entered. Therefore, Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable, which results in taking the appropriate RHR-SDC ACTIONS.

(continued) and movement of recently irradiated fuel and

Distribution SystemsShutdown B 3.8.10 RIVER BEND B 3.8-92 Revision No. 0 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, A.2.4, and A.2.5 (continued)

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.

SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the required AC, DC, and AC vital bus electrical power distribution subsystems are functioning properly, with the buses energized. The verification of proper voltage availability on the required buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The 7 day Frequency takes into account the redundant capability of the electrical power distribution subsystems, as well as other indications available in the control room that alert the operator to subsystem malfunctions.

REFERENCES

1.

USAR, Chapter 6.

2.

USAR, Chapter 15.

and

Inservice Leak and Hydrostatic Testing Operation B 3.10.1 RIVER BEND B 3.10-2 Revision No. 146 BASES APPLICABLE Allowing the reactor to be considered in MODE 4 when the reactor coolant SAFETY ANALYSES temperature is > 200°F, during, or as a consequence of, hydrostatic or leak testing, or as a consequence of control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test, effectively provides an exception to MODE 3 requirements, including OPERABILITY of primary containment and the full complement of redundant Emergency Core Cooling Systems (ECCS). Since the tests are performed nearly water solid, at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be very low. Under these conditions, the potential for failed fuel and a subsequent increase in coolant activity above the limits of LCO 3.4.8, "Reactor Coolant System (RCS) Specific Activity," are minimized. In addition, the secondary containment will be OPERABLE, in accordance with this Special Operations LCO, and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment described in Reference 2. Therefore, these requirements will conservatively limit radiation releases to the environment.

In the event of a large primary system leak, the reactor vessel would rapidly depressurize, allowing the low pressure core cooling systems to operate.

The capability of the low pressure coolant injection and low pressure core spray subsystems, as required in MODE 4 by LCO 3.5.2, "ECCS -

Shutdown," would be more than adequate to keep the core flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred.

For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and during postulated accident conditions.

As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply.

Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation at reactor coolant temperatures > 200°F, can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures (continued)

In the unlikely event of any primary system leak that could result in draining the RPV, the reactor vessel would depressurize.

The make-up capability required in MODE 4 by LCO 3.5.2, "RPV Water Inventory Control,"

would be more than adequate to keep the RPV water level above TAF under this low decay heat load condition.

v t e Letter Sequence Request
EPID:L-2017-LLA-0383, TSTF-542 (Approved, Closed)
Initiation Request Acceptance Supplement, Supplement Administration Questions Answers Results Approval Other:
MONTHYEAR2017-12-04 [Table View]

RBG-47801, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control

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RBG-47801, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control

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