Text

License Amendment Request Technical Specification (TS) Change Related to Revised Fuel Handling Accident Analysis and Adoption of TS Improvements Consistent with NUREG-1432
	ML19241A264
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/29/2019
From:	Gaston R; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	2CAN081903
	Download: ML19241A264 (96)
	v • d • e

Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 2CAN081903 August 29, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

License Amendment Request Technical Specification (TS) Change Related to Revised Fuel Handling Accident Analysis and Adoption of TS Improvements Consistent with NUREG-1432 Arkansas Nuclear One, Unit 2 NRC Docket No. 50-368 Renewed Facility Operating License No. NPF-6

REFERENCE:

U. S. Nuclear Regulatory Commission letter to Entergy Operations, Inc.,

"Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment RE: Use of Alternate Source Term" (TAC No. ME 3678) (2CNA041102) (ML110980197),

dated April 26, 2011 As required by 10 CFR 50.90, Entergy Operations, Inc. (Entergy), hereby requests NRC approval of proposed changes to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (TSs) to address an updated fuel handling accident (FHA) analysis and to adopt changes to gain greater consistency with NUREG 1432, Revision 4, "Standard Technical Specifications for Combustion Engineering Plants." NRC Safety Evaluation Report, Section 3.3.2 (Reference), discusses the updated ANO-2 FHA analysis.

Entergy requests to modify the TSs affected by the updated FHA analysis to gain further consistency with the NUREG 1432 improved technical specifications (ITS). The changes are most closely associated with previous changes made to NUREG 1432 via Technical Specification Task Force (TSTF) Travelers TSTF-51, TSTF-272, TSTF-268, TSTF-471, and TSTF-571-T. Entergy proposes to adopt changes consistent with NUREG 1432 and the related TSTFs in conjunction with changes necessary to support the updated FHA analysis. In so doing, a small number of additional TSs are affected in addition to those required to be changed in support of the updated FHA analysis.

2CAN081903 Page 2 of 3 Entergy presented a summary of the proposed changes in an NRC public teleconference (pre-submittal meeting) on June 13, 2019. Comments received during the teleconference and conversations held thereafter have been incorporated in the enclosed license amendment request.

The inclusion of the TSTF changes, which are intended to gain greater consistency with NUREG 1432 along with the needed TS Applicability changes in support of the updated FHA analysis, is considered appropriate due to the commonality of the changes, and to limit the number of submittals and submittal review fees that would be required to address each change via individual amendment requests.

The proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

The enclosure provides a description and assessment of the proposed changes. Attachment 1 of the enclosure provides the existing TS pages demarcated to show the proposed changes. of the enclosure includes a markup of the associated TS Bases pages, for information only. Attachment 3 of the enclosure provides the retyped (revised) TS pages. Note that TS Definitions Page 1-3 is currently under review by the NRC in association with the proposed adoption of TSTF-563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program." The markup and revised TS Page 1-3 contained in this letter do not include changes associated with TSTF-563.

New regulatory commitments are included in Attachment 4 of the enclosure to this letter.

Approval of the proposed amendment is requested by September 1, 2020. However, due to the number of changes involved, additional NRC review time may be required. Because Entergy has established administrative controls to ensure continued safe operation in support of the updated FHA analysis, approval beyond the requested September 1, 2020, date will not result in any adverse impact to nuclear or public safety. Once approved, the amendment shall be implemented within 90 days.

In accordance with 10 CFR 50.91, Entergy is notifying the State of Arkansas of this amendment request by transmitting a copy of this letter and enclosure to the designated State Official.

If there are any questions or if additional information is needed, please contact Tim Arnold, Manager, Regulatory Assurance, at 479-858-7826.

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

Executed on August 29, 2019.

Sincerely, ORIGINAL SIGNED BY RON GASTON Ron Gaston RWG/dbb

2CAN081903 Page 3 of 3

Enclosure:

Evaluation of the Proposed Change Enclosure Attachments:

1. Technical Specification Page Markups

2. Technical Specification Bases Page Markups (Information Only)

3. Retyped Technical Specification Pages

4. List of Regulatory Commitments cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Enclosure to 2CAN081903 Evaluation of the Proposed Change

Enclosure to 2CAN081903 Page 1 of 37 EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION The proposed license amendment would modify multiple Technical Specifications (TSs) for Arkansas Nuclear One, Unit 2 (ANO-2) to address non-conservative TS Applicabilities associated with the movement of fuel assemblies. This is necessary due to the previous adoption of Alternate Source Terms (AST) which included an update to the ANO-2 fuel handling accident (FHA) analysis. This update created a new requirement to address the movement of new (unirradiated) fuel assemblies over irradiated fuel assemblies.

In addition to the above, Entergy Operations, Inc. (Entergy) also requests changes in these and related TSs to gain greater consistency with NUREG 1432, Revision 4, "Standard Technical Specifications for Combustion Engineering Plants" (heretofore referred to as improved technical specifications or ITS). These changes can best be described referring to the previous Technical Specification Task Force (TSTF) travelers that update NUREG 1432 over the years. Therefore, background information related to the following TSTF travelers is included in this amendment request for ease of review.

ADAMS TSTF# Revision Title Approval Date Reference Revise Containment Requirements 51 2 During Handling Irradiated Fuel and 11/01/1999 ML993190284 Core Alterations Refueling Boron Concentration 272 1 12/21/1999 ML993630256 Clarification Operations Involving Positive 286 2 07/06/2000 ML003730788*

Reactivity Additions Eliminate use of term Core Alterations 471 1 12/07/2006 ML062860320 in Actions and Notes Revise Actions for Inoperable Source 571-T 0 10/04/2018** ML17346A587 Range Neutron Flux Monitor

Letter notifying the Nuclear Energy Institute (NEI) dated July 6, 2000.

- Letter notifying TSTF of release of TSTFs -51, -286, and -471 for adoption with reference to TSTF-571-T. This letter did not approve TSTF-571-T.

Entergy presented a summary of the proposed changes in an NRC public teleconference (pre-submittal meeting) on June 13, 2019 (reference ML19182A033 for meeting summary). The following comments (summarized) were received during the teleconference and have been incorporated in this license amendment request.

Verify matrix table (Section 2.4 of this enclosure) correctly corresponds to the appropriate TSTFs Verify Discussion of Differences (DODs) correctly correspond to the basis for each change (i.e., related to a TSTF, gain consistency with ITS, etc.).

Enclosure to 2CAN081903 Page 2 of 37 Provide appropriate justification for the existing 72-hour refueling canal boron sample frequency.

Ensure a clear line of sight with respect to how each applicable change meets the intent of the associated TSTF.

Ensure text box annotations on TS markup pages each have an arrow clearly demarcating which change the text box is associated with.

Include reference to TSTF markup pages and applicable NUREG 1432, Revision 4, pages.

2.0 DETAILED DESCRIPTION 2.1 Background The proposed changes were previously submitted beginning in April 2012 (Reference 2) following a pre-submittal meeting held on October 5, 2011 (Reference 1). However, after approximately two years of NRC review, Entergy withdrew the request due to NRC concern with the changes proposed for ANO-2 TS 3.9.2 (source range neutron flux monitors). A listing of correspondence related to the previous Reference 2 request is included in the Section 6.0 of this enclosure. A historical summary is included in the subject withdrawal letter (Reference 6).

The previous Reference 2 request was originated to resolve TS Applicability changes needed as a result of an updated FHA analysis associated with the adoption of AST for ANO-2. As presented in the ANO-2 AST license amendment request (LAR), Attachment 3, Section 2.2 (Reference 7), the original FHA analysis assumed failure of 60 fuel rods in a single fuel assembly. The updated analysis assumes the failure of all fuel rods in two assemblies (472 rods). The updated calculation was provided in Attachment 2 to a supplemental letter associated with the aforementioned AST amendment request (Reference 8). The NRC Safety Evaluation Report, Section 3.3.2 (Reference 9), discusses the updated analysis.

Most of the ANO-2 TSs affected by the updated FHA analysis are applicable "during the movement of irradiated fuel." Because the FHA analysis now assumes failure of the impacted fuel assembly in addition to the dropped fuel assembly, the associated TSs and TS Bases are revised to include the movement of new (unirradiated) fuel assemblies also, when being moved over irradiated fuel assemblies seated in the Spent Fuel Pool (SFP) or reactor pressure vessel.

The above portion of this LAR is required to address non-conservative TSs having an Applicability associated with the movement of fuel assemblies. Currently such operations are administratively controlled as described in NRC Administrative Letter (AL) 98-10, "Dispositioning of TSs that are Insufficient to Assure Plant Safety." Refueling and operating procedures were previously revised to ensure the movement of new fuel above irradiated fuel assemblies is controlled in accordance with the updated FHA analysis. In accordance with the guidance in AL 98-10, this portion of the LAR is required to resolve a non-conservative TS and is not a voluntary request from a licensee to change its licensing basis. Therefore, this portion of the request is not subject to forward fit considerations as described in the letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010 (ML01960180).

Enclosure to 2CAN081903 Page 3 of 37 In addition to the above and consistent with the aforementioned previous LAR (Reference 2),

Entergy requests to modify the TSs affected by the updated FHA analysis to gain further consistency with the NUREG 1432 ITS. The changes are most closely associated with previous changes related to NUREG 1432 via TSTF-51, TSTF-272, TSTF-268, TSTF-471, and TSTF-571-T. Entergy proposes to adopt changes consistent with NUREG 1432 and the related TSTFs in conjunction with changes necessary to support the updated FHA. In so doing, a small number of additional TSs are affected in addition to those required to be changed in support of the updated FHA analysis. For ease of review, a summary discussion of each of these TSTFs is included below.

TSTF-51 and TSTF-471 In the pressurized water reactor (PWR) ITS NUREGs (NUREG-1430, -1431, and -1432), the defined term "Core Alterations" appeared in the Actions and Surveillance Notes of less than ten specifications. The term is not used in any ITS Applicability statements and only occurs in the ANO-2 custom TS Applicability for Containment Penetrations (shutdown) and personnel communication requirements during fuel handling in the Containment Building. Suspending core alterations (as opposed to suspending the movement of fuel assemblies) has no effect on the initial conditions or mitigation of any Design Basis Accident (DBA) or transient. These requirements apply an operational burden with no corresponding safety benefit. Therefore, this TSTF removed the use of the defined term "Core Alterations" from the TSs.

TSTF-51 also provided a differentiation between "recently" irradiated fuel and irradiated fuel that is not considered "recently" irradiated. Following a unit shutdown or trip, there is a period of time in which an FHA could result in exceeding offsite or Control Room dose limits. Any fuel movement during this time period is considered the movement of "recently" irradiated fuel.

TSTF-51 lifted several restrictions on fuel movements following this initial time period where related offsite and/or Control Room dose limits would not be exceeded following an FHA.

TSTF-471 corrected an oversight related to implementation of TSTF-51. TSTF-51 was intended to delete any reference to the term "Core Alterations" in the TSs. Following TSTF-51 implementation, the industry noted other places in the TSs where the term continued to exist.

Therefore, TSTF-471 was written to eliminate these remaining references. The basis for this change is equivalent to the discussion of Core Alterations included under TSTF-51 above. The NRC approved this change for Calvert Cliffs, Unit 1, in TS Amendment 279 (Reference 24).

TSTF-272 TSTF-272 corrected a deficiency in the Refueling Boron TS 3.9.1. This TS limits the boron concentration of the Reactor Coolant System (RCS) and the refueling canal during refueling to ensure that the reactor remains subcritical while in Mode 6. However, with the reactor vessel head installed, no potential for dilution of the RCS via common communication with the refueling canal exists. In this condition, it is not necessary to place a limit on the boron concentration in the refueling canal. The Applicability is revised with a Note which states that the limits only apply to the refueling canal when those volumes are in communication with the RCS. This change is consistent with the intent of the TS and eliminates restrictions that have no impact on nuclear or personnel safety.

The NRC approved similar changes for Millstone 2 in TS Amendment 263 (Reference 20).

Enclosure to 2CAN081903 Page 4 of 37 TSTF-286 and TSTF-571-T The NRC notified NEI in letter dated July 6, 2000 (Reference 16) that TSTF-286 had been approved for incorporation into the TSs. During the development of ITS Revision 3, the NRC requested that the template wording "cause introduction into the RCS, coolant with boron concentration less than required to meet the minimum SDM of LCO" be changed to "cause introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO" This request has been included in the ANO-2 TS application of the ITS wording contained in this letter.

This TSTF addressed Limiting Conditions for Operation (LCOs) and Actions related to positive reactivity additions to the RCS. The purpose of the TSTF was to permit operators to control RCS inventory and temperature while maintaining positive control of core reactivity. Therefore, most related TSs removed requirements that prevented "positive reactivity addition" or "reduction in boron concentration" in lieu of a more generic requirement that ensured Shutdown Margin (SDM) would be maintained during RCS inventory or temperature adjustments.

In a letter from the NRC to the TSTF dated November 7, 2013 (Reference 10), the NRC staff expressed concern with the Actions related to conditions where one source range neutron flux monitor is inoperable. During movement of fuel assemblies, sources, and reactivity control components with one monitor inoperable, there is the potential for the operable monitor to become effectively decoupled from the core reactivity condition. For example, if one monitor is inoperable, and certain, strategically located fuel assemblies are removed, the operable monitor may no longer be capable of monitoring the reactivity condition of fuel assemblies that are located in the far half of the core.

TSTF-571-T modified the source range neutron flux monitor TS to prohibit the movement of fuel assemblies, sources, and reactivity control components when a source range monitor is inoperable. A provision was included to allow such movement, if it is needed, to repair the inoperable monitor or to place a reactivity-related component in a safe condition. The changes ensure that no actions are taken that could alter the core reactivity when a monitor is inoperable. In letter dated October 4, 2018 (Reference 11), the NRC concluded that adoption of TSTF-571-T would resolve the concern discussed above.

2.2 Current TS Requirements The majority of the TSs affected by the proposed changes contain an Applicability similar to:

During movement of irradiated fuel in the reactor pressure vessel or During core alterations In addition, most of the associated TS Actions are similar to:

Suspend all operations involving CORE ALTERATIONS or positive reactivity changes or Suspend all activities involving the handling of irradiated fuel

Enclosure to 2CAN081903 Page 5 of 37 The above are related to TSs affected by the updated FHA analysis, TSTF-51, TSTF-286, and TSTF-471. With respect to TSTF-272, the current ANO-2 LCO 3.9.1, "Boron Concentration,"

states, in part:

With the reactor vessel head unbolted or removed, the boron concentration of the reactor coolant and the refueling canal shall be maintained uniform and With the reactor vessel head atop the reactor vessel, the refueling canal is not in communication with the RCS, whether or not the head is bolted. TSTF-272 corrects the wording to account for such conditions.

TSs affected by the updated FHA analysis or ITS consistency changes are at times modified to address editorial errors, redundant information, or other items as discussed in Section 3.0 of this letter. The markup of the TS pages included in Attachment 1 of this enclosure provides a detail of each change with direct reference to the reason for the change (whether related to the FHA analysis, a TSTF, or other type change). See Section 2.4 below for a listing of TSs affected by the proposed amendment.

2.3 Reason for the Proposed Change As discussed previously, the major change associated with this amendment request is required in support of an update to the FHA analysis which was part of the previous adoption of AST for ANO-2. In order to save significant, time, personnel resources, and costs, and due to the number of TSs affected, Entergy is also proposing changes to fuel handling and core reactivity related TSs to gain greater consistency with the requirements of the ITS.

FHA related changes are required because the updated FHA analysis now assumes the drop of new (unirradiated) fuel assembly onto irradiated fuel assemblies as part of the dose consequence analysis. The current TSs associated with fuel handling activities only address the movement of irradiated fuel assemblies.

The changes supporting greater consistency with the ITS and other proposed changes provide significant enhancement to the associated TSs in order to permit ease of Operator understanding and compliance with the individual requirements. In addition, some of the changes enhance the margin to safety by providing additional flexibility to affect proper control of plant parameters during certain conditions.

In light of the above, Entergy proposes to revise the subject TSs to both resolve the non-conservative aspect previously described and to gain greater consistency with NUREG 1432.

2.4 Description of the Proposed Change The proposed change resolves associated non-conservatism in specific TSs and provides several other enhancements. Because of the number of TSs affected, a specific description of each individual change is not included here; however, each change is effectively illustrated in of this enclosure, along with specific reference to the type of change (i.e., whether the change is related to ITS consistency, editorial, FHA analysis, etc.). The following are generic examples of the changes:

Enclosure to 2CAN081903 Page 6 of 37 FHA Analysis during movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies Note that the above also replaces any use of the defined term "Core Alterations" which is deleted from TS Section 1.0, "Definitions," in NUREG 1432 by TSTF-471.

TSTF-51 and TSTF-471 During CORE ALTERATIONS movement of recently irradiated fuel assemblies or movement of new fuel assemblies over recently irradiated fuel assemblies in the reactor pressure vessel.

TSTF-272 With the reactor vessel head unbolted or removed, Tthe boron concentration of the reactor coolant and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met MODE 6*

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removedOnly applicable to the refueling canal when connected to the RCS.

As will be discussed later, the deletion of the phrase "With the reactor vessel head unbolted or removed" is unrelated to TSTF-272, but is requested since this phrase is redundant to the TS Definition of MODE 6.

TSTF-286 suspend all operations that would causeinvolving introduction of coolant into the RCS witha reduction in boron concentration less than required to meet SDM of LCO 3.1.1.1of the Reactor Coolant System TSTF-571-T Suspend movement of fuel, sources, and reactivity control components within the reactor vessel.

A listing of each TS affected by this amendment request is included below, along with the types of changes affecting each TS that are illustrated in Attachment 1 of this enclosure. Attachment 1 of this enclosure includes a tie to a DOD included in Section 3.0 below. Some TSs may not appear to be directly impacted by the FHA or an associated TSTF. For example, an ANO-2 TS may have referred to "Core Alterations" which would require change under TSTF-51 or TSTF-471 (replacing the term with "the movement of recently irradiated fuel"). The change under the TSTF would then require the wording to be modified to that needed in support of the updated FHA analysis. In these cases, the demarcation in the following table is enclosed in parenthesis.

Enclosure to 2CAN081903 Page 7 of 37 Because ANO-2 has not converted the TSs to the ITS format, an explanation of minor wording differences between the changes proposed and the original TSTF markups is appropriate. Entergy does not consider such changes to be deviations from NUREG 1432 because the changes act to accommodate the difference in format between the ANO-2 TSs and ITS, and the difference in TS usage rules between the two TS types. Any significant differences are specifically identified in the Attachment 1 TS page markups and discussed in detail in Section 3.0 of this enclosure.

Note the table column labeled "C" represents a technical correction or clarification. A label of "E" indicates an editorial correction or administrative enhancement. These are the same identifiers utilized in the TS markups contained in Attachment 1.

TSTFs FHA 51 272 286 471 571-T C E TS 1.12 CORE ALTERATION X TS 3.1.1.3 RCS Dilution Flow Rate (X) X X TS 3.3.3.1 Radiation Monitoring X X X X TS 3.4.1.2 RCS Loops - Mode 3 X X TS 3.4.1.3 RCS Loops - Modes 4 and 5 X X X TS 3.7.6.1 CREVS and CREACS X TS 3.8.1.2 AC Sources - Shutdown X X X X TS 3.8.2.2 AC Distribution - Shutdown X X X X X TS 3.8.2.4 DC Sources/Distribution -

X X X X Shutdown TS 3.9.1 Boron Concentration X X X TS 3.9.2 Source Range Instrumentation X (X) X X TS 3.9.4 Containment Building Penetration X X X TS 3.9.5 Communications X (X)

TS 3.9.8.1 Shutdown Cooling - Normal Level X TS 3.9.9 Water Level - Reactor Vessel X X

Enclosure to 2CAN081903 Page 8 of 37

3.0 TECHNICAL EVALUATION

FUEL HANDLING ACCIDENT The updated FHA analysis assumes damage to both the dropped fuel assembly and the impacted fuel assembly. Therefore, TS conditions and/or applicabilities must be revised to include the movement of any fuel assembly (new or irradiated fuel) over irradiated fuel assemblies. The proposed amendment addresses the ANO-2 updated FHA analysis based on the previously approved use of AST. The reanalysis of a fuel drop (and resultant damage to fuel rods) considered the weight of other components beyond the fuel assembly itself (hoist grapple, control rods) in accordance with NRC Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

As part of the AST adoption, Westinghouse performed an updated FHA analysis for ANO-2.

The calculation and results were provided in the proposed ANO-2 amendment to adopt the use of AST. As presented in the ANO-2 AST amendment request, Attachment 3, Section 2.2 (Reference 7), the previous FHA analysis assumed failure of 60 fuel rods in a single fuel assembly. The updated analysis assumes the failure of all fuel rods in two assemblies, the dropped assembly and the impacted assembly (472 rods). The revised calculation was provided in Attachment 2 to a supplemental letter associated with the aforementioned AST amendment request (Reference 8). The NRC Safety Evaluation (SE), Section 3.3.2 (Reference 9), discusses the revised analysis. Excerpts from this section of the SE state:

The licensee assumed that, as a design basis, spent fuel will have decayed for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after shutdown before being handled. The licensee assumed that 472 pins will be damaged (two full fuel assemblies), as the result of the postulated FHA, and thus release all of its available gap activity over a 2-hour period. This bounds the current ANO-2 licensing basis, as described in ANO-2 FSAR Section 15.1.23, and is consistent with the guidance expressed in RG 1.183.

The licensee concluded that the radiological consequences at the EAB, lPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the FHA are a TEDE of 6.3 rem at the EAB for the worst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 6.3 rem at the outer boundary of the LPZ, and 5 rem in the control room for the duration of the accident. Based on the above discussion of the licensee's FHA analysis, the NRC staff concludes that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2 FSAR as design bases. The NRC staff also performed independent calculations to verify the conservatism of certain parameters used by the licensee. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.2 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2 in this SE. The staff concludes that the EAB, LPZ, and control room doses estimated by the licensee for the FHA accident meet the applicable accident dose criteria and are, therefore, acceptable.

NRC letter dated October 4, 2018 (Reference 11), required (in part) that licensees applying for adoption of TSTF-51 describe any limitations or controls that would prevent the movement of unirradiated fuel assemblies, sources, or reactivity control components capable of damaging a fuel assembly located within the core prior to the time period defined as "recently". For ANO-2,

Enclosure to 2CAN081903 Page 9 of 37 consistent with the assumptions of the updated FHA analysis performed in support of the adoption of AST, "recently" is defined as less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months post-shutdown. Following this 100-hour period, an FHA cannot result in the release of radioactivity from the site that would exceed 10 CFR 50.67 limits, assuming no containment. The 100-hour post-shutdown assumption is governed by ANO-2 TS 3.9.3.a, which requires the reactor to be subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to moving irradiated fuel within the reactor vessel. The updated FHA analysis, which assumes a fuel assembly drop including the weight of a control rod and the hoist grapple, bounds the drop of any other component which could be manipulated within the core.

In addition, a review of the required refueling preparations following shutdown does not support a condition that would permit the manipulation of fuel, sources, or reactivity control components within the core in less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months (~4 days) post-shutdown. For example, the schedule for the fall 2018 ANO-2 refueling outage included nearly 6 full days of normal activities from reactor shutdown to a condition where the refueling canal was flooded and the upper guide structure removed, allowing access to the core internals.

Currently, most affected TSs are applicable "during the movement of irradiated fuel." Because the updated FHA analysis now assumes failure of the impacted assembly in addition to the dropped assembly, the associated TSs and TS Bases are revised to include the movement of new fuel assemblies also, when being moved over irradiated fuel assemblies seated in the SFP or reactor pressure vessel.

Note that some TS "applicability" statements are included or repeated in Actions, Notes, or Surveillance Requirements (SRs). Therefore, the list of affected TSs includes the areas of the associated TS that contains the phrase "during movement of irradiated fuel" or similar. The applicability statements, regardless of their location within the individual TS, are revised to state, or be similar to, the following:

"during the movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies" The revised wording accounts for the possibility of incurring fuel damage should a fuel assembly (new or irradiated) drop, impacting an irradiated fuel assembly seated in the SFP or reactor pressure vessel. This change was not previously required because the previous FHA analysis only assumed damage to the dropped assembly (no damage occurred to the fuel assembly seated in the SFP or reactor pressure vessel).

Consideration of Other Load Drops To determine the appropriate TS applicability wording that accommodates the updated FHA analysis, Entergy considered other loads that could travel over irradiated fuel assemblies. While the previous analysis conservatively assumed 60 fuel rods were damaged following an FHA, the actual results indicated no fuel damage, based on an assembly weight of approximately 1451 lbs (the fuel assembly design used today weighs approximately 1419 lbs, reference ANO-2 Safety Analysis Report (SAR) Table 4.2-1). Both the previous and updated FHA analyses assume a vertical drop of a long slender object impacting directly onto an assembly seated in the reactor vessel or the SFP, with the revised analysis adding the weight of the grapple and a control element assembly (CEA). Therefore, any similar object weighing less than approximately 1451 lbs was bounded by the previous FHA analysis.

Enclosure to 2CAN081903 Page 10 of 37 The updated FHA analysis provided by Westinghouse during the adoption of AST evaluated a non-fuel load drop of 2000 lbs. The load was considered either of long slender geometry or cube geometry. The analysis concluded that the most significant fuel damage occurred if the load remained vertical and the energy from the load drop was transferred to a single assembly.

Even so, only the fuel rods in the impacted assembly were damaged since the "dropped" load did not involve fuel (236 fuel rods, one half of that assumed in the revised offsite dose analysis).

Note that loads greater than 2000 lbs in the Containment Building are prohibited from being suspended or traveling over fuel assemblies at ANO without a specific evaluation intended to verify that the updated FHA analysis results are not exceeded in the event of a load drop.

Loads greater than 2000 lbs are prohibited from being suspended or traveling over fuel assemblies in the SFP in accordance with TS 3.9.7.

The following information was considered to determine whether "other load" drops need be considered in the application of this amendment request:

1. The FHA (fuel assembly drop analysis) as addressed in the TSs has historically only considered fuel assemblies. Attempting to address "unknown" loads would be difficult and could insert ambiguity into the TSs. In addition, the TS Bases would have to contain a wealth of information to interpret the TS, virtually placing the TS itself under licensee control.

2. No "heavy load" (> 2000 lbs) was identified during a review of common loads transported over fuel seated in the core (with the reactor vessel head removed) or the SFP. Note that the "dummy" assembly at ANO-2 weighs approximately 1330 lbs.

3. The updated FHA analysis indicates that impact over two seated assemblies, or any rotation or horizontal drop of the load, would not substantially damage fuel.

4. The load drop analysis does not credit any cushioning of the impact by the upper end fittings on the seated assembly nor from the upper portion of the fuel storage racks in the SFP.

5. Administrative procedures such as OP-1005.002, "Control of Heavy Loads," work to limit movement of any object over the fuel, both for load drop concerns and also to comply with foreign material exclusion processes. The procedure also lists the weight of 38 Containment Building and many more SFP area loads that routinely are moved (mostly during outage). The SFP work platform is the only load that may be located over fuel during fuel shuffle that is in the above weight range, but is a horizontal structure. Individual procedures, such as OP-2402.079, "Operation of the Containment Polar Crane (2L2),"

establish safe load paths and "no fly" zones for subject loads. For example, "no-fly" zones in the Containment Building include the area directly over the reactor vessel and over the south piping penetration area (where safety related piping is located). Finally, ANO-2 complies with NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."

6. Entergy provided response to NRC Bulletin 96-02, "Movement of Heavy Loads over Spent Fuel, over Fuel in the Reactor Core, or over Safety-Related Equipment," for ANO-2 regarding control of such loads in letter dated May 17, 1996 (Reference 12). The NRC did not list ANO-2 as a plant with open issues in its summary of responses (Reference 13).

7. As stated previously, the previous FHA analysis indicated that no fuel damage occurred (although 60 fuel rods were conservatively assumed to fail, but only within the "dropped" assembly).

Enclosure to 2CAN081903 Page 11 of 37 Based on the above, Entergy believes the drop of "other loads" (non-fuel loads) need not be addressed further in support of this amendment request.

As stated previously, Entergy is using this opportunity to revise certain ANO-2 TSs to gain greater consistency with the ITS. The justification for these changes is best described by reference to TSTF travelers that updated NUREG-1432 in previous years, along with the current ITS Bases associated with a given TS. Therefore, the following discussions focus on the purpose of the individual TSTFs and ITS Bases.

TSTF-51, REVISION 2 and TSTF-471, REVISION 1 The defined term "Core Alterations" appeared in the Actions and Surveillance Notes of less than ten specifications in the pressurized water reactor (PWR) ITS NUREGs. The term is not used in any ITS Applicability statements and only occurs in the old ANO-2 custom TS Applicability for Containment Penetrations (shutdown) and the personnel communication requirements during fuel handling in the Containment Building (non-ITS). Suspending Core Alterations (as opposed to suspending the movement of fuel assemblies) has no effect on the initial conditions or mitigation of any DBA or transient. These requirements apply an operational burden with no corresponding safety benefit. Therefore, this TSTF removed the use of the defined term "Core Alterations" from the TSs. This TSTF was approved for industry adoption on November 1, 1999 (ML993190284).

The ANO-2 TSs define "Core Alterations: as:

CORE ALTERATION shall be the movement or manipulation of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

Core Alterations take place during Mode 6 (refueling operation). There are three accidents considered for ANO-2 during Mode 6: 1) a FHA, 2) a boron dilution event, and 3) failure of the Shutdown Cooling (SDC) loop.

The analysis for an FHA conservatively assumes that a fuel assembly is dropped during fuel handling activities in the Containment Building or the SFP area of the Auxiliary Building.

Interlocks and procedural/administrative controls make such an event highly unlikely. However, if an assembly were damaged (i.e., one or more fuel rods damaged), the accumulated fission product gases and iodines in the fuel element gap would be released to the surrounding water.

Release of solid fission products from the fuel would be negligible because of the low fuel temperature that exists in Mode 6, which greatly limits diffusion. There are no mitigation actions credited in the analysis to reduce the FHA offsite dose consequences. The Control Room Emergency Ventilation System (CREVS) provides protection for Control Room operators during all previously evaluated accidents and is required by TSs to be operable during the movement of fuel assemblies, regardless of operational mode. Manipulation of non-fuel core components (CEAs, sources, incore instruments, etc.) is not assumed to cause failure of the fuel clad.

Enclosure to 2CAN081903 Page 12 of 37 Based on the above, the "Core Alterations" term serves no significant purpose with regard to the FHA-related margin of safety and may be replaced with criteria associated only with the movement of fuel assemblies.

A boron dilution event is initiated by a water source that results in the boron concentration approaching or dropping below the value required to maintain TS required SDM. This event is mitigated by stopping the dilution. The suspension of Core Alterations has no significant effect on restoration of the TS required SDM. In addition, the location or movement of core components does not affect the initiation of, or mitigation of, a boron dilution event. Note that the boron dilution event is considered unlikely for ANO-2 due to the significant period of time for operator detection (increase in RCS or canal level, drop in other inventories) and response (isolate source and/or borate the RCS) before SDM would be significantly challenged (reference ANO-2 SAR Section 15.1.4.3). When filled, the refueling canal holds over 430,000 gallons of borated water, not counting the volume contained within RCS piping; therefore, a significant amount of unborated water would be required to reduce the refueling boron concentration just one ppm from the TS 3.9.1 minimum boron concentration limit of 2500 ppm. The TS 3.9.1 required reactivity condition of 2500 ppm boron or a keff of 0.95 provides significant margin to inadvertent criticality. In addition, while all CEAs "could" be removed without resulting in criticality based on the Mode 6 boron concentration, multiple CEAs are normally not removed from the core during refueling outages. CEA movements are normally performed in the SFP.

This provides additional SDM beyond that provided by the required boron concentration.

In accordance with procedures, a loss of SDC requires cessation of all refueling activities once any suspended fuel assembly has been placed in a safe condition. In addition, OP-2203.029, "Loss of Shutdown Cooling," requires evacuation of the Containment Building and closure of the Containment building upon loss of SDC. However, the loss of SDC and subsequent potential heat-up of the RCS has minimal impact on core reactivity and, therefore, is unrelated to Core Alterations, especially core component manipulations other than movement of fuel assemblies.

TS 3.9.8.1 (SDC) requires positive reactivity additions to be secured when no SDC loop is in operation, with the exception that SDC may be secured for up to one hour per 8-hour period without cessation of core alterations. Based on the significant boron concentration present in the RCS during Mode 6 operation, core component or fuel assembly manipulation with or without SDC would have little effect on overall core reactivity with regard to maintaining sufficient SDM. In light of required corrective actions taken upon a loss of SDC and because a loss of SDC has minimal impact on core reactivity, the deletion of the "Core Alterations" has no significant effect in relation to this event.

Because the initial condition assumptions in the safety analyses for an FHA, boron dilution event, and loss of SDC event continue to be met, the scenarios wherein these accidents could occur, and the required operability of the associated systems, are not reduced by the deletion of "Core Alterations" term. In addition, either the suspension of fuel movement and/or suspension of unrestricted dilution activities, as applicable, is required in cases where the "Core Alterations" term is proposed for deletion.

TSTF-471 corrected an oversight related to implementation of TSTF-51 above. TSTF-51 was intended to delete reference to the term "Core Alterations" in the TSs. Following TSTF-51 implementation, the industry noted other places in the TSs where the term continued to exist.

Therefore, TSTF-471 was written to eliminate these remaining references. The basis for this change is equivalent to the discussion of core alterations included under TSTF-51 above.

Enclosure to 2CAN081903 Page 13 of 37 TSTF-51 also changed the applicability of several TSs from "during the movement of irradiated fuel" to "during the movement of recently irradiated fuel." Most plants, including ANO-2, performed revised offsite dose analyses associated with an FHA in the 1990s and discovered offsite dose remained acceptable for an FHA that occurred long after reactor shutdown. For ANO-2, the decay time necessary to result in acceptable offsite dose limits upon an FHA, assuming no containment and no filtration, is 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after shutdown. Because of these re-analyses, TSTF-51 permits plants to insert the term "recently" in appropriate TS Applicabilities, Conditions, and SRs, which would permit associated TS equipment to be inoperable during the movement of fuel assemblies provided the decay time requirement had been met. However, based on the updated FHA discussed previously, the dose consequences are not acceptable with regard to ANO-2 Control Room personnel following the 100-hour period.

Because the General Design Criterion (GDC)-19 Operator dose criteria cannot be met without reliance on the CREVS for filtration, Entergy is not adopting the "recently" term of the TSTF for TSs associated with Control Room radiation monitoring or the CREVS. Likewise, since Control Room cooling may be needed during CREVS operation post-FHA, the "recently" term is also not adopted for the Control Room Emergency Air Conditioning System (CREACS). Note that the dose analysis results for the updated FHA, crediting CREVS, are contained in ANO-2 SAR Table 15.1.23-2.

An NRC reviewers note contained in TSTF-51 requires adoption of two commitments:

1) a commitment to maintain the ability to close containment building penetrations following an FHA, and 2) a commitment to maximize availability of radiation monitoring and ventilation systems that can aid in mitigating the offsite dose consequences following an FHA. In accordance with the TSTF, Entergy is adopting these commitments (see Attachment 4).

The proposed TS change regarding the elimination of the "Core Alterations" term from TSs facilitates refueling operations during Mode 6 and provides operational flexibility during core alterations activities. Since the requirements to suspend the movement of irradiated fuel assemblies within the Containment Building (or cessation of unrestricted RCS dilution, as applicable) remains, the elimination of TS Actions such as "suspend Core Alterations," and other references to "Core Alterations" has no effect on the initial conditions or mitigation of any DBA or transient. The proposed changes associated with this TSTF are consistent with the intent of the ITS (and TSTF-51/TSTF-471) and eliminates restrictions that have no impact on nuclear or personnel/public safety.

TSTF-272, REVISION 1 This TSTF was approved for industry adoption on December 21, 1999 (ML993630256).

This TSTF corrected a deficiency in Refueling Boron TS 3.9.1. This TS limits the boron concentration of the RCS and the refueling canal during refueling to ensure that the reactor remains subcritical while in Mode 6. However, with the reactor vessel head installed, no potential for dilution of the RCS via common communication with the refueling canal exists. In this condition, it is not necessary to place a limit on the boron concentration in the refueling canal. The Applicability is revised with a Note which states that the limits only apply to the refueling canal when this volume is in communication with the RCS.

Enclosure to 2CAN081903 Page 14 of 37 ANO-2 SR 4.0.4 (ITS SR 3.0.4) states, in part, that entry into a MODE or other specified condition in the Applicability shall only be made when the LCOs Surveillances have been met within the specified Frequency. ANO-2 TS 3.9.1, SR 3.9.1.1, requires verification of SDM (boron concentration) prior to removing or unbolting the reactor vessel head. In addition, ANO-2 SR 3.9.1.2 requires boron concentration in the RCS and refueling canal to be verified every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, compliance with SR 4.0.4 ensures boron requirements in both the RCS and refueling canal are met prior to establishing communication between the two volumes.

Note that the 72-hour boron concentration verification frequency is sufficient based on the large volume of water contained in the refueling canal and, subsequently, the significant amount of unborated water that would be required to reduce concentration in a manner that could challenge the required SDM (see discussion under TSTF-51 above for additional detail).

Entergy is also adopting the TSTF-272 insert associated with the TS 3.9.1 Bases as shown in the Attachment 1 markup (editorials discussed in DOD section later):

Prior to reconnecting portions of the refueling canal to the RCS, this SR must be met per SR 4.0.4. If any dilution activity has occurred while the refueling canal was disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

The proposed changes associated with this TSTF are consistent with the intent of the ITS (and TSTF-272) and eliminates restrictions that have no impact on nuclear or personnel/public safety.

TSTF-286, REVISION 2 As stated previously, this TSTF was approved for industry adoption on March 20, 2000. Note, however, that the reference letter to NEI was dated July 6, 2000 (ML003730788). During the development of ITS Revision 3, the NRC requested that the template wording "cause introduction into the RCS, coolant with boron concentration less than required to meet the minimum SDM of LCO" be changed to "cause introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO" This request has been included in the ANO-2 TS application of this ITS wording contained in this letter.

TSTF-286 addressed LCOs and Actions related to positive reactivity additions to the RCS. The purpose of the TSTF was to permit Operators to control RCS inventory and temperature while maintaining positive control of core reactivity. Therefore, most related TSs removed requirements that prevented "positive reactivity addition" or "reduction in boron concentration" in lieu of a more generic requirement that ensured SDM would be maintained during RCS inventory or temperature adjustments.

TSTF-286 allows licensees to revise the plant TS LCO Actions and Notes that require suspension of operations involving positive reactivity additions or preclude reduction in boron concentration by placing a limit on positive reactivity addition to within the TS-required SDM limit. TSTF-286 thus provides the flexibility necessary for continued safe reactor operations, while also limiting any potential for excessive positive reactivity addition to the core. During conditions in which these Actions may be required, various activities for unit operation must be continued to maintain RCS inventory and control RCS temperature. The activities that involve inventory makeup from sources with boron concentrations less than the current RCS concentration (i.e., boron dilution) need not be precluded in the TSs provided the required SDM

Enclosure to 2CAN081903 Page 15 of 37 is maintained for the worst-case overall effect on the core. Note that an unexpected boron dilution event is considered unlikely for ANO-2 due to the significant period of time for Operator detection and response before SDM would be significantly challenged (reference ANO-2 SAR Section 15.1.4.3).

The proposed changes associated with this TSTF are consistent with the intent of the ITS (and TSTF-286) and eliminates restrictions that do not have a significant impact on nuclear or personnel/public safety.

TSTF-571-T, REVISION 0 During movement of fuel assemblies, sources, and reactivity control components with one monitor inoperable, there is the potential for the operable monitor to become effectively decoupled from the core reactivity condition. For example, if one monitor is inoperable, and certain, strategically located fuel assemblies are removed, the operable monitor may no longer be capable of monitoring the reactivity condition of fuel assemblies that are located in the far half of the core.

The installed source range neutron flux monitors are Gamma Metrics detectors operating in the proportional region of the gas filled detector characteristic curve. The detectors monitor the neutron flux in counts per second. The instrument range indicated in the Control Room is 1 to 1E+6 cps and includes audible count rate. The detectors also provide continuous visual indication in the control room and an audible alarm to alert the operators. In Mode 6, the monitors must be operable to determine changes in core reactivity. There is no other direct means available to monitor core reactivity levels.

ANO-2 TS 3.9.2 currently requires the cessation of positive reactivity additions during periods when one source range neutron flux monitor is inoperable. TSTF-571-T modifies this TS by adding action to prohibit the movement of fuel assemblies, sources, and reactivity control components when a monitor is inoperable. TSTF-571-T also provides a provision which allows such movement, if it is needed, to repair the inoperable monitor or to place reactivity control components in a safe condition. The changes ensure that actions are limited which could alter the core reactivity when a monitor is inoperable. In letter dated October 4, 2018 (Reference 11),

the NRC stated, in part, that licensees pursuing adoption of TSTF-286 should also adopt the changes included in TSTF-571-T, the latter of which addresses previously NRC-identified concerns with TSTF-286 as it relates to source range monitors.

The proposed changes associated with this TSTF are consistent with the intent of the ITS (and TSTF-571-T), providing conservative actions to minimize the potential for an undetected reactivity change to occur and enhancing the maintenance of nuclear and personnel/public safety.

Enclosure to 2CAN081903 Page 16 of 37 DISCUSSION OF DIFFERENCES (DODs)

The following TSs changed by TSTFs subject to this amendment request do not exist in the ANO-2 TSs and, therefore, no corresponding changes are made herein. This difference does not impact the acceptability of the changes which are contained in this amendment request nor alter the intent of the ITS.

ITS# Title 3.3.8 FBACS Actuation Instrumentation 3.7.13 Fuel Building Air Cleanup System (FBACS) 3.8.8 Inverters - Shutdown The reference numbers for each bulleted item below directly correlate with numbering denoted in text boxes in the associated marked-up TS pages of Attachment 1.

1. The TSTF-51 insertion term "recently" is not included in Note 2 of Table 3.3-6 (TS Page 3/4 3-25) or the associated Action 21 (TS Page 3/4 3-26a) since this Note and Action are associated with the CREVS radiation monitors. "Recently" is defined for ANO-2 as the movement of fuel within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of reactor shutdown. The CREVS radiation monitors act to automatically isolate the Control Room and start the emergency filtration system upon receipt of a high radiation signal. Although offsite dose consequences are acceptable following 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of reactor shutdown, the dose to Control Room personnel is not acceptable without crediting the filtration capability of the CREVS. Therefore, the CREVS radiation monitors will continue to be required to be operable during the movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies, regardless of the time after shutdown, in order to provide Control Room Operators radiological protection during FHA events. This is more restrictive than the allowances provided by the ITS (TSTF-51) and is considered an acceptable difference.

TSTF-51 also made like changes to ITS 3.7.11, "Control Room Emergency Air Cleanup System (CREACS)" (equivalent to the ANO-2 CREVS) and ITS 3.7.12, "Control Room Emergency Air Temperature Control System (CREATCS)" (equivalent to the ANO-2 Control Room Emergency Air Conditioning System or CREACS). The "recently" term is also not adopted for ANO-2 TS 3.7.6.1, "Control Room Emergency Ventilation and Air Conditioning System," consistent with the above discussion.

TSTF-51 is adopted, however, for AC and DC power CREVS TS support systems. ANO, Unit 1 (ANO-1) TSs require CREVS operability in addition to the ANO-2 TSs, since CREVS is a shared system for the common ANO-1 and ANO-2 Control Room envelope.

In addition, the TS Definition of "operability" requires all necessary support systems to also be operable, irrespective of the support system's TS Applicability. Therefore, adoption of TSTF-51 within the CREVS AC/DC power support systems will not prevent required support systems to be in an operable state needed to support CREVS operability, whenever the associated CREVS train is required to be operable.

Enclosure to 2CAN081903 Page 17 of 37

2. ANO-2 TS 3.1.1.3, "Boron Dilution," does not exist in the ITS. This TS is intended to verify mixing flow is available (minimum of 2000 gpm) whenever a reduction in RCS boron is being performed. The Action associated with this TS is similar to other TSs (both ITS and ANO-2 specific) in that a reduction in boron concentration is not allowed when the minimum RCS flow rate is not met. TSTF-286 modified this Action type to allow boron reductions provided SDM requirements are maintained. Entergy proposes to adopt the TSTF-286 changes for this Action. This change meets the purpose of TSTF-286 which was intended to permit Operators to control RCS inventory and temperature while maintaining positive control of core reactivity.

Note that the TSTF-286 related markups refer to only the Modes 3, 4, and 5 SDM requirements (ITS 3.1.1). Because ANO-2 TS 3.1.1.3 is applicable in all Modes, references to all ANO-2 SDM-related TSs are included in the Action markup. Because one of these "SDM" TSs states a boron concentration (TS 3.9.1), the TSTF-286 wording is modified to state that either the SDM or boron concentration requirements are maintained, depending on which ANO-2 SDM-related TS is applicable for the given mode of operation.

This is an administrative difference which ensures the intent of TSTF-286 is maintained.

3. With reference to ITS 3.4.6, "RCS Loops - Mode 4," the TSTF-286 related markups refer to the SDM requirements of ITS 3.1.1, which are applicable in Modes 3, 4, and 5.

Because the ANO-2 SDM TS for Mode 5 is separate from the SDM requirements for other Modes, reference to both ANO-2 SDM TSs are included in the Action markup. This is an administrative difference which ensures the intent of TSTF-286 is maintained.

In addition, the ITS contains individual TSs for RCS Loops in Mode 4, Mode 5 - Loops Filled, and Mode 5 - Loops Not Filled. All Mode 4 and Mode 5 RCS Loop requirements are contained in a single ANO-2 TS 3.4.1.3, "Reactor Coolant System - Shutdown."

Therefore, the TSTF-286 RCS Loop related changes are effectively captured in this single ANO-2 TS for the three aforementioned ITS Modes.

4. With reference to ITS 3.9.1, "Boron Concentration," TSTF-272 added a Note that stated the LCO was only applicable when the "refueling canal and reactor cavity" were connected to the RCS. ANO-2 TSs and procedures, with respect to RCS and boron concentrations, refer only to the "refueling canal," i.e., the reactor cavity is assumed to be enveloped within the meaning of "refueling canal." At ANO-2, there is no physical separation between the refueling canal and the reactor cavity. In addition, the equivalent ITS LCO contains these terms in brackets, as each would be site specific. Based on the above, the "reactor cavity" phrase is omitted from adoption of the ITS TSTF-272 equivalent. This is an administrative/editorial difference and continues to meet the intent of TSTF-272.

5. With reference to ITS 3.9.2, "Nuclear Instrumentation," TSTF-286 permitted RCS dilutions provided SDM was maintained (reference Required Action A.2). However, TSTF-471 revised Required Action A.1 to suspend all positive reactivity additions. Therefore, ITS Required Action A.1 now makes ITS Required Action A.2 mute as no dilutions may occur.

In light of this conflict, Entergy proposes to adopt only the intent of Required Action A.1 as this is more restrictive than Required Action A.2.

Entergy is also adopting changes consistent with TSTF-571-T in TS 3.9.2 in order to address a previous generic NRC concern with the movement of reactivity components when a source range neutron flux monitor is inoperable. ANO-2 TS 3.9.2 Action a is

Enclosure to 2CAN081903 Page 18 of 37 expanded to require the suspension of movement of fuel, sources, and reactivity control components within the reactor vessel when one or more monitors are inoperable. A Note 1 is also added to Action a, consistent with the Note provided in TSTF-571-T, permitting movement of reactivity-related components, if necessary to place in a safe condition or if necessary to support repair of an inoperable monitor.

The associated TSTFs and the ITS refer to only one monitor being inoperable when applying these actions. This is because ITS usage rules require entry into all Conditions which may apply at a given time. The ANO-2 standard TS usage rules, in general, require entry only into those actions that specifically meet the described plant condition.

Therefore, if both ANO-2 monitors were inoperable, Action a may not necessarily be entered as it currently only addresses a single monitor being inoperable. However, the intent of the ITS is that these actions (suspend positive reactivity additions and suspend movement of reactivity-related components) should be performed when one or both monitors are inoperable. To ensure the intent of the ITS is addressed, ANO-2 TS 3.9.2 Action a is modified to state (addition underlined): "With one or more of the above required monitors inoperable". Because the intent of TSTF-571-T and the ITS is maintained, Entergy considers this difference to be acceptable.

The ANO-2 TSs require a Channel Functional Test of the source range detectors within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of Core Alterations. This test is not required by the ITS and is, therefore, more restrictive than the ITS. Consistent with the intent of TSTF-51 and TSTF-471, "Core Alterations" is replaced with reference to the movement of recently irradiated fuel, modified to accommodate the updated ANO-2 FHA analysis. Because the intent of TSTF-51 and TSTF-471 is met, and because the revision accounts for the updated FHA analysis, Entergy considers this difference to be acceptable.

6. TSTF-51 modified the Applicability and Condition A of ITS 3.9.3, "Containment Penetrations." However, the corresponding ANO-2 TS 3.9.4, "Containment Building Penetration," repeats reference to "Core Alterations" in SR 4.9.4.1, which was removed from the ITS Applicability and Condition by TSTF-51. Therefore, Entergy proposes to modify ANO-2 SR 4.9.4.1 consistent with the modifications performed to the ITS by TSTF-51, while also accommodating the updated ANO-2 FHA analysis. Because the intent of TSTF-51 is maintained, Entergy has determined that this difference is acceptable.

7. ANO-2 TS 3.9.5, "Communications," does not exist in the ITS. This TS requires communication between the Control Room and fuel handling personnel in the Containment Building when on-loading or off-loading the core. The TS, however, refers to "Core Alterations," which was deleted from the ITS via TSTF-51 and TSTF-471. Entergy has modified TS 3.9.5 consistent with the intent of TSTF-51, TSTF-471, and the updated ANO-2 FHA analysis, with exception that the term "recently" is not incorporated. Omission of the "recently" term is considered appropriate since direct communication between the Control Room and fuel handling personnel supports coordination of fuel movement and can be utilized to promptly report adverse conditions, regardless of the decay time that has expired. The changes affect the Applicability, Action, and SR contained within the TS.

Because the intent of the ITS is maintained and the updated FHA analysis requirements are accommodated, Entergy considers this difference to be acceptable.

Enclosure to 2CAN081903 Page 19 of 37

8. ANO-2 TS 3.9.8.1, "Shutdown Cooling - One Loop," Action b, refers to "Core Alterations."

This TS is similar to ITS 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level," which requires only one SDC loop to be operable. Neither TSTF-51 or TSTF-471 included a markup of this ITS page because the "Core Alterations" defined term did not exist in ITS 3.9.4. Consistent with these TSTFs, "Core Alterations" is removed from this Action with the incorporation of TSTF-286. The revised Action wording meets the intent of TSTF-51 in that Action a prevents the addition of decay heat load to the core, similar to ITS 3.9.4, Required Action A.3, which states: "Suspend loading irradiated fuel assemblies in the core." Based on the above, the revised wording meets the intent of TSTF-51 and is consistent with the similar requirements of ITS 3.9.4.

The Applicability and Action statements of ANO-2 TS 3.9.9, "Water Level - Reactor Vessel," in part refers to the movement of CEAs instead of the pre-TSTF-51 ITS wording of "Core Alterations." Consistent with TSTF-51, reference to CEAs is removed since TSTF-51 removed reference to "Core Alterations," which included the movement of CEAs in its definition. This difference is consistent with the intent of TSTF-51.

ADMINISTRATIVE/EDITORIAL/CLARIFICATION CHANGES The numbering used in the DOD section above is continued through this section for ease in review. These numbers correlate directly with numbering denoted in text boxes placed in the associated marked-up TS pages of Attachment 1.

9. TS 3.1.1.3 "Boron Dilution" and TS 3.9.8.1, "Shutdown Cooling - One Loop" Because TSTF 286 (and the ITS) utilizes the acronym "RCS" in lieu of "reactor coolant system,"

TS 3.1.1.3 and TS 3.9.8.1 use of the "reactor coolant system" phrase is replaced with "RCS".

This change is administrative in nature. Note that "Reactor Coolant System" is shown deleted as part of changes associated with TSTF 286 adoption.

10. Table 3.3-6 and 4.3-3 "Radiation Monitoring Instrumentation" In reference to ANO-2 TS Table 3.3-6, Item 2a (TS Page 3/4 3-25), a new Note 3 is proposed that clarifies the original intent of the TS requirements for radiation monitoring and automatic isolation of the Containment Purge system. Currently the TS requires operability of these features in Modes 5 and 6. However, the Containment Purge system is not always in operation.

As written, the TS would require the radiation monitoring and isolation capability to remain operable even when the Containment Purge system is secured. The addition of Note 3 specifies that operability is required only during 1) Containment Purge operations, or 2) ongoing Containment Building continuous ventilation operations when moving recently irradiated fuel assemblies or moving new fuel assemblies over irradiated fuel assemblies in the Containment Building, consistent with the updated FHA, the ITS, and TSTF-51.

Table 3.3-6 Actions 16a, 16b, and 16c (TS Page 3/4 3-26a), and Notes 2 and 3 of Table 4.4-3 (TS Page 3/4 3-27) are revised to capitalize noun names and/or TS definitions. Wording is added such as "the Containment Building" to enhance sentence clarity. The ending phrase "in the Containment Building" is also adopted for the Containment Purge radiation monitor

Enclosure to 2CAN081903 Page 20 of 37 functional test, consistent with the ITS, which ensures the channel requirements are appropriately associated with the movement of fuel within the Containment Building. These changes are administrative in nature.

Action 16c (TS Page 3/4 3-26) is revised to delete the specific Offsite Dose Calculation Manual (ODCM) Limitation reference, which is currently incorrect. This level of detail is not necessary to properly apply the Action. Deletion of this excessive detail removes the licensee burden of requesting a change to the TS when indexing changes are made to the ODCM. This change is administrative in nature.

Notes 2 and 3 of Table 4.4-3 (TS Page 3/4 3-27) are revised to clearly differentiate between "purge" and "continuous ventilation" modes of operation with regard to the Containment Purge ventilation system and radiation monitoring instrumentation. The once per 12-hour Channel Check and once per 31-day Channel Functional Test should be associated with continuous ventilation operations, since purge operations only occur once in an outage and normally lasts no more than an hour or two. Both notes still require a Channel Check and Channel Functional Test within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or 31 days prior to initiating a purge, respectively. A "purge" is treated as a potentially significant radioactive release, similar to gaseous and liquid release from radioactive holding tanks at the facility. Once the purge is complete, the Containment Building atmosphere is free from appreciable amounts of radioactivity and the ventilation system may be continuously operated thereafter for human comfort purposes. Since an FHA or other event involving radiological release could occur when operating in the continuous ventilation mode, it is appropriate to maintain operability of the associated radiation monitoring instrumentation until the system is secured. The aforementioned changes ensure the instrumentation is verified to remain operable at set intervals.

Likewise, the "Modes in which Surveillance Required" column for testing of the Containment Purge features depicted in Table 4.3-3 (TS Page 3/4 3-27) is revised from the current "Modes 5 and 6", to "in accordance with applicable Notes" associated with each required test. In conjunction with this change, the Channel Calibration is revised with the addition of reference to Note 3, which ensures the calibration is completed prior to purge operations during a refueling outage. The new reference to Note 3 requires the calibration to be performed within 31 days prior to initiating a Containment purge, consistent with the requirement for the Channel Functional Test. Any additional requirements for the Channel Calibration would fall under the Surveillance Frequency Control Program (SFCP). Entergy believes this change better meets the intent of ensuring appropriate operability of the necessary features prior to relying on these features to perform their necessary function.

In addition to the above, TS Page 3/4 3-28 is blank and is, therefore, deleted. A Note is added to the footer of Page 3/4 3-27 stating that the next page is 3/4 3-36. This change is administrative in nature.

Based on the above, Entergy believes the proposed changes ensure proper and conservative application of the subject TSs and are, therefore, justified.

Enclosure to 2CAN081903 Page 21 of 37

11. TS 3.1.1.3 "Boron Dilution," TS 3.4.1.2 "RCS - Hot Standby," TS 3.4.1.3 "RCS -

Shutdown," TS 3.8.1.2 "A.C. Sources - Shutdown," and TS 3.9.8.1, "Shutdown Cooling -

One Loop" With respect to TS 3.4.1.2 and 3.4.1.3:

1. The term "operable" is now capitalized in the ANO-2 RCS cooling loops TS 3.4.1.2. This is a TS defined term. All defined terms are capitalized throughout the TSs.

2. The term "in" is deleted from the TS 3.4.1.2 LCO statement. The presence of this term is an editorial error.

3. The term "all" is removed from the Actions of these TSs to be consistent with TSTF-286 and ITS wording. This term was not required to meet the intent of the Action and inclusion of this term is inconsistent with ITS wording.

With respect to TS 3.8.1.2, the term "immediately" is added and the phrase "all operations" is deleted for consistency with the ITS. The intent of the Action is unaltered by these changes.

Entergy considers these changes to be administrative.

12. TS 3.8.2.2 "A.C. Distribution - Shutdown" The phrase "and energized" and term "any" is removed from the Action associated with this TS.

A bus that is de-energized cannot be performing its specified safety function and, therefore, cannot be operable. This phrase does not appear in the other shutdown electrical TSs or the ITS. The deletion of the term "any" does not alter the intent of the Action and provides consistency with ITS wording. Entergy considers this change administrative/editorial in nature in that the specified safety function of the required equipment is unchanged by removal of this phrase.

13. TS 3.9.1 "Boron Concentration" For consistency with other ANO-2 TSs and the ITS, reference to "reactor coolant" in the LCO of ANO-2 TS 3.9.1 is revised to "reactor coolant system". This change is administrative in nature.

The current LCO for ANO-2 TS 3.9.1 includes the following phrase:

With the reactor vessel head unbolted or removed In addition, the current Applicability Note in ANO-2 TS 3.9.1 states:

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

Enclosure to 2CAN081903 Page 22 of 37 These statements are a repeat of the TS Definition of "Mode 6". Section 1 of the ANO-2 TSs (Definitions), Table 1.1 states:

6 REFUELING** 0.95 0 140°F

- Reactor vessel head unbolted or removed and fuel in the vessel.

No other Mode 6 related ANO-2 TSs or the ITS include this redundant information. Therefore, the current TS 3.9.1 Applicability Note is replaced with the Note from TSTF-272, which eliminates the above redundant phrase. In addition, the similar phrase in the LCO of TS 3.9.1 is deleted. Entergys adoption of the TSTF-272 Note is in accordance with the TSTF. The deletion of redundant information is an administrative change and in no way changes the current requirements of the TS.

14. TS 3.9.4 "Containment Building Penetration" For consistency, "containment" is revised in the Applicability, Action, and SR of ANO-2 TS 3.9.4 to "Containment Building." This is an editorial change only and has no impact on the intent of the TS.

15. TS 3.9.9 "Water Level - Reactor Vessel" With regard to reactor vessel water level, the phrase "elevation corresponding to the" top of irradiated fuel is added to the LCO. This ensures that proper water level is established prior to initiating refueling of the reactor core following a defueled condition. At this point in an outage, there is no fuel in the core and, based on current LCO wording, verifying 23 feet of water "above fuel in the core" would not be possible. The addition of the aforementioned phrase removes ambiguity in this regard. Entergy considers this change to be an enhancement.

The movement of fuel "within the reactor vessel" contained in the Applicability and Action of this TS is revised to "within the Containment Building." This reference is also added to the SR. The required water level should be met even when fuel is being moved in other areas of the refueling canal, not just in the reactor vessel. In addition, the phrase "while in Mode 6" is deleted from the Applicability since fuel assemblies cannot be accessed within the reactor until Mode 6 has been achieved (reference TS Definition for Mode 6 which in part states "reactor vessel head removed"). Therefore, referencing this mode of operation in the Applicability is redundant. Entergy considers these changes to be enhancements.

Other Administrative Differences The following ITS (or associated TSTF) requirements are not contained in the ANO-2 TSs and, therefore, no changes to these TSs are proposed with respect to the subject TSTFs. These differences do not conflict with the intent of the subject TSTFs.

TSTF-51 deleted the phrase "Core Alterations" from the Containment Purge and CREVS radiation monitoring requirements in ITS 3.3.8 and 3.3.9, respectively. Because this phrase is not present in the respective ANO-2 TS 3.3.3.1 for either system, the deletion of this phrase was not necessary.

Enclosure to 2CAN081903 Page 23 of 37 TSTF-51 made changes to the ITS 3.3.10, "Fuel Handling Isolation Signal (FHIS) and ITS 3.7.14, "Fuel Building Air Cleanup System (FBACS). The ANO-2 TSs do not contain similar TSs; therefore, no ANO-2 TS-related changes are required.

ITS SR 3.3.8.4 requires a Channel Functional Test of Containment Purge Isolation System monitors. TSTF-471 removed the "Core Alterations" phrase from the SR Note, leaving reference to only the movement of irradiated fuel assemblies. ANO-2 TS 3.3.3.1 (Table 4.3-3) does not contain reference to "Core Alterations"; therefore, no TSTF-471 related changes are required for the ANO-2 SR (see Instrument 2.a in Table 4.3-3).

ITS 3.3.9, "Control Room Isolation Signal" (with reference to TSTF-286), includes changes associated with the cessation of any reduction in RCS boron concentration.

ANO-2 TS 3.3.3.1, "Radiation Monitoring Instrumentation," does not contain any related action; therefore, changes associated with TSTF-286 are not required for this TS.

ITS 3.3.13, "Logarithmic Power Monitoring Channels" (with reference to TSTF-286),

includes changes associated with the cessation of any reduction in RCS boron concentration. ANO-2 TS 3.3.1.1, "Reactor Protective Instrumentation," does not contain any related action; therefore, changes associated with TSTF-286 are not required for this TS.

TSTF-51 removed "Core Alteration" references and requirements from ITS 3.7.11, "Control Room Emergency Air Cleanup System (CREACS)" (equivalent to the ANO-2 CREVS) and ITS 3.7.12, "Control Room Emergency Air Temperature Control System (CREATCS)" (equivalent to the ANO-2 Control Room Emergency Air Conditioning System or CREACS). ANO-2 TS 3.7.6.1, "Control Room Emergency Ventilation and Air Conditioning System," does not contain reference to "Core Alterations"; therefore no TSTF-51 related changes associated with the deletion of "Core Alterations" are required for ANO-2 TS 3.7.6.1.

TSTF-51 made changes to the ITS 3.8.8, "Inverters - Shutdown." TSTFs 286 and 471 likewise made changes to this inverter TS. The ANO-2 TSs do not contain a similar TS; therefore, no ANO-2 TS related changes are required.

TSTF-286 made changes to ITS 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation

- Low Water Level." However, the corresponding ANO-2 TS 3.9.8.2 does not contain actions associated with no coolant flow. Actions related to conditions where coolant flow is less than that needed for proper mixing is addressed by ANO-2 TS 3.1.1.3 (see DODs 2 and 9 above). ANO-2 TS 3.1.1.3 is being revised as part of this amendment request, consistent with TSTF-286. TS 3.1.1.3 requires RCS dilutions to be controlled consistent with TSTF-286 when SDC flow is less than 2000 gpm; this is more restrictive than the ITS 3.9.5, Condition B, bases of no flow being available. Based on the above, no changes are proposed for ANO-2 TS 3.9.8.2.

Enclosure to 2CAN081903 Page 24 of 37 Conclusion The proposed changes are consistent with the bases/justification of the referenced TSTFs, the ITS, the aforementioned NRC Split Report, and/or the updated ANO-2 FHA analysis.

Therefore, Entergy has concluded that the proposed changes are acceptable and do not result in a significant impact to nuclear or public safety. A markup of the proposed TS changes in their entirety is included in Attachment 1 of this enclosure and a clean (revised) version of the affected TS pages included in Attachment 3. The TS markup pages include text boxes linked to a DOD discussed above, and where associated with the ITS or a TSTF, the corresponding Combustion Engineering Owners Group (CEOG) TSTF or ITS page is also referenced within the text box, where applicable.

Note that TS Definitions Page 1-3 is currently under review by the NRC (Reference 25) in association with the proposed adoption of TSTF-563. The markup and revised TS Page 1-3 contained in this letter do not include changes associated with TSTF-563.

An information-only markup of the TS Bases is provided in Attachment 2 of this enclosure. The ANO-2 TS Bases are not as detailed as the ITS Bases and, therefore, changes are made only where the detail contained in associated TSTF markups is contained within the ANO-2 TS Bases. The affected TS Bases will be revised in accordance with the TS 6.5.14, "Technical Specification (TS) Bases Control Program" and in accordance with 10 CFR 50.59 upon approval of this amendment request. This is considered a commitment as included in Attachment 4 of this submittal.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(b) Technical Specifications, states that TSs shall be derived from the analyses and evaluation included in the safety analysis report. The proposed change, in part, aligns the ANO-2 TSs with the updated FHA analysis.

10 CFR 50.36(c) states that TSs are required to include items in the following categories:

(1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation (LCOs), (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls. 10 CFR 50.36(c)(2)(i) states that the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility and that when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. The proposed change maintains LCOs and remedial actions to be followed if an LCO is not met.

10 CFR 50.36(c)(3) requires TSs to include SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met. The proposed change maintains appropriate SRs that meet this regulatory requirement.

Enclosure to 2CAN081903 Page 25 of 37 10 CFR 50.36(a)(1) states, in part: "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." TS bases associated with the proposed change have been developed and included in the amendment request, where applicable, for information only.

10 CFR 50, Appendix A, General Design Criteria (GDC), contains specific design requirements for nuclear power plants. The following provides discussion of applicable GDCs along with any other related regulation or guidance.

GDC 17 requires onsite electric power systems be provided with sufficient capacity and capability to assure that (1) specified acceptable fuel design limits (SAFDLs) and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

GDC 19 provides requirements for maintaining a habitable control room and includes limitations on radiological dose that may be received by control room operators.

In accordance with GDC 26, GDC 28, and GDC 29, reactivity shall be controllable, such that subcriticality is maintained under cold conditions and SAFDLs are not exceeded during normal operation and anticipated operational occurrences.

GDC 41 requires containment atmosphere cleanup systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

GDC 61 requires that the fuel storage and handling systems be designed to assure adequate safety under normal and postulated accident conditions.

GDC 62, "Prevention of criticality in fuel storage and handling," requires that criticality be prevented by physical systems and processes.

GDC 64 requires the means for monitoring the reactor containment atmosphere effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

The acceptance limits for offsite radiation exposure (with respect to ANO-2) are contained in 10 CFR 50.67 and NRC Regulatory Guide (RG) 1.183.

RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Revision 0, provides assumptions used for evaluating the potential radiological consequences of a fuel handling accident (FHA) in the spent fuel pool (SFP) area.

The SFP / refueling canal level limitation is such that sufficient iodine activity would be retained to limit offsite doses from the accident to within 10 CFR 50.67 limits.

Enclosure to 2CAN081903 Page 26 of 37 The proposed change does not alter the design of ANO-2. As a result, the applicability of the GDC and the RG is not affected.

Section IV, The Commission Policy, of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, published in the Federal Register on July 22, 1993 (58 FR 39132), states, in part:

The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.

[T]he Commission will also entertain requests to adopt portions of the improved, even if the licensee does not adopt all STS improvements.

In accordance with this Policy Statement, improved STS have been developed and will be maintained for each NSSS [nuclear steam supply system] owners group. The Commission encourages licensees to use the improved STS as the basis for plant-specific Technical Specifications.

The proposed change does not affect compliance with the regulations or regulatory guidance, is consistent with the Standard Technical Specification (STS), and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Precedent With respect to the TS changes associated with the updated FHA analysis, the most similar submittal was provided by the Waterford Steam Electric Station Unit 3 (Reference 17) as approved by the NRC on April 25, 2012 (Reference 18).

Most plants adopted the equivalent of TSTFs -51, -272, and -286 during TS conversion to ITS (some also included TSTF-471). With respect to specific adoption of each TSTF, the following references are relevant to this application.

TSTF-51 January 22, 2002, Watts Bar Nuclear Plant, Unit 1 (Reference 19)

TSTF-272 January 11, 2002, Millstone Nuclear Power Station, Unit No. 2 (Reference 20)

TSTF-286 June 8, 2011, Ginna Nuclear Power Plant (Reference 21)

June 28, 2006, Millstone Power Station, Unit Nos. 2 and 3 (Reference 22)

Enclosure to 2CAN081903 Page 27 of 37 TSTF-471 February 15, 2007, Point Beach Nuclear Plant, Units 1 and 2 (Reference 23)

September 21, 2006, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (Reference 24) 4.3 No Significant Hazards Consideration Analysis Entergy Operations, Inc. (Entergy) has evaluated the proposed changes to Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification (TSs) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

Entergy proposes a change to the ANO-2 TSs to resolve non-conservatisms identified as a result of an update to the stations Fuel Handling Accident (FHA) analysis performed in support of the previous adoption of Alternate Source Terms. This change modifies the TS limitations specific to the movement of fuel assemblies. Generic modifications to the TSs are also incorporated to improve consistency with NUREG 1432, "Standard Technical Specifications for Combustion Engineering Plants," Revision 4 (commonly referred to as the improved technical specifications or ITS). In addition, administrative changes are proposed such as the correction of errors, removal of ambiguity, or otherwise enhancements in TS wording.

Basis for no significant hazards consideration determination:

As required by 10 CFR 50.91(a), Entergy analysis of the issue of no significant hazards consideration (NSHC) is presented below. Note that although several changes are a result of conforming more closely to NUREG 1432, the following analysis of the references Technical Specification Task Force (TSTF) travelers that previously revised NUREG 1432.

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

Response: No Updated FHA TS changes associated with the updated FHA analysis ensure the initial assumptions of the FHA are maintained and, therefore, act to minimize the consequences of an accident by ensuring TS required features are operable during the movement of fuel assemblies. The updated FHA analysis was previously accepted by the NRC during adoption of Alternate Source Terms (AST) for ANO-2. The probability of a fuel assembly drop (or any load drop) is unchanged by the updated FHA analysis. Therefore, the updated FHA analysis does not involve a significant increase in the probability of an accident previously evaluated.

Entergy has reviewed station procedures and controls in order to verify that no other loads, other than a new or irradiated fuel assembly, need be addressed with regard to an FHA (i.e., no other known load carried over irradiated fuel assemblies exists which would not be bounded by the fuel drop analysis or be expected to cause fuel damage if dropped). The proposed TS changes ensure required systems are operable during operations that could

Enclosure to 2CAN081903 Page 28 of 37 lead to an FHA. As previously approved by the NRC via the adoption of AST for ANO-2, the updated FHA analysis adequately bounds Control Room and offsite dose within federal limitations. Based on the above, the proposed FHA-related changes to the TSs do not result in a significant increase in the consequences of an accident previously evaluated.

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

TSTF-51 and TSTF 471 The design basis accident (DBA) assumed for ANO-2 related to the proposed changes is the FHA. The boron dilution event is evaluated in the ANO-2 Safety Analysis Report (SAR), but considered an unlikely event due to the time available for operator detection and response, along with prevalent administrative controls. A loss of Shutdown Cooling (SDC) event has little relationship to and minimal impact with regard to an FHA. TSTF-51 and TSTF-471 replace the use of the previously defined "core alterations" term with requirements associated with the movement of fuel assemblies, since the drop of a fuel assembly is the only event that could reasonably lead to an FHA or a significant challenge to the plant.

In addition, TSTF-51 reduces restrictions following sufficient radioactive decay of fuel assemblies since the offsite dose consequences of an FHA following this decay period (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for ANO-2) would remain within 10 CFR 50.67 limits. Note that this allowance is not adopted for TS Control Room ventilation or radiation monitoring systems (associated with meeting 10 CFR 50, Appendix A, General Design Criteria (GDC) 19).

The removal of references to "core alterations" in favor of restrictions associated with the movement of fuel assemblies eliminates current restrictions associated with the manipulation of other core components (i.e., sources or reactivity control components within the core) since such manipulation cannot result in an FHA, boron dilution event, or loss of SDC. In addition, manipulation of these other components cannot present a significant challenge to shutdown margin (SDM) because the TS required RCS boron concentration for Mode 6 operation provides substantial margin to criticality.

Changes associated with TSTF-51 and TSTF-471, as adopted, do not modify limitations in such a way that the consequences of an FHA would be greater than that assumed in the updated FHA analysis (i.e., 10 CFR 50.67 and GDC 19 limitations are not exceeded following an FHA).

Based on the above, the proposed changes associated with the adoption of TSTF-51 and TSTF-471 do not result in a significant increase in the probability or consequences of an accident previously evaluated.

TSTF-272 Changes associated with TSTF-272 place additional restrictions on Mode 6 operations by ensuring the boron concentration of the water in the refueling canal meets the same TS limits required for the Reactor Coolant System (RCS) when the RCS is in direct hydraulic

Enclosure to 2CAN081903 Page 29 of 37 communication with the refueling canal (i.e., reactor vessel head removed and refueling canal filled). These changes are unrelated to any accident initiator and further prohibit any challenge to the fuel in the reactor vessel by ensure sufficient boron concentration is maintained during Mode 6 operations. Therefore, these changes do not result in a significant increase in the probability or consequences of an accident previously evaluated.

TSTF-286 Changes associated with TSTF-286 permit operator control of RCS inventory and temperature when certain TS requirements are not met, provide the overall required SDM of the RCS is maintained. The activities that involve inventory makeup from sources with boron concentrations less than the current RCS concentration (i.e., boron dilution) need not be precluded in the TSs provided the required SDM is maintained for the worst-case overall effect on the core. Note that an unexpected boron dilution event is considered unlikely for ANO-2 due to the significant period of time for operator detection and response before SDM would be significantly challenged (reference ANO-2 Safety Analysis Report Section 15.1.4.3). In addition, while a boron dilution event is evaluated in the accident analysis, the only "accident" assumed for ANO-2 during Mode 6 operations is the FHA.

Permitting RCS inventory and temperature adjustments is unrelated to any assumptions associated with an FHA. Therefore, these changes do not result in a significant increase in the probability an accident (or a boron dilution event) previously evaluated. Because an unexpected boron dilution event provides sufficient opportunity for detection and recovery, the proposed changes associated with TSTF-286 likewise do not result in a significant increase in the consequences of an accident (or boron dilution event) previously evaluated.

TSTF-571-T The proposed change revises the Actions for inoperable source range neutron flux monitors to prohibit the movement of fuel assemblies, sources, and reactivity control components when monitor is inoperable. The Actions taken when a monitor is inoperable are not initiators to any accident previously evaluated. The monitors are not credited to mitigate any previously evaluated accident. The proposed change restricts the licensee's actions while a monitor is inoperable beyond the current requirements. Therefore, the consequences of an accident previously evaluated are not significantly increased.

Administrative/Editorial/Miscellaneous Changes Enhancements and administrative changes proposed for TSs affected by the previously discussed updated FHA or changes associated with increasing consistency with the ITS are unrelated to any accident initiator. Administrative changes likewise cannot impact the consequences of any accident previously evaluated.

The following is a listing of other changes proposed in this amendment request which modify the TSs (not considered within the editorial/administrative realm).

Enclosure to 2CAN081903 Page 30 of 37 A new Note 3 is proposed that clarifies the original intent of the TS requirements for radiation monitoring and automatic isolation of the Containment Purge system. As written, the TS would require the radiation monitoring and isolation capability to remain operable even when the Containment Purge system is secured. The addition of Note 3 specifies that operability is required only during 1) Containment Purge operations, or 2) ongoing Containment Building continuous ventilation operations when moving recently irradiated fuel assemblies or moving new fuel assemblies over irradiated fuel assemblies in the Containment Building, consistent with the updated FHA and TSTF-51. Other associated enhancements are made to the Containment Purge requirements in support of the above changes or to provide additional clarification.

The phrase "elevation corresponding to the" top of irradiated fuel is added to the Limiting Condition for Operation (LCO) of TS 3.9.9, "Water Level - Reactor Vessel."

This ensures that proper water level is established prior to initiating refueling of the reactor core following a defueled condition.

The movement of fuel "within the reactor vessel" contained in the Applicability and Action of TS 3.9.9 is revised to "within the Containment Building." This reference is also added to the Surveillance Requirement. The required water level should be met even when fuel is being moved in other areas of the refueling canal, not just in the reactor vessel. In addition, the phrase "while in Mode 6" is deleted from the Applicability since fuel assemblies cannot physically be removed from the reactor until Mode 6 has been achieved.

Enhancements associated with the Containment Purge system radiation instrumentation ensure Surveillance testing is performed when the system is in service, regardless if an actual Purge is taking place. In addition, the proposed changes ensure appropriate testing is performed prior to placing the system in service each refueling outage. The proposed changes are neutral or more restrictive and, therefore, cannot increase the consequences of an accident previously evaluated.

Clarifications to limitations on refueling water level and the location of fuel assemblies are more restrictive changes, ensuring proper controls have been established before activities are commenced. No impact to the consequences of any accident result from these changes. The changes to these TSs, in addition to the aforementioned changes to Containment Purge requirements, do not increase the probability of an accident occurring.

Based on the above, the proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.

Enclosure to 2CAN081903 Page 31 of 37

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

Response: No Updated FHA TS changes associated with the updated FHA involve no physical changes to the plant.

These changes act to ensure required structures, systems, and components (SSCs) are operable when moving irradiated fuel assemblies or new fuel assemblies over irradiated fuel assemblies to limit any Control Room or offsite dose consequences to within acceptable limits. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

TSTF-51 and TSTF 471 TS changes associated with ITS improvements related to these TSTFs involve no physical changes to the plant. The removal of references to "core alterations" in favor of restrictions associated with the movement of fuel assemblies eliminates current restrictions associated with the manipulation of other core components (i.e., sources or reactivity control components within the core). Such manipulations cannot result in an FHA, boron dilution event, or loss of SDC. In addition, such manipulations cannot result in an appreciable change in core reactivity due to the high RCS boron concentration required during refueling operations by the TSs. TSTF-51 changes associated with a reduction in restrictions following sufficient radioactive decay of fuel assemblies are not considered accident precursors. The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

TSTF-272 Changes associated with TSTF-272 place additional restrictions on Mode 6 operations by ensuring the boron concentration of the water in the refueling canal meets the same TS limits required for the RCS when the RCS is in direct hydraulic communication with the refueling canal (i.e., reactor vessel head removed and refueling canal filled). These changes are unrelated to any accident initiator and further prohibit any challenge to the fuel in the reactor vessel by ensure sufficient boron concentration is maintained during Mode 6 operations. The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

TSTF-286 Changes associated with TSTF-286 permit operator control of RCS inventory and temperature when certain TS requirements are not met, provide the overall required SDM of the RCS is maintained. No physical plant changes are related to these TS changes.

Enclosure to 2CAN081903 Page 32 of 37 The only accident or event that could be affected by this change is the boron dilution event, which has been previously evaluated. The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism.

Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

TSTF-571-T The proposed change revises the Actions for inoperable source range neutron flux monitors to prohibit the movement of fuel assemblies, sources, and reactivity control components when a monitor is inoperable. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). No credible new failure mechanisms, malfunctions, or accident initiators that would have been considered a design basis accident in the ANO-2 Safety Analysis Report (SAR) are created.

Administrative/Editorial/Miscellaneous Changes Enhancements and administrative changes proposed for TSs affected by the above updated FHA or ITS improvements are unrelated to any accident initiator and involve no physical changes to the plant.

Enhancements associated with the Containment Purge system radiation instrumentation ensure Surveillance testing is performed when the system is in service, regardless if an actual Purge is taking place. In addition, the proposed changes ensure appropriate testing is performed prior to placing the system in service each refueling outage. Clarifications to limitations on refueling water level and the location of fuel assemblies are more restrictive changes, ensuring proper controls have been established before activities are commenced.

The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism. Based on the above, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No Updated FHA TS changes associated with the updated FHA act to ensure required SSCs are operable when moving irradiated fuel assemblies or new fuel assemblies over irradiated fuel assemblies to limit any Control Room or offsite dose consequences to within acceptable limits. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Enclosure to 2CAN081903 Page 33 of 37 TSTF-51 and TSTF 471 The removal of references to "core alterations" in favor of restrictions associated with the movement of fuel assemblies eliminates current restrictions associated with the manipulation of other core components (i.e., sources or reactivity control components within the core). Such manipulations cannot result in an FHA, boron dilution event, or loss of SDC. In addition, such manipulations cannot result in an appreciable change in core reactivity due to the high RCS boron concentration required during refueling operations by the TSs. TSTF-51 also reduces restrictions following sufficient radioactive decay of fuel assemblies since the consequence of an FHA following this decay period would remain within 10 CFR 50.67 limits. Note that this allowance is not adopted for Control Room ventilation or radiation monitoring systems (governed under GDC 19). Changes associated with TSTF-51 and TSTF-471, as adopted, do not modify limitations in such a way that the consequences of an FHA would be greater than that assumed in the FHA analysis (i.e.,

10 CFR 50.67 and GDC 19 limitations are not exceeded following an FHA). Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

TSTF-272 Changes associated with TSTF-272 place additional restrictions on Mode 6 operations by ensuring the boron concentration of the water in the refueling canal meets the same TS limits required for the RCS when the RCS is in direct hydraulic communication with the refueling canal (i.e., reactor vessel head removed and refueling canal filled). These changes are more restrictive than the current TS and, therefore, do not involve a significant reduction in a margin of safety.

TSTF-286 Changes associated with TSTF-286 permit operator control of RCS inventory and temperature when certain TS requirements are not met, provide the overall required SDM of the RCS is maintained. The only accident or event that could be affected by this change is the boron dilution event which has been previously evaluated. While the margin between existing boron concentration and that required to meet SDM requirements may be reduced, margin is gained by permitting operators to take corrective action to maintain RCS inventory and temperature within limits during periods when such operations are otherwise prohibited. While not quantifiable, the changes associated with TSTF-286 have a general balanced effect in relation to the margin of safety. Because an unexpected boron dilution event provides sufficient opportunity for detection and recovery, the proposed changes associated with TSTF-286 do not involve a significant reduction in a margin of safety.

TSTF-571-T The proposed change revises the Actions for inoperable source range neutron flux monitors to prohibit the movement of fuel assemblies, sources, and reactivity control components when a monitor is inoperable. No safety limits are affected. No Limiting Conditions for Operation or Surveillance limits are affected. The design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards

Enclosure to 2CAN081903 Page 34 of 37 (or alternatives approved for use by the NRC) continue to be met as described in the plants' licensing basis. The proposed change does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Administrative/Editorial/Miscellaneous Changes Enhancements and administrative changes proposed for TSs affected by the above updated FHA or ITS improvements are unrelated to any accident initiator or mitigation strategy. Enhancements associated with the Containment Purge system radiation instrumentation ensure Surveillance testing is performed when the system is in service, regardless if an actual Purge is taking place. In addition, the proposed changes ensure appropriate testing is performed prior to placing the system in service each refueling outage. Clarifications to limitations on refueling water level and the location of fuel assemblies are more restrictive changes, ensuring proper controls have been established before activities are commenced. Based on the above, these proposed changes do not involve a significant reduction in a margin of safety.

Therefore, the proposed changes contained within this amendment request do not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above, Entergy concludes that the requested change involves no significant hazards consideration, as set forth in 10 CFR 50.92(c), "Issuance of Amendment."

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and 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.

Enclosure to 2CAN081903 Page 35 of 37

6.0 REFERENCES

1. U. S. Nuclear Regulatory Commission letter to Entergy Operations, Inc., "Summary of October 5, 2011, Meeting with Entergy Operations, Inc., on Pre-submittal License Amendment Request for Changes due to Revised Fuel Handling Accident Analysis along with Adoption of TSTF-51, TSF-272, TSTF-286, TSTF-471, and Relocation/Deletion of Non-Improved Technical Specification (ITS) Shutdown Technical Specifications" (2CNA111101) (TAC No. ME6887), dated November 1, 2011

2. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Technical Specification (TS) Change Related to Revised Fuel Assembly Drop Analysis and Adoption of TSTF-51, TSTF-272, TSTF-286, and TSTF-471" (2CAN041201) (ML12096A022), dated April 4, 2012

3. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Supplemental - Technical Specification (TS) Change Related to Revised Fuel Assembly Drop Analysis and Adoption of TSTF-51, TSTF-272, TSTF-286, and TSTF-471" (2CAN071201) (ML12192A089), dated July 9, 2012

4. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Supplemental - Technical Specification (TS) Change Related to Revised Fuel Assembly Drop Analysis and Adoption of TSTF-51, TSTF-272, TSTF-286, and TSTF-471" (2CAN061302) (ML13170A197), dated June 18, 2013

5. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Supplemental - Technical Specification (TS) Change Related to Revised Fuel Assembly Drop Analysis and Adoption of TSTF-51, TSTF-272, TSTF-286, and TSTF-471" (2CAN071301) (ML13183A124), dated July 1, 2013

6. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Withdrawal - Technical Specification (TS) Change Related to Revised Fuel Assembly Drop Analysis and Adoption of TSTF-51, TSTF-272, TSTF-286, and TSTF-471" (2CAN041401) (ML14113A604), dated April 23, 2014

7. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Technical Specification Changes and Analyses Relating to Use of Alternative Source Term" (2CAN031001) (ML100910241), dated March 31, 2010

8. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request - Technical Specification Changes and Analyses Relating to Use of Alternative Source Term - Supplemental Information" (2CAN061004) (ML102000199),

dated June 23, 2010

9. U. S. Nuclear Regulatory Commission letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment RE: Use of Alternate Source Term" (TAC No. ME 3678) (2CNA041102) (ML110980197), dated April 26, 2011

Enclosure to 2CAN081903 Page 36 of 37

10. U. S. Nuclear Regulatory Commission letter to Technical Specification Task Force (TSTF),

"Potential Issues with Plant-Specific Adoption of Travelers TSTF-51, Revision 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations,"

TSTF-286, Revision 2, "Operations Involving Positive Reactivity Additions," and TSTF-471, Revision 1, "Eliminate Use of Term Core Alterations in Actions and Notes""

(ML13246A358), dated November 7, 2013

11. U. S. Nuclear Regulatory Commission letter to Technical Specification Task Force, "Plant-Specific Adoption of Travelers TSTF-51, Revision 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," TSTF-471, Revision 1, "Eliminate Use of Term Core Alterations in Actions and Notes," and TSTF-286, Revision 2, "Operations Involving Positive Reactivity Additions"" (ML17346A587), dated October 4, 2018

12. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "NRC Bulletin 96-02 30 Day Response" (0CAN059606) (ML9605200224), dated May 17, 1996

13. U. S. Nuclear Regulatory Commission letter to Entergy Operations, Inc., "Completion of Licensing Action for NRC Bulletin 96-02, Movement of Heavy Loads over Spent Fuel, over Fuel in the Reactor Core, or over Safety-Related Equipment, dated April 11, 1996, for Arkansas Nuclear One, Units 1 and 2" (0CNA049826), dated April 27, 1998

14. U. S. Nuclear Regulatory Commission letter to Nuclear Energy Institute, Approval of TSTFs

-017, R.2; 036, R.4; 037, R.2; 051, R.2; 348; 350; and -351 (ML993190284), dated November 1, 1999

15. U. S. Nuclear Regulatory Commission letter to Nuclear Energy Institute, Approval of TSTFs

-263, R.3, -265, R.2, -272, R.1, -275, R.2, -295, and -306 (ML993630256), dated December 21, 1999

16. U. S. Nuclear Regulatory Commission letter to Nuclear Energy Institute, Approval of TSTFs

- 017, R.2; - 036, R.4; - 037, R.2; - 051, R.2; - 052, R.3; - 125, R.1; - 207, R.5; - 212, R.1 -

239; - 276, R.2; - 283, R.3; - 286, R.2; - 306, R.2; - 308, R.1; - 324, R.1; - 339, R.2; - 340, R.3; - 348; - 349, R.1; - 350; - 351; - 352, R.1; - 353; - 361, R.1; - 362; - 363; - 364 and - 365 (ML003730788), dated July 6, 2000

17. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request to Revise the Technical Specifications Based Upon Revised Fuel Handling Accident Analysis - Waterford Steam Electric Station Unit 3" (ML11105A131),

dated April 13, 2011

18. U. S. Nuclear Regulatory Commission letter to Waterford Steam Electric Station, Unit 3, "Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re: Request to Revise the Technical Specifications based upon a Revised Fuel Handling Accident Analysis" (TAC No. ME6049) (ML120940171), dated April 25, 2012

19. U. S. Nuclear Regulatory Commission letter to Tennessee Valley Authority, "Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment to Incorporate Part of TSTF-51, Revision 2, into the Technical Specifications to Eliminate Certain ESF Operability Requirements During Core Alterations" (TAC No. MB2005) (ML020100062), dated January 22, 2002

Enclosure to 2CAN081903 Page 37 of 37

20. U. S. Nuclear Regulatory Commission letter to Dominion Nuclear Connecticut, Inc.,

"Millstone Nuclear Power Station, Unit No. 2 - Issuance of Amendment RE: Changes to Technical Specification Definitions for Core Alteration and Refueling Operations" (TAC No. MB1779) (ML013440338), dated January 11, 2002

21. U. S. Nuclear Regulatory Commission letter to Ginna Nuclear Power Plant, LLC, "Ginna Nuclear Power Plant - Amendment RE: Technical Specification Task Force (TSTF)-286, "Operations Involving Positive Reactivity Additions"" (TAC No. ME5444) (ML110980569),

dated June 8, 2011

22. U. S. Nuclear Regulatory Commission letter to Dominion Nuclear Connecticut, Inc.,

"Millstone Power Station, Unit Nos. 2 and 3 - Issuance of Amendments RE: Technical Specification Change Request for Positive Reactivity Additions During Shutdown Operations" (TAC Nos. MC6379 and MC6380) (ML060550271), dated June 28, 2006

23. U. S. Nuclear Regulatory Commission letter to Nuclear Management Company, LLC, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendment RE: Core Alterations" (ML063450073), dated February 15, 2007

24. U. S. Nuclear Regulatory Commission letter to Calvert Cliffs Nuclear Power Plant, Inc.,

"Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Amendment Re: Deletion of Term Core Alterations" (TAC Nos. MC7330 and MC7331) (ML062350447), dated September 21, 2006

25. Entergy Operations, Inc. letter to U. S. Nuclear Regulatory Commission, "License Amendment Request to Revise Technical Specifications to Adopt TSTF 563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program" (2CAN041905) (ML19120A086), dated April 30, 2019 ATTACHMENTS

1. Technical Specification Page Markups

2. Technical Specification Bases Page Markups (Information Only)

3. Retyped Technical Specification Pages

4. List of Regulatory Commitments

Enclosure, Attachment 1 to 2CAN081903 Technical Specification Page Markups (20 pages)

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c. Digital computer channels - The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed 471 Pg 1.1-2 and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control element assemblies are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except controlled leakage) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

ARKANSAS - UNIT 2 1-3 Amendment No. 157,220,255,266,

REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the Rreactor Ccoolant Ssystem (RCS) shall be 2000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.

E/

APPLICABILITY: ALL MODES. E/ DOD 9 DOD 11 ACTION:

With the flow rate of reactor coolant through the RCSreactor coolant system < 2000 gpm, immediately suspend all operations that would cause introduction of coolant into the RCS 286 / E /

withinvolving a reduction in boron concentration less than required to meet the SDM or boron DOD 2 concentration of LCO 3.1.1.1, LCO 3.1.1.2, or LCO 3.9.1, as applicableof the Reactor Coolant System.

C/

DOD 2 SURVEILLANCE REQUIREMENTS 4.1.1.3 The flow rate of reactor coolant through the RCSreactor coolant system shall be determined to be 2000 gpm within one hour prior to the start of and in accordance with the Surveillance Frequency Control Program during a reduction in the Reactor Coolant System boron concentration by either:

a. Verifying at least one reactor coolant pump is in operation, or

b. Verifying that at least one low pressure safety injection pump or containment spray pump is in operation as a shutdown cooling pump and supplying 2000 gpm through the RCSreactor coolant system.

ARKANSAS - UNIT 2 3/4 1-4 Amendment No. 126,255,315,

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1. AREA MONITORS

a. Spent Fuel Pool Area Monitor 1 Note 1 1.5 x 10-2 R/hr 10 101 R/hr 13

b. Containment High Range 2 1, 2, 3, & 4 Not Applicable 1 - 107 R/hr 18

2. PROCESS MONITORS

a. Containment Purge and Exhaust Isolation 1 5 & 6Note 3 2 x background 10 - 106 cpm 16

b. Control Room Ventilation Intake Duct Monitors 2 Note 2 2 x background 10 - 106 cpm 17,20,21

c. Main Steam Line Radiation Monitors 1/Steam Line 1, 2, 3, & 4 Not Applicable 10 104 mR/hr 19 Note 1 - With fuel in the spent fuel pool or building.

FHA / 51 Note 2 - MODES 1, 2, 3, 4, and during movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over Pg 3.3-39 irradiated fuel assemblies.

& 40 /

Note 3 - Applicable during: C/ DOD 1

a. PURGE of the Containment Building or, DOD 10

b. Containment Building continuous ventilation operations when moving recently irradiated fuel assemblies or moving new FHA fuel assemblies over recently irradiated fuel assemblies in the Containment Building.

ARKANSAS - UNIT 2 3/4 3-25 Amendment No. 63,130,145,206,231,255, 51 Pg 3.3-35/36

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 13 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 16 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, complete the following:

51 Pg 3.3-35/36

a. If performing CORE ALTERATIONS or moving recently irradiated fuel FHA assemblies or moving new fuel assemblies over recently irradiated fuel assemblies within the Containmentreactor Bbuilding, secure the E/ Ccontainment Ppurge Ssystem or suspend CORE ALTERATIONS and the DOD 10 movement of recently irradiated fuel assemblies and movement of new fuel FHA assemblies over recently irradiated fuel assemblies within the Containmentreactor Bbuilding.

b. If a Ccontainment PURGE is in progress, secure the Ccontainment Ppurge Ssystem.

c. If continuously ventilating the Containment Building, verify the associated SPING monitor operable or perform the applicable ACTION(s)S of the Offsite Dose Calculation Manual, Appendix 2, Table 2.2-1;, otherwiser, secure the Ccontainment Ppurge Ssystem. C/

DOD 10 ACTION 17 - In MODE 1, 2, 3, or 4, with no channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the control room emergency ventilation system (CREVS) in the recirculation mode of operation or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

ACTION 18 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, (1) either restore the inoperable channel to OPERABLE status within 7 days or (2) prepare and submit a Special Report to the NRC within 30 days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status. With both channels inoperable, initiate alternate methods of monitoring the containment radiation level within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in addition to the actions described above.

ACTION 19 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the NRC within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ARKANSAS - UNIT 2 3/4 3-26 Amendment No. 63,130,145,206,231, 255,301,

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 20 - In MODE 1, 2, 3, or 4 with the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, within 7 days restore the inoperable channel to OPERABLE status or initiate and maintain the CREVS in the recirculation mode of operation. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

ACTION 21 - During movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies with one or two channels inoperable, FHA /

immediately place one OPERABLE CREVS train in the emergency recirculation DOD 1 mode or immediately suspend the movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies.

ARKANSAS - UNIT 2 3/4 3-26a Amendment No. 301,

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE INSTRUMENT CHECK CALIBRATION FUNCTIONAL TEST REQUIRED

1. AREA MONITORS

a. Spent Fuel Pool Area Monitor SFCP SFCP SFCP Note 1

b. Containment High Range SFCP SFCP Note 4 SFCP 1, 2, 3, & 4

2. PROCESS MONITORS

a. Containment Purge and In accordance with Exhaust Isolation Note 2 SFCPNote 3 Note 3 applicable Notes5

&6

b. Control Room Ventilation Note 5 SFCP SFCP SFCP Note 6 C/

Intake Duct Monitors DOD 10

c. Main Steam Line SFCP SFCP SFCP 1, 2, 3, & 4 Radiation Monitors E/

Note 1 - With fuel in the spent fuel pool or building. DOD 10 Note 2 - Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to initiating Ccontainment PURGEpurge operations and in accordance with the Surveillance Frequency Control Program during Ccontainment PURGEpurge or continuous ventilation operations.

Note 3 - Within 31 days prior to initiating Ccontainment PURGEpurge operations and in accordance with the Surveillance Frequency Control Program during Ccontainment continuous ventilationpurge operations when moving recently irradiated fuel FHA assemblies or moving new fuel assemblies over recently irradiated fuel assemblies in the Containment Building.

Note 4 - Acceptable criteria for calibration are provided in Table II.F.1-3 of NUREG-0737. 51 Note 5 - MODES 1, 2, 3, 4, and during movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over FHA /

irradiated fuel assemblies. DOD 1 Note 6 - When the Control Room Ventilation Intake Duct Monitor is placed in an inoperable status solely for performance of this Surveillance, entry into associated ACTIONS may be delayed up to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

ARKANSAS - UNIT 2 3/4 3-27 Amendment No. 63,130,145,206,231,255,315, Next Page is 3/4 3-36

THIS PAGE INTENTIONALLY LEFT BLANK E/

(Next page is 3/4 3-36) DOD 10 ARKANSAS - UNIT 2 3/4 3-28 Amendment No. 53,134,151,163,191

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. The reactor coolant loops listed below shall be in OPERABLEoperable:

1. Reactor Coolant Loop (A) and at least one associated reactor coolant pump.

2. Reactor Coolant Loop (B) and at least one associated reactor coolant pump. ITS Pg 3.4.5-1

b. At least one of the above Reactor Coolant Loops shall be in operation.* /E/

DOD 11 APPLICABILITY: MODE 3.

ACTION:

a. With less than the above required reactor coolant loops OPERABLEoperable, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant loop in operation, suspend all operations that would causeinvolving introduction of coolant into the RCS witha reduction in boron 286 concentration less than required to meet SDM of LCO 3.1.1.1of the Reactor Pg Coolant System and immediately initiate corrective action to return the required 3.4-9 loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction of coolant into the RCSdilution of the reactor 286 coolant withsystem boron concentration less than required to meet SDM of LCO 3.1.1.1, and Pg (2) core outlet temperature is maintained at least 10 °F below saturation temperature.

3.4-8 ARKANSAS - UNIT 2 3/4 4-2 Amendment No. 24,29,315,

REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop (A) and its associated steam generator and at least one associated reactor coolant pump.

2. Reactor Coolant Loop (B) and its associated steam generator and at least one associated reactor coolant pump.

3. Shutdown Cooling Loop (A) #.

4. Shutdown Cooling Loop (B) #.

b. At least one of the above coolant loops shall be in operation.*

ITS Pg 3.4.6-2 APPLICABILITY: Modes 4 and 5. / E / DOD 11 ACTION:

a. With less than the above required coolant loops OPERABLE, immediately initiate status as soon as possible and initiate action to make at least one steam generator available for decay heat removal via natural circulation. LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

b. With no coolant loop in operation, suspend all operations that would 286 causeinvolving a reduction introduction of coolant into the RCS with boron Pg concentration less than required to meet SDM of LCO 3.1.1.1 or LCO 3.1.1.2, as 3.4-11 applicable,of the Reactor Coolant System and immediately initiate corrective & 14 /

action to return the required coolant loop to operation. C/

DOD 3 SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required shutdown cooling loop(s) shall be determined OPERABLE per the INSERVICE TESTING PROGRAM.

4.4.1.3.2 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be 23% indicated level in accordance with the Surveillance Frequency Control Program.

4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

All reactor coolant pumps and decay heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction of coolant 286 Pg intodilution of the RCSreactor coolant system with boron concentration less than required to 3.4-10 meet SDM of LCO 3.1.1.1 or LCO 3.1.1.2, as applicable, and (2) core outlet temperature is & 13 / C maintained at least 10°F below saturation temperature. / DOD 3

The normal or emergency power source may be inoperable in Mode 5.

ARKANSAS - UNIT 2 3/4 4-2a Amendment No. 24,29,233,301,305, 315,

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION AND AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two independent control room emergency ventilation and air conditioning systems shall be OPERABLE. (Note 1)

APPLICABILITY: MODES 1, 2, 3, 4, or during movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel FHA /

assemblies. DOD 1 ACTION:

MODES 1, 2, 3, and 4

a. With one control room emergency air conditioning system (CREACS) inoperable, restore the inoperable system to OPERABLE status within 30 days.

b. With one control room emergency ventilation system (CREVS) inoperable for reasons other than ACTION d, restore the inoperable system to OPERABLE status within 7 days.

c. With one CREVS inoperable for reasons other than ACTION d and one CREACS inoperable, restore the inoperable CREVS to OPERABLE status within 7 days and restore the inoperable CREACS to OPERABLE status within 30 days.

d. With one or more CREVS inoperable due to an inoperable CRE boundary:

1. Immediately initiate action to implement mitigating actions, and

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

3. Restore the CRE boundary to OPERABLE status within 90 days

e. With two CREVS inoperable for reasons other than ACTION d (Note 2):

1. Immediately initiate action to implement mitigating actions, and

2. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , verify LCO 3.4.8, "Specific Activity," is met, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore at least one CREVS to OPERABLE status.

f. With two CREACS inoperable (Note 2), restore at least one CREACS to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With ACTIONS a, b, c, d, e, and/or f not met, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

Note 1: The control room envelope (CRE) boundary may be open intermittently under administrative controls.

Note 2: ACTION e is not applicable if the second CREVS is intentionally made inoperable.

ACTION f is not applicable if the second CREACS is intentionally made inoperable.

ARKANSAS - UNIT 2 3/4 7-17 Amendment No. 206,219,255,288, 301,304,

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION AND AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION FHA /

During MovementHandling of Irradiated Fuel Assemblies or Movement of New Fuel DOD 1 Assemblies over Irradiated Fuel Assemblies

g. With one CREACS inoperable, restore the inoperable system to OPERABLE status within 30 days or immediately place the OPERABLE system in operation; otherwise, suspend all activities involving the movementhandling of irradiated fuel assemblies or movement of FHA /

new fuel assemblies over irradiated fuel assemblies. DOD 1

h. With one CREVS inoperable, restore the inoperable system to OPERABLE status within 7 days or immediately place the control room in the emergency recirc mode of operation; otherwise, suspend all activities involving the movementhandling of irradiated fuel FHA /

assemblies or movement of new fuel assemblies over irradiated fuel assemblies.

DOD 1

i. With one CREVS inoperable for reasons other than ACTION d and one CREACS inoperable:

1. restore the inoperable CREVS to OPERABLE status within 7 days or immediately place the CRE in the emergency recirc mode of operation, and

2. restore the inoperable CREACS to OPERABLE status within 30 days or immediately place the OPERABLE system in operation;

3. otherwise, suspend all activities involving the movementhandling of irradiated fuel FHA /

assemblies or movement of new fuel assemblies over irradiated fuel assemblies. DOD 1

j. With both CREACS inoperable, immediately suspend all activities involving the FHA /

movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over DOD 1 irradiated fuel assemblies.

k. With both CREVS inoperable or with one or more CREVS inoperable due to an inoperable CRE boundary, immediately suspend all activities involving the movementhandling of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel FHA /

assemblies. DOD 1 ARKANSAS - UNIT 2 3/4 7-17a Amendment No. 206,219,255,281, 288,304,

ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

b. One diesel generator with:

1. A day fuel tank containing a minimum volume of 300 gallons of fuel,

2. A fuel storage system, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6, or during movement of recently irradiated fuel assemblies or movement of new fuel assemblies over recently irradiated fuel FHA assemblies.

ITS Pg 3.8.2-2 / DOD 11 51 ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving the movement of recently irradiated fuel FHA assembliesCORE ALTERATIONS, the movement of new fuel assemblies over recently irradiated fuel assemblies, orand operations involving positive reactivity additions that could result in loss of required SDM or boron concentrationchanges. 286 Pg 3.8-20 471 Pg 3.8.2-2 SURVEILLANCE REQUIREMENT 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for Requirement 4.8.1.1.2a.5.

ARKANSAS - UNIT 2 3/4 8-5 Amendment No. 149,255,

ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE:

1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Load Center Bus 4 - 480 volt Motor Control Center Busses 2 - 120 volt A.C. Vital Busses APPLICABILITY: MODES 5 and 6, or during movement of recently irradiated fuel assemblies or movement of new fuel assemblies over recently irradiated fuel FHA 471 Pg 3.8.10-1 assemblies.

ACTION: 51 Pg 3.8-19, 20, & 41 ITS 3.8.10-1

/ C / DOD 12 With less than the above complement of A.C. busses OPERABLE and energized, immediately suspend core alterations, the movement of recently irradiated fuel assemblies, the movement of FHA new fuel assemblies over recently irradiated fuel assemblies, and any operations involving positive reactivity additions that could result in loss of required SDM or boron concentration. 286 Pg 3.8-41 SURVEILLANCE REQUIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.

ARKANSAS - UNIT 2 3/4 8-7 Amendment No. 227,

ELECTRICAL POWER SYSTEMS DC SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following DC electrical equipment and bus shall be energized and OPERABLE:

1 - 125-volt DC bus, and 1 - 125-volt battery bank and charger supplying the above DC bus.

APPLICABILITY: MODES 5 and 6, or during movement of recently irradiated fuel assemblies FHA or movement of new fuel assemblies over recently irradiated fuel assemblies.

ACTION: 51 Pg 3.8-29 & 41

a. With the required battery charger inoperable:

i. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and 471 Pg 3.8.5-2 ii. Verify battery float current 2 amps once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the requirements of ACTION a not met or with the above complement of DC equipment and bus otherwise inoperable, immediately suspend core alterations, the movement of recently irradiated fuel assemblies, the movement of new fuel FHA assemblies over recently irradiated fuel assemblies, and any operations involving positive reactivity additions that could result in loss of required SDM or boron concentration. 286 Pg 3.8-29 SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. bus shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 125-volt battery bank and charger shall be demonstrated OPERABLE per Surveillance Requirements 4.8.2.3.1, 4.8.2.3.2, and 4.8.2.3.3; however, while each of these Surveillance Requirements must be met, Surveillance Requirements 4.8.2.3.2 and 4.8.2.3.3 are not required to be performed.

ARKANSAS - UNIT 2 3/4 8-10 Amendment No. 94,227,297,

3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, Tthe boron concentration of the reactor coolant system and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met:

ITS Pg

a. Either a Keff of 0.95 or less, which includes a 1% k/k conservative allowance for 3.9.1-1 / E uncertainties, or / DOD 13

b. A boron concentration of 2500 ppm, which includes a 50 ppm conservative allowance for uncertainties.

APPLICABILITY: MODE 6*.

471 Pg 3.9.1-1 ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at 40 gpm of 2500 ppm boric acid solution until Keff is reduced to 0.95 or the boron concentration is restored to 2500 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9.1.2 The boron concentration of the reactor coolant and the refueling canal shall be determined by chemical analysis in accordance with the Surveillance Frequency Control Program.

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removedOnly applicable to the refueling canal when connected to the RCS. 272 Pg 3.9-1 /

DOD 4 ARKANSAS - UNIT 2 3/4 9-1 Amendment No. 82,169,255,315, Correction Letter dated 10/24/95,

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6. ITS Pg 3.9.2-1

/ DOD 5 ACTION:

a. With one or more of the above required monitors inoperable, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity additionschanges.

471 Pg 3.9.2-1 571 Pg AND 3.9.2-1 /

Suspend movement of fuel, sources, and reactivity control components within the DOD 5 reactor vessel.1

b. With both of the above required monitors inoperable, determine the boron concentration of the reactor coolant system at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK in accordance with the Surveillance Frequency Control Program,

b. A CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program, and

c. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of the movement of recently irradiated fuel assemblies or the movement of new fuel FHA /

assemblies over recently irradiated fuel assembliesCORE ALTERATIONS. DOD 5 51 471 Pg 3.9.2-1 Note 1: Fuel assemblies, sources, and reactivity control components may be moved if 571 Pg necessary to restore an inoperable source range neutron flux monitor or to complete 3.9.2-1 movement of a component to a safe condition.

ARKANSAS - UNIT 2 3/4 9-2 Amendment No. 315,

REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment door is capable* of being closed,

b. A minimum of one door in each airlock is capable* of being closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed* by a manual or automatic isolation valve, blind flange, or equivalent, or

2. Capable* of being closed by an OPERABLE containment purge and exhaust isolation system.

APPLICABILITY: During CORE ALTERATIONS or movement of recently irradiated fuel assemblies or movement of new fuel assemblies over recently irradiated FHA 51 Pg fuel assemblies within the Ccontainment Building.

3.9.4 /

DOD 6 ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of recently irradiated fuel assemblies or movement of new fuel assemblies over recently irradiated fuel assemblies in the FHA Ccontainment Building. The provisions of Specification 3.0.3 are not applicable.

E/

DOD 14 SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment penetrations shall be determined to be in its above required conditions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS or movement of recently irradiated fuel assemblies or movement of new fuel FHA assemblies over recently irradiated fuel assemblies in the Ccontainment Building.

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls. Administrative controls shall ensure that appropriate personnel are aware that when containment penetrations, including both personnel airlock doors and/or the equipment door are open, a specific individual(s) is designated and available to close the penetration following a required evacuation of containment, and any obstruction(s) (e.g., cables and hoses) that could prevent closure of an airlock door and/or the equipment door be capable of being quickly removed.

ARKANSAS - UNIT 2 3/4 9-4 Amendment No. 166,203,230,315, Next page is 3/4 9-6

REFUELING OPERATIONS COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY: During CORE ALTERATIONS movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies in the FHA 51 / reactor pressure vessel.

DOD 7 ACTION:

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies in the reactor pressure vesselall CORE FHA ALTERATIONS. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and in accordance with the Surveillance Frequency Control Program during movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies in FHA the reactor pressure vesselCORE ALTERATIONS.

ARKANSAS - UNIT 2 3/4 9-6 Amendment No. 315,

REFUELING OPERATIONS SHUTDOWN COOLING AND COOLANT CIRCULATION SHUTDOWN COOLING - ONE LOOP LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one shutdown cooling loop shall be in operation.

APPLICABILITY: MODE 6. E/

DOD 11 ACTION:

a. With less than one shutdown cooling loop in operation, except as provided in b.

below, suspend all operations involving an increase in the reactor decay heat load or 286 that would cause introduction of coolant into the RCS witha reduction in boron Pg concentration less than required to meet the boron concentration of LCO 3.9.1of the 3.9-6/ Reactor Coolant System. Close all containment penetrations providing direct access DOD 9 from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The shutdown cooling loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided no operations are permitted that would cause introduction of 286 coolant into the RCS with boron concentration less than that required to meet the Pg minimum required boron concentration of LCO 3.9.1during the performance of 3.9-6 CORE ALTERATIONS.

51 /

c. The provisions of Specification 3.0.3 are not applicable. DOD 8 SURVEILLANCE REQUIREMENTS 4.9.8.1 A shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flow rate of 2000 gpm in accordance with the Surveillance Frequency Control Program.

ARKANSAS - UNIT 2 3/4 9-9 Amendment No. 29,104,315,

REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.9 At least 23 feet of water shall be maintained over the elevation corresponding to the top of irradiated fuel assemblies seated within the reactor pressure vessel.

APPLICABILITY: During movement of fuel assemblies or CEAs within the Containment 51 Pg Buildingthe reactor pressure vessel while in MODE 6, except during 3.9-10 /

latching and unlatching of CEAs. DOD 8 ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or CEAs within the Containment Buildingthe pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.9 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program thereafter during movement of fuel assemblies within the C/

Containment Buildingor CEAs. DOD 15 ARKANSAS - UNIT 2 3/4 9-10 Amendment No. 167,315,

Enclosure, Attachment 2 to 2CAN081903 Technical Specification Bases Page Markups (Information Only)

(11 pages)

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 2000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2000 GPM will circulate an equivalent Reactor Coolant System volume of 6,650 cubic feet in approximately 25 minutes. The reactivity change rate associated with boron concentration reductions will therefore be within the capability of operator recognition and control.

The ACTION permits normal plant control operations that individually add limited positive reactivity (e.g., temperature or boron fluctuations associated with RCS inventory management 286 or temperature control), provided such activities are accounted for in the calculated SDM.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle. The Surveillance Requirements consisting of beginning of cycle measurements and end of cycle MTC predictions ensure that the MTC remains within acceptable values. The confirmation that the measured values are within a tolerance of +/- 0.16 X 10-4 k/k/°F from the corresponding design values (MTC predicted values based on core data) prior to 5% power and near 40 EFPD provides assurances that the MTC will be maintained within acceptable values throughout each fuel cycle. CE NPSD 911-A and CE NPSD 911 Amendment 1-A, "Analysis of Moderator Temperature Coefficients in Support of a Change in the Technical Specifications End of Cycle Negative MTC Limits",

provide the analysis that established the design margin of +/- 0.16 X 10-4 k/k/°F. The option to eliminate the EOC MTC measurement requires that the reload analysis and predicted design value be performed using the CE methodology.

For fuel cycles that meet the applicability requirements of WCAP-16011-P-A, Revision 0, "Startup Test Activity Reduction Program," SR 4.1.1.4.2.a may be met prior to exceeding 5% of RATED THERMAL POWER after each fuel loading by confirmation that the predicted MTC, when adjusted for the measured RCS boron concentration, is within the MTC limits.

WCAP-16011-P-A also provides the basis for using only the near 40 EFPD surveillance test result to justify elimination of the near two-thirds of expected core burnup surveillance when applicability requirements are met. Performance of only one measurement at power is justified based on the WCAP-16011-P-A conclusion that ITC startup test data between different operating conditions is poolable. For the purposes of this specification, "within 7 days" ensures the required tests will be performed within 7 days prior to, or within 7 days following the point in core life specified for the test.

The applicability requirements in WCAP-16011-P-A ensure core designs are not significantly different than those used to benchmark predictions and require that the measured RCS boron concentration meets specific test criteria. This provides assurance that the MTC obtained from the adjusted predicted MTC is accurate.

For fuel cycles that do not meet the applicability requirements in WCAP-16011-P-A, the verification of MTC required prior to entering MODE 1 after each fuel loading is performed by measurement of the isothermal temperature coefficient.

ARKANSAS - UNIT 2 B 3/4 1-2 Amendment No. 24,82,157,229 Rev. 16,18,31,53,

INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

The PURGE as defined in the definitions section is a release under a purge permit, whereas continuous ventilation is defined as operation of the purge system after the requirements of the purge permit have been satisfied. When securing the containment purge system to meet the ACTION requirements of this Specification, at least one supply valve and one exhaust valve is to be closed, and the supply and exhaust fans secured. Because SPING 5 is utilized to fulfill offsite dose monitoring for this pathway, the SPING must undergo a CHANNEL FUNCTIONAL TEST within 31 days prior to and a CHANNEL CHECK within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the performance of a PURGE. The ANO ODCM describes required actions associated with an inoperable SPING 5 relevant to PURGE operations, the ventilation mode of operation, and operations during the movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) or movement of new fuel assemblies over FHA recently irradiated fuel assemblies within the Containment Building.

The principal function of the control room intake duct monitors is to provide an enclosed 51 environment from which the unit can be operated following an uncontrolled release of radioactivity. Due to the unique arrangement of the shared control room envelope, one control room isolation channel receives a high radiation signal from the ANO-1 control room ventilation intake duct monitor and the redundant channel receives a high radiation signal from the ANO-2 control room ventilation intake duct monitor. With neither channel of the control room radiation monitoring system operable, the CREVS must be placed in a condition that does not require the isolation to occur (i.e., one operable train of CREVS is placed in the emergency recirculation mode of operation). Reactor operation may continue indefinitely in this state.

Note 6 to Table 4.3-3 allows up to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to perform the monthly CHANNEL FUNCTIONAL TEST of the Control Room ventilation intake duct radiation monitors without declaring the LCO not met. If the test is not completed in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , affected intake duct monitors must be declared inoperable and applicable ACTION(s) of TS Table 3.3-6 applied.

Channels not restored to an OPERABLE status in accordance with Actions 17 or 20, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Remaining within the Applicability of the LCO is acceptable because the plant risk in MODE 4 is similar to or lower than MODE 5 (reference CE NPSD-1186-A, Technical Justification for the Risk Informed Modification to Selected Required Action End States for CEOG PWRs, October, 2001). In MODE 4 there are more accident mitigation systems available and there is more redundancy and diversity in core heat removal mechanisms than in MODE 5. However, voluntary entry into MODE 5 may be made as it is also an acceptable low-risk state.

ARKANSAS - UNIT 2 B 3/4 3-3 Amendment No. 22,29,60,123,132,191 Rev. 8,17,40,56,60,

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above the limits specified by Specification 3.2.4 during all normal operations and anticipated transients.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or shutdown cooling (SDC) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two SDC loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one SDC pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System (RCS). The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

If only one required SDC train is OPERABLE and in operation and no required RCS loops are OPERABLE, redundancy for heat removal is lost and the plant must be placed in a configuration that minimizes overall plant risk. This redundancy is obtained by making at least one SG available for decay heat removal via natural circulation because:

1. MODE 4 operation poses overall lower risk of core damage and large early radiation release than does MODE 5 (reference CE NPSD-1186-A, Technical Justification for the Risk Informed Modification to Selected Required Action End States for CEOG PWRs, October, 2001). This is particularly true with SDC impaired.

2. In MODE 4, RCS and steam generator conditions may be maintained such that failure of the operating SDC train may be mitigated by natural circulation heat removal through one or more steam generators.

Remaining within the Applicability of the LCO is acceptable because the plant risk in MODE 4 is similar to or lower than MODE 5. However, voluntary entry into MODE 5 may be made as it is also an acceptable low-risk state. These Actions are modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 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 4, 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.

286 With no reactor coolant loop in operation, suspending the introduction of coolant into the RCS Insert B4 with boron concentration less than required to meet the minimum SDM of LCO 3.1.1.1 or LCO 3.1.1.2, as applicable, is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core; however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.

ARKANSAS - UNIT 2 B 3/4 4-1 Amendment No. 24,29,149,199 Rev. 1,10,33,46,56,

PLANT SYSTEMS BASES 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION (CREVS) AND AIR CONDITIONING SYSTEM (CREACS) (continued)

APPLICABILITY (continued)

During movement of irradiated fuel assemblies or movement of new fuel assemblies over FHA irradiated fuel assemblies, the CREVS must be OPERABLE to cope with a release due to a fuel handling accident.

Unit 1 and Unit 2 control rooms are a single environment for emergency ventilation and air conditioning concerns. Since the control room emergency ventilation and air conditioning equipment is shared between units, the plant status of both units must be considered when determining applicability of the specification.

ACTIONS a.

With one CREACS train inoperable, action must be taken to restore OPERABLE status within 30 days. In this ACTION, the remaining OPERABLE CREACS train is adequate to maintain the control room temperature within limits. However, the overall reliability is reduced because a failure in the OPERABLE CREACS train could result in a loss of CREACS function. The 30 day ACTION statement is based on the low probability of an event occurring requiring control room isolation, the consideration that the remaining train can provide the required capabilities, and alternate non-safety related cooling means that are available.

b.

With one CREVS train inoperable for reasons other than the loss of capability for automatic actuation on a high radiation signal or for reasons other than an inoperable CRE boundary, action must be taken to restore the OPERABLE status within 7 days. In this ACTION, the remaining OPERABLE CREVS train is adequate to perform the CRE radiation protection function. However, the overall reliability is reduced because a failure in the OPERABLE CREVS train could result in loss of CREVS function. The 7 day ACTION statement is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability. If automatic actuation on high radiation is lost, the ACTIONS of LOC 3.3.3.1 provide sufficient actions to ensure continued safe operation.

c.

With one CREVS train inoperable for reasons other than ACTION d and one CREACS train inoperable, actions must be taken to restore the CREVS to an OPERABLE status within 7 days and to restore the CREACS train to an OPERABLE status within 30 days.

d.

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.

ARKANSAS - UNIT 2 B 3/4 7-9 Rev. 11,41,47,56,63,

PLANT SYSTEMS BASES 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION (CREVS) AND AIR CONDITIONING SYSTEM (CREACS) (continued)

ACTIONS (continued)

e. and f. (continued)

During the period that the CREVS trains are inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from potential hazards while both trains of CREVS are inoperable. In the event of a DBA, the mitigating actions will reduce the consequences of radiological exposures to the CRE occupants. In addition, TS 3.4.8, "RCS Specific Activity," allows limited operation with the RCS activity significantly greater than the LCO limit. This presents a risk to the plant operator during an accident when all CREVS trains are inoperable. Therefore, it must be verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months that LCO 3.4.8 is met. This ACTION does not require additional RCS sampling beyond that normally required by LCO 3.4.8.

With both CREACS trains inoperable, contingency measures may be required to ensure the CRE remains habitable and to limit temperature effects on electrical equipment housed within the Control Room. OP-2104.007 contains additional guidance to aid maintenance of the CRE during this limited 24-hour period.

At least one CREVS train and/or one CREACS train, as applicable, must be returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The AOT is based on WCAP-16125-NP-A, "Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown," Revision 2, August 2010, which demonstrated that the 24-hour AOT is acceptable based on the infrequent use of the ACTION and the small incremental effect on plant risk.

If both CREVS trains and/or CREACS trains, as applicable, remain inoperable beyond the 24-hour AOT in MODE 1, 2, 3, or 4, the unit must be placed in a MODE that minimizes the accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowed outage times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

g. FHA If during movementhandling of irradiated fuel assemblies or during movement of new fuel assemblies over irradiated fuel assemblies, the system cannot be restored within 30 days, then either the OPERABLE CREACS train must be immediately placed in service or all activities involving the movementhandling of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies must be suspended. Placing the OPERABLE CREACS train in service ensures any active failure will be readily detected. The alternative to immediately suspend movement of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies is acceptable since movementhandling of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies could release radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

ARKANSAS - UNIT 2 B 3/4 7-11 Rev. 11,33,41,44,48,56,63,68,

PLANT SYSTEMS BASES 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION (CREVS) AND AIR CONDITIONING SYSTEM (CREACS) (continued)

ACTIONS (continued)

h. FHA If during movementhandling of irradiated fuel assemblies or during movement of new fuel assemblies over irradiated fuel assemblies, the system cannot be restored within 7 days, then either the OPERABLE CREVS train must be immediately placed in emergency recirculation mode or all activities involving the movementhandling of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies must be suspended. Placing the OPERABLE CREVS train in emergency recirculation mode ensures that no failures preventing automatic actuation will occur, and that any active failure will be readily detected.

The alternative to immediately suspend movement of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies is acceptable since movementhandling of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies could release radioactivity that might require isolation of the CRE.

This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

i.

If during movementhandling of irradiated fuel assemblies or during movement of new fuel assemblies over irradiated fuel assemblies, one CREVS train is inoperable for reasons other than ACTION d and one CREACS train are inoperable, actions must be taken to restore the CREVS to an OPERABLE status within 7 days or immediately place the OPERABLE CREVS in the emergency recirculation mode and actions must be taken to restore the CREACS train to an OPERABLE status within 30 days or immediately place the OPERABLE CREACS train in service. If these actions cannot be accomplished, then all activities involving the movementhandling of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies must be suspended. This does not preclude movement of fuel to a safe position.

j.

If during movementhandling of irradiated fuel assemblies or during movement of new fuel assemblies over irradiated fuel assemblies, both CREACS trains are inoperable, actions must be taken immediately to suspend movement of irradiated fuel assemblies or the movement of new fuel assemblies over irradiated fuel assemblies since this is an activity that could release radioactivity that could enter the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude movement of fuel to a safe position.

k.

If during the movementhandling of irradiated fuel assemblies or during movement of new fuel assemblies over irradiated fuel assemblies both CREVS trains are inoperable or with one or more CREVS trains inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

ARKANSAS - UNIT 2 B 3/4 7-12 Amendment No. 132,206 Rev. 1,11,41,47,56,59,63,68,

3/4.8 ELECTRICAL POWER SYSTEMS BASES The OPERABILITY of the AC and DC power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant AC and DC power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.

51 The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated ITS distribution systems during shutdown and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and 2) sufficient 471 instrumentation and control capability is available for monitoring and maintaining the unit status, and 3) mitigating systems will be available following a fuel handling accident. Upon loss of a required power source, suspension of core alterations, the movementhandling of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the FHA previous 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ), the movement of new fuel assemblies over recently irradiated fuel assemblies, and activities that could result in loss of required SDM (Mode 5) or boron 286 concentration (Mode 6)involving positive reactivity additions act to minimize the probability of Insert B6 the occurrence of postulated events. Suspension of these activities shall not preclude placing fuel assemblies in a safe position.

Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that required in 286 the RCS for minimum SDM or refueling boron concentration. This may result in an overall Insert B6 reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.

If the inoperable AC electrical power sources or an inoperable DC electrical power subsystem cannot be restored to an OPERABLE status within the allowable outage times, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Remaining within the Applicability of the LCO is acceptable because the plant risk in MODE 4 is similar to or lower than MODE 5 (reference CE NPSD-1186-A, Technical Justification for the Risk Informed Modification to Selected Required Action End States for CEOG PWRs, October, 2001). In MODE 4 there are more accident mitigation systems available and there is more redundancy and diversity in core heat removal mechanisms than in MODE 5. However, voluntary entry into MODE 5 may be made as it is also an acceptable low-risk state. These Actions are modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 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 4, 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.

ARKANSAS - UNIT 2 B 3/4 8-1 Amendment No. 141,146,215,227,237 Rev. 33,54,56,

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION ITS 471 The limitations on reactivity conditions during REFUELING ensure that: 1) the reactor will remain subcritical during fuel handling operationsCORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.

The Applicability is modified by a Note. The Note states that the limits on boron concentration are only applicable to the refueling canal and the refueling cavity when those volumes are 272 connected to the RCS. When the refueling canal and the refueling cavity are isolated from the Insert B1 RCS, no potential path for boron dilution exists.

Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other 286 operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by the ACTION.

Surveillance Requirement (SR) 4.9.1.2 ensures the coolant boron in the RCS, and connected portions of the refueling canal and the refueling cavity, is within limits. The boron concentration of the coolant in each required volume is determined periodically by chemical analysis. Prior to reconnecting portions of the refueling canal to the RCS, this SR must be met per SR 4.0.4. If 272 any dilution activity has occurred while the refueling canal was disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

3/4.9.2 INSTRUMENTATION The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core, such as may be caused by a boron dilution accident. Note, however, that the boron dilution event is considered unlikely for ANO-2 due to the significant period of time for operator detection and response before SDM would be significantly challenged (reference ANO-2 SAR Section 15.1.4.3).

Suspending positive reactivity additions and the movement of fuel, sources, and reactivity control components ensures that positive reactivity is not inadvertently added to the reactor core while the source range neutron flux monitor is inoperable. Action "a" is modified by a Note that states that fuel assemblies, sources, and reactivity control components may be moved if necessary to facilitate repair or replacement of the inoperable source range neutron flux monitor. It may be 571-T necessary to move these items away from the locations in the core close to the source range neutron flux monitor to minimize personnel radiation dose during troubleshooting or repair. The Note also permits completion of movement of a component to a safe position, should the source range neutron flux monitor be discovered inoperable during component movement.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

ARKANSAS - UNIT 2 B 3/4 9-1 Amendment No. 43,166,203,230,240 Rev. 48,58,

REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT PENETRATIONS 51 The requirements on containment penetration closure ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel handling accident (FHA)element rupture during movement of recently irradiated fuel or FHA movement of new fuel assemblies over recently irradiated fuel assemblies based upon the lack of containment pressurization potential while in the REFUELING MODE. Due to radioactive decay, a FHA which does not involve movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) or the movement 51 of new fuel assemblies over recently irradiated fuel assemblies will result in doses that are well within the guideline values specified in 10 CFR 50.67 even without containment closure capability.

Containment penetrations, the personnel airlock doors, and/or the equipment door may be open during movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) or movement of new fuel assemblies over FHA recently irradiated fuel assemblies within the containment and during CORE ALTERATIONS provided a minimum of one closure method (manual or automatic valve, blind flange, or 471 equivalent) in each penetration, one door in each airlock, and the equipment door are capable of being closed in the event of a FHAfuel handling accident. This allowance assumes that 23 feet of water is maintained above the fuel seated within the reactor vessel to ensure any offsite FHA dose consequence remains within 10 CFR 50.67 limits in the event of a FHAfuel handling accident. Note that eEquivalent isolation methods must be approved and may include use of a material that can provide a temporary atmospheric pressure ventilation barrier. For closure, the equipment door will be held in place by a minimum of four bolts.

During the movement of irradiated fuel, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly.

The basis for inclusion of the "recently" term as it relates to the movement of irradiated fuel is the reduction in dose consequences due to such decay. Likewise, the goal of maintaining 51 ventilation system and radiation monitoring availability is to reduce dose consequences even (commitment) further below that provided by natural decay. Therefore, the availability of ventilation systems and radiation monitors that serve this purpose should be considered commensurate with their contribution to public safety, regardless of whether the analyzed decay time (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) has passed.

In addition to the above, a single normal or contingency method to promptly close containment penetrations should be established regardless of whether the analyzed decay time (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) 51 has passed. Such prompt methods need not completely block the penetration or be capable of (commitment) resisting pressure. The purpose of establishing such prompt methods of containment closure is to is to enable ventilation systems to draw release gases from a postulated FHA in the proper direction such that it can be treated and monitored.

ARKANSAS - UNIT 2 B 3/4 9-1 Amendment No. 43,166,203,230,240 Rev. 48,58,

REFUELING OPERATIONS BASES 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during the movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical FHA reactor core within the previous 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) or the movement of new fuel assemblies over recently irradiated fuel assembliesCORE ALTERATIONS.

3/4.9.6 REFUELING MACHINE OPERABILITY FHA The OPERABILITY requirements for the refueling machine ensure that: 1) the refueling machine will be used for movement of CEAs with fuel assemblies and that it has sufficient load capacity to lift a fuel assembly, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly, CEA and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

For the spent fuel storage building crane, the normal configuration is with the power disconnect and travel interlock in place to ensure that a load in excess of 2000 pounds is not inadvertently carried over spent fuel. The use of the spent fuel storage building crane to lift the fuel pool gates requires travel beyond the area where the power disconnect and travel interlock provide protection. In this configuration additional controls are required to ensure the limiting condition for operation is met. The safe load path and heavy load permit provide the necessary controls to ensure loads in excess of 2000 pounds are not carried over spent fuel when the fuel pool gates are being lifted. Before the lift is made the surveillance requirement must still be satisfied.

3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 °F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

The requirement to have two shutdown cooling loops OPERABLE when there is less than 23 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.

ARKANSAS - UNIT 2 B 3/4 9-2 Amendment No. 24,29 Rev. 10,

REFUELING OPERATIONS BASES 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION (continued)

Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant 286 inventory must be from sources that have a boron concentration greater than that required in Insert B11 the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 12% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.11 DELETED 3/4.9.12 FUEL STORAGE The spent fuel storage racks are designed to assure that when fuel assemblies of less than or equal to 4.95 w/o U-235 enrichment are stored within the limits of Table 3.9.1 subcritical array with K eff 0.95 will be maintained when a concentration of 452 ppm of soluble boron is present in the spent fuel water. These conditions have been verified by criticality analyses.

The requirement for 2000 ppm boron concentration is to assure the fuel assemblies will be maintained in a subcritical array with K eff 0.95 in the event of a postulated accident. Analysis has shown that, during a postulated accident with the fuel stored within the limits of this specification, that a Keff of 0.95 will be maintained when the boron concentration is at or above 881 ppm.

The peripheral cells are defined as those storage cells closest to the spent fuel pool wall that have fuel assemblies located in them. Therefore, if the storage cell closest to the spent fuel pool wall is kept empty, then the second storage cell from the spent fuel pool wall may be filled with lower burnup fuel meeting the requirements of Table 3.9-1.

ARKANSAS - UNIT 2 B 3/4 9-4 Amendment No. 43,166,178,224,228 Rev. 3,28,29,

Enclosure, Attachment 3 to 2CAN081903 Retyped Technical Specification Pages (19 pages)

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c. Digital computer channels - The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control element assemblies are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except controlled leakage) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

ARKANSAS - UNIT 2 1-3 Amendment No. 157,220,255,266,

REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the Reactor Coolant System (RCS) shall be 2000 gpm whenever a reduction in RCS boron concentration is being made.

APPLICABILITY: ALL MODES.

ACTION:

With the flow rate of reactor coolant through the RCS < 2000 gpm, immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM or boron concentration of LCO 3.1.1.1, LCO 3.1.1.2, or LCO 3.9.1, as applicable.

SURVEILLANCE REQUIREMENTS 4.1.1.3 The flow rate of reactor coolant through the RCS shall be determined to be 2000 gpm within one hour prior to the start of and in accordance with the Surveillance Frequency Control Program during a reduction in the RCS boron concentration by either:

a. Verifying at least one reactor coolant pump is in operation, or

b. Verifying that at least one low pressure safety injection pump or containment spray pump is in operation as a shutdown cooling pump and supplying 2000 gpm through the RCS.

ARKANSAS - UNIT 2 3/4 1-4 Amendment No. 126,255,315,

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1. AREA MONITORS

a. Spent Fuel Pool Area Monitor 1 Note 1 1.5 x 10-2 R/hr 10 101 R/hr 13

b. Containment High Range 2 1, 2, 3, & 4 Not Applicable 1 - 107 R/hr 18

2. PROCESS MONITORS

a. Containment Purge and Exhaust Isolation 1 Note 3 2 x background 10 - 106 cpm 16

b. Control Room Ventilation Intake Duct Monitors 2 Note 2 2 x background 10 - 106 cpm 17,20,21

c. Main Steam Line Radiation Monitors 1/Steam Line 1, 2, 3, & 4 Not Applicable 10 104 mR/hr 19 Note 1 - With fuel in the spent fuel pool or building.

Note 2 - MODES 1, 2, 3, 4, and during movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies.

Note 3 - Applicable during:

a. PURGE of the Containment Building or,

b. Containment Building continuous ventilation operations when moving recently irradiated fuel assemblies or moving new fuel assemblies over recently irradiated fuel assemblies in the Containment Building.

ARKANSAS - UNIT 2 3/4 3-25 Amendment No. 63,130,145,206,231,255,

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 13 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 16 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, complete the following:

a. If moving recently irradiated fuel assemblies or moving new fuel assemblies over recently irradiated fuel assemblies within the Containment Building, secure the Containment Purge System or suspend the movement of recently irradiated fuel assemblies and movement of new fuel assemblies over recently irradiated fuel assemblies within the Containment Building.

b. If a Containment PURGE is in progress, secure the Containment Purge System.

c. If continuously ventilating the Containment Building, verify the associated SPING monitor operable or perform the applicable ACTION(s) of the Offsite Dose Calculation Manual; otherwise, secure the Containment Purge System.

ACTION 17 - In MODE 1, 2, 3, or 4, with no channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the control room emergency ventilation system (CREVS) in the recirculation mode of operation or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

ACTION 18 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, (1) either restore the inoperable channel to OPERABLE status within 7 days or (2) prepare and submit a Special Report to the NRC within 30 days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status. With both channels inoperable, initiate alternate methods of monitoring the containment radiation level within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in addition to the actions described above.

ACTION 19 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the NRC within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ARKANSAS - UNIT 2 3/4 3-26 Amendment No. 63,130,145,206,231, 255,301,

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 20 - In MODE 1, 2, 3, or 4 with the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, within 7 days restore the inoperable channel to OPERABLE status or initiate and maintain the CREVS in the recirculation mode of operation. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

ACTION 21 - During movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies with one or two channels inoperable, immediately place one OPERABLE CREVS train in the emergency recirculation mode or immediately suspend the movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies.

ARKANSAS - UNIT 2 3/4 3-26a Amendment No. 301,

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE INSTRUMENT CHECK CALIBRATION FUNCTIONAL TEST REQUIRED

1. AREA MONITORS

a. Spent Fuel Pool Area Monitor SFCP SFCP SFCP Note 1

b. Containment High Range SFCP SFCP Note 4 SFCP 1, 2, 3, & 4

2. PROCESS MONITORS

a. Containment Purge and In accordance with Note 2 Note 3 Note 3 Exhaust Isolation applicable Notes

b. Control Room Ventilation SFCP SFCP SFCP Note 6 Note 5 Intake Duct Monitors

c. Main Steam Line SFCP SFCP SFCP 1, 2, 3, & 4 Radiation Monitors Note 1 - With fuel in the spent fuel pool or building.

Note 2 - Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to initiating Containment PURGE operations and in accordance with the Surveillance Frequency Control Program during Containment PURGE or continuous ventilation operations.

Note 3 - Within 31 days prior to initiating Containment PURGE operations and in accordance with the Surveillance Frequency Control Program during Containment continuous ventilation operations when moving recently irradiated fuel assemblies or moving new fuel assemblies over recently irradiated fuel assemblies in the Containment Building.

Note 4 - Acceptable criteria for calibration are provided in Table II.F.1-3 of NUREG-0737.

Note 5 - MODES 1, 2, 3, 4, and during movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies.

Note 6 - When the Control Room Ventilation Intake Duct Monitor is placed in an inoperable status solely for performance of this Surveillance, entry into associated ACTIONS may be delayed up to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

ARKANSAS - UNIT 2 3/4 3-27 Amendment No. 63,130,145,206,231,255,315, Next Page is 3/4 3-36

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. The reactor coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop (A) and at least one associated reactor coolant pump.

2. Reactor Coolant Loop (B) and at least one associated reactor coolant pump.

b. At least one of the above Reactor Coolant Loops shall be in operation.*

APPLICABILITY: MODE 3.

ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 and immediately initiate corrective action to return the required loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1, and (2) core outlet temperature is maintained at least 10 °F below saturation temperature.

ARKANSAS - UNIT 2 3/4 4-2 Amendment No. 24,29,315,

REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop (A) and its associated steam generator and at least one associated reactor coolant pump.

2. Reactor Coolant Loop (B) and its associated steam generator and at least one associated reactor coolant pump.

3. Shutdown Cooling Loop (A) #.

4. Shutdown Cooling Loop (B) #.

b. At least one of the above coolant loops shall be in operation.*

APPLICABILITY: Modes 4 and 5.

ACTION:

a. With less than the above required coolant loops OPERABLE, immediately initiate status as soon as possible and initiate action to make at least one steam generator available for decay heat removal via natural circulation. LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

b. With no coolant loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 or LCO 3.1.1.2, as applicable, and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required shutdown cooling loop(s) shall be determined OPERABLE per the INSERVICE TESTING PROGRAM.

4.4.1.3.2 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be 23% indicated level in accordance with the Surveillance Frequency Control Program.

4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

All reactor coolant pumps and decay heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 or LCO 3.1.1.2, as applicable, and (2) core outlet temperature is maintained at least 10 °F below saturation temperature.

The normal or emergency power source may be inoperable in Mode 5.

ARKANSAS - UNIT 2 3/4 4-2a Amendment No. 24,29,233,301,305, 315,

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION AND AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two independent control room emergency ventilation and air conditioning systems shall be OPERABLE. (Note 1)

APPLICABILITY: MODES 1, 2, 3, 4, or during movement of irradiated fuel assemblies or movement of new fuel assemblies over irradiated fuel assemblies.