Text

License Amendment Request Technical Specification Deletions, Additions, and Relocations
	ML19350B324
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/16/2019
From:	Gaston R; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	2CAN121901
	Download: ML19350B324 (118)
	v • d • e

10 CFR 50.90 2CAN121901 December 16, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

License Amendment Request Technical Specification Deletions, Additions, and Relocations Arkansas Nuclear One, Unit 2 NRC Docket No. 50-368 Renewed Facility Operating License No. NPF-6

Reference:

Nuclear Regulatory Commission Federal Register Notice 58 FRN 39132, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," dated July 22, 1993.

As required by 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests changes to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (TSs) to enhance consistency with NUREG-1432, "Standard Technical Specifications - Combustion Engineering Plants," Revision 4, and the referenced Nuclear Regulatory Commission (NRC) policy statement. The ANO-2 TSs have not yet been converted to the standard technical specifications (STS) of NUREG-1432. Subsequently, Entergy desires to add, delete, relocate, and/or otherwise modify certain ANO-2 TSs, consistent with the referenced NRC policy statement and the STS. The proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

Approval of the proposed amendment is requested by January 29, 2021.

New regulatory commitments are included in this amendment request.

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.

Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing

2CAN121901 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on December 16, 2019.

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

Enclosure:

Evaluation of the Proposed Change Attachments to

Enclosure:

1.

Technical Specification Page Markups

2.

Technical Specification Bases Page Markups

3.

Retyped Technical Specification Pages

4.

Technical Requirements Manual Draft Markups

5.

List of Regulatory Commitments cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Designated Arkansas State Official

Enclosure to 2CAN121901 Evaluation of the Proposed Change

Enclosure to 2CAN121901 Page 1 of 42 EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION Entergy Operations, Inc. (Entergy) requests U.S. Nuclear Regulatory Commission (NRC) review and approval of a proposed amendment to the Arkansas Nuclear One, Unit 2 (ANO-2) Renewed Facility Operating License NPF-6, Appendix A, Technical Specifications (TSs), to revise several TS requirements by the addition, deletion, or relocation of certain TS Limiting Conditions of Operation (LCOs), Actions, and Surveillance Requirements (SRs). Relocated TSs will be placed in the ANO-2 Technical Requirements Manual (TRM) or the associated TS Bases. The requested change involves no significant hazards consideration.

The NRCs Final Policy Statement on TS improvements for nuclear power reactors (Reference 1) and 10 CFR 50.36 allowed certain requirements to be relocated from the TSs to other licensee-controlled documents. Candidate TSs are generally relocated to licensee-controlled documents such as the TRM. The Final Policy Statement specified that TS LCOs which do not meet any of the four criteria listed in 10 CFR 50.36(c)(2)(ii) may be proposed for removal from the TSs and relocated to licensee-controlled documents.

The ANO-2 TSs are modeled after the original Combustion Engineering (CE) standard TSs of NUREG-0212. The changes proposed by Entergy in this amendment request are intended to provide added consistency between the ANO-2 TSs and the improved standard of NUREG-1432, "Standard Technical Specifications - Combustion Engineering Plants,"

Revision 4 (Reference 2).

2.0 DETAILED DESCRIPTION

2.1 Background

In the 1970s through the early 1980s, the Standard Technical Specifications (STS) for CE Pressurized Water Reactors (PWRs) (CE-STS) was a generic document (i.e., NUREG-0212) prepared by the NRC for use in the licensing process of CE PWRs (Reference 3). The original ANO-2 TSs were subsequently issued as NUREG-0336 on July 18, 1978 (Reference 4).

NUREG-0212, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors," was revised periodically to reflect changes in generic licensing requirements.

Revision 1 of NUREG-0212 (Reference 5) removed the TS requirement for Boron Dilution.

Later, the CE Owners Group (CEOG) initiated a program to restructure the STSs in order to develop improved STSs for CEOG plants in accordance with the 1985 recommendations of the NRC Technical Specification Improvement Project and Atomic Industrial Forum Subcommittee for Technical Specification Improvement. From these recommendations the current 10 CFR 50.36(c)(2)(ii) criteria were developed. Subsequently, improved vendor-specific STSs were developed and issued by the NRC in September 1992 (Reference 6). The improved STS were published as NRC reports including NUREG-1432. Changes to the STS NUREGs are developed through the industry and the NRC by the Technical Specification Task Force (TSTF) traveler process. For example, TSTF-266, "Eliminate the Remote Shutdown System Table of Instrumentation and Controls" (Reference 7), removed the list of Remote Shutdown instrumentation from the TSs as approved by the NRC in 1999 (Reference 8). This list was subsequently added to the associated TS Bases.

Enclosure to 2CAN121901 Page 2 of 42 2.2 Current Requirement The following table lists each affected TS and a summary of the current requirements associated with each. Due to the amount of information involved, please refer to Attachments 1 and 2 of this enclosure for a complete markup of the affected TSs and associated TS Bases, respectively.

TS #

Brief Title Brief Summary of Requirements 3.1.1.3 Boron Dilution Requires a minimum Reactor Coolant System (RCS) flow during dilution evolutions.

3.3.3.1 Radiation Monitoring Item 2.c requires one Main Steam Line (MSL) radiation monitor per steam line.

3.3.3.5 Remote Shutdown Requires certain instruments to assist remote shutdown operations.

3.4.6.2 RCS Leakage Surveillance Requirement (SR) 4.4.6.2.1.b requires monitoring of the Reactor Vessel Head (RVH) flange leakoff temperature.

3.7.2.1 SG P/T Places limits on temperature differential between the RCS and the Steam Generator (SG) secondary fluid.

3.7.5.1 Flood Protection Establishes flood protection measures based on lake level.

3.7.9.1 Sealed Sources Limits surface contamination of certain sealed sources.

3.7.12 SFP Structural Integrity Provides inspection requirements to verify continued structural integrity of the Spent Fuel Pool (SFP).

3.9.3.a Decay Time Limits fuel movement from the core until a specified decay time has been met.

3.9.5 Communications Requires communications between the Control Room and refueling personnel.

3.9.6 Refueling Machine Establishes limits and interlocks for the Refueling Machine (located in the Containment Building).

3.9.7 SFP Crane Travel Limits loads over the fuel stored in the SFP.

3.11.1 Liquid Holdup Limits quantity of radioactivity in unprotected outside temporary liquid storage tanks.

3.11.2 Gas Storage Limits quantity of radioactivity in gas storage tanks.

3.11.3 Explosive Gas Places controls on the mixture of hydrogen and oxygen contained in a waste gas storage tank.

Enclosure to 2CAN121901 Page 3 of 42 2.3 Reason for the Proposed Change The proposed amendment eliminates redundant TS requirements, adds a program to govern certain TS radiological control requirements, and relocates several TS requirements to licensee-controlled documents. This effort is intended to provide greater consistency with NUREG 1432, Revision 4 (Reference 2), and the referenced NRC Final Policy Statement (Reference 1).

2.4 Description of the Proposed Change The proposed amendment removes instrument listings from TS 3.3.3.5, "Remote Shutdown."

As an Operator aid, this information will be added to the associated ANO-2 TS Bases.

Requirements of some other TSs are also removed or proposed for relocation to the ANO-2 TRM. At ANO-2, the TRM is considered part of the Safety Analysis Report (SAR) and is controlled under 10 CFR 50.59.

ANO-2 TS 3.1.1.3, "Boron Dilution," requirements are proposed for deletion because these requirements are covered by other LCOs. The proposed amendment also adds an Explosive Gas and Storage Tank Radioactivity Monitoring Program to govern activities associated with liquid and gaseous storage tanks, which subsequently results in the deletion of the related TSs (TSs 3.11.1, 3.11.2, and 3.11.3).

TS requirements associated with the RVH flange leakoff temperature indication (TS 3.4.6.2),

sealed sources (TS 3.7.9.1), and refueling communications (TS 3.9.5) are deleted due to be controlled by other regulation or otherwise not meeting the TS inclusion criteria of 10 CFR 50.36(c)(2)(ii).

The proposed changes are consistent with the STS in that the subject requirements being deleted or relocated are not contained in the STS. The following table provides a summary of changes associated with each affected ANO-2 TS.

TS #

Brief Title Proposed Change 3.1.1.3 Boron Dilution TS and TS Bases are deleted. Sufficient RCS flow to support proper mixing during dilutions is provided by TSs 3.4.1.a, 3.4.1.2, 3.4.1.3, and 3.9.8.1.

3.3.3.1 Radiation Monitoring The MSL Radiation Monitors, associated Action 19 of Table 3.3-6, the associated SRs of Table 4.3-3, and the associated TS Bases are relocated to the TRM.

3.3.3.5 Remote Shutdown The Remote Shutdown instruments listed in Table 3.3-9 and associated SRs listed in Table 4.3-6 are removed from the TS. As an Operator aid, the lists will be added to the respective TS 3.3.3.5 Bases.

3.4.6.2 RCS Leakage RVH flange leakoff temperature SR 4.4.6.2.1.b is deleted.

This temperature indication does not meet the 10 CFR 50.36(c)(2)(ii) criteria for TS inclusion. The TS 3.4.6.2 Bases do not discuss this instrument.

Enclosure to 2CAN121901 Page 4 of 42 TS #

Brief Title Proposed Change 3.7.2.1 SG P/T TS and TS Bases are relocated to the TRM.

3.7.5.1 Flood Protection TS and TS Bases are relocated to the TRM.

3.7.9.1 Sealed Sources TS and TS Bases is deleted. Controlled by 10 CFR 70.39.

3.7.12 SFP Structural Integrity TS and TS Bases are relocated to the TRM.

3.9.3.a Decay Time TS and TS Bases are relocated to the TRM.

3.9.5 Communications TS and TS Bases is deleted. 10 CFR 50, Appendix B.

provides requirements for establishing procedures to control plant evolutions.

3.9.6 Refueling Machine TS and TS Bases are relocated to the TRM.

3.9.7 SFP Crane Travel TS and TS Bases are relocated to the TRM.

3.11.1 Liquid Holdup TS is relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program. Associated TS Bases are deleted.

3.11.2 Gas Storage TS is relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program. Associated TS Bases are deleted.

3.11.3 Explosive Gas TS is relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program. Associated TS Bases are deleted.

6.5.8 Explosive Gas Add new Explosive Gas and Storage Tank Radioactivity Monitoring Program consistent with the STS.

Because several TSs are removed, the TS page preceding the removed TS may be included in this request in order to specify the "next" TS page number in the footer or to remove reference to a "next" page. For example, the removal of TS 3.1.1.3 eliminates TS Page 3/4 1-4; therefore, the footer of TS Page 3/4 1-3 is revised to state that the next page is 3/4 1-5.

Relative to the above page numbering, there exists a TS page that is numbered "3/4 7-11 3/4 7-12 3/4 7-13". Information from this page was removed via a previous amendment and, therefore, this page is being deleted from the TS, with the footer of the preceding page (3/4 7-10) revised to indicate the next page to be 3/4 7-15 (due to the removal of the deleted page and the removal of TS 3.7.2.1 as listed in the table above).

No technical changes are made to the subject preceding pages. Because the footer revisions and the removal of a page associated with the previously deleted TS are administrative in nature, no further discussion of these changes are included in this submittal. The affected preceding pages are listed below:

3/4 1-3 3/4 7-10 3/4 7-18 3/4 9-2 3/4 9-4

Enclosure to 2CAN121901 Page 5 of 42 With the relocation of TS 3.3.3.1 MSL Radiation Monitor requirements and the associated Action 19, Actions 20 and 21 are moved from TS Page 3/4 3-26a to TS Page 3/4 3-26.

TS Page 3/4 3-26a is subsequently deleted in its entirety. In addition, TS Page 34 3-28 (i.e., a previously deleted TS) is removed from the TSs and the notation "Next page is 3/4 3-36" is relocated to the footer of previous page 3/4 3-27. These changes are also considered administrative and are not discussed further in this submittal.

The removal of Remote Shutdown TS Tables 3.3-9 and 4.3-6 results in deletion of TS Pages 3/4 3-37 and 3/4 3-38. Therefore, the footer of TS Page 3/4 3-36 is modified to state that the next page is 3/4 3-39. This change is considered administrative and is not discussed further in this submittal.

With the deletion of SR 4.4.6.2.1.b as denoted in the table above, only one SR remains under SR 4.4.6.2.1. Therefore, the sub-bullet "a." is deleted and the requirement of this sub-bullet combined into SR 4.4.6.2.1. This change is considered administrative and is not discussed further in this submittal.

Attachments 1 and 2 of this enclosure contain a markup of affected TS pages and TS Bases pages, respectively. Attachment 3 contains the retyped (clean) version of the affected TS pages. To aid in NRC review, a draft markup of the TRM changes (relevant to information relocated to the TRM) is provided in Attachment 4.

Attachments 2 and 4 of this enclosure are provided for information only. Entergy will revise the ANO-2 TS Bases and ANO-2 TRM consistent with the markups provided in Attachments 2 and 4 of this enclosure in accordance with the TS Bases Control Program of ANO-2 TS 6.5.14 and 10 CFR 50.59, as applicable. Entergy considers this to be a regulatory commitment which is also stated in Attachment 5.

3.0 TECHNICAL EVALUATION

On July 22, 1993, the NRC issued its Final Policy Statement (Reference 1), expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Atomic Energy Act of 1954, as amended (42 U.S.C. 2232), and 10 CFR 50.36. The Final Policy Statement described the safety benefits of the STS and encouraged licensees to use the STS as the basis for plant-specific TS amendments, and for complete conversions to the improved STS. Further, the Final Policy Statement gave guidance for evaluating the required scope of the TSs and defined the guidance criteria to be used in determining which of the LCOs and associated surveillances should remain in the TS. By this approach, existing LCO requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TS and those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The Commission codified the four criteria in 10 CFR 50.36 (Reference 9).

These four criteria contained in 10 CFR 50.36(c)(2)(ii) can be used to determine the requirements that must be included in the TSs. Items not meeting any of the four criteria can be removed from TSs, and as appropriate, placed in a licensee-controlled document. Entergy is then allowed to change the relocated requirements, if necessary, in accordance with 10 CFR 50.59 or another NRC-approved method. This should result in significant reductions in

Enclosure to 2CAN121901 Page 6 of 42 time and expense to modify requirements that have been removed from the TSs following an evaluation of effects on nuclear and public safety. The criteria and an evaluation of each technical specification proposed for relocations are provided below.

The four criteria defined by 10 CFR 50.36(c)(2)(ii) for determining whether a structure, system, or component (SSC), process variable, design feature, or operating feature is required to be included in the TS LCOs, are as follows:

(A)

Installed instrumentation that is used to detect, and indicate in the control room, significant abnormal degradation of the reactor coolant pressure boundary; (B) a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier; (C) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier; (D) a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The proposed amendment deletes or relocates to the TRM several TS requirements which are redundant, controlled by other regulation, or otherwise do not meet the TS inclusion criteria stated above. Some information from TS 3.3.3.5, "Remote Shutdown," is removed and, as an Operator aid, placed in the associated TS Bases. Three radiological effluent related TSs are deleted in favor of a new Explosive Gas and Storage Tank Radioactivity Monitoring Program which is proposed for adoption in ANO-2 TS Section 6.0, "Administrative Controls." Each affected TS and associated justification for change is discussed below. Due to the amount of TS-specific information involved, please refer to Attachments 1 through 4 of this enclosure for cases where an entire TS is being deleted or relocated.

TS 3.1.1.3 - Boron Dilution The Chemical and Volume Control System (CVCS) regulates both the chemistry and the quantity of coolant in the RCS. Changing the boron concentration in the RCS is a normal plant operation, compensating for reactivity effects such as fuel burnup, xenon buildup, and temperature changes. Boron dilution is a manual operation conducted under strict procedural controls, which specify permissible limits on the rate and magnitude of any required change in boron concentration. Boron concentration in the RCS can be decreased either by controlled addition of unborated makeup water, with a corresponding removal of reactor coolant (feed and bleed), or by using one of the letdown ion exchangers.

Six different general operational modes were analyzed for the boron dilution incident: dilution during refueling, cold shutdown, hot shutdown, hot standby, and during low and full power operation. During normal plant operation, the operation of more than one charging pump is not the normal mode. However, in each case it was assumed that the boron dilution results from injecting unborated demineralized water into the RCS at the maximum possible rate of

Enclosure to 2CAN121901 Page 7 of 42 138 gallons per minute (gpm) (the combined capacity of three charging pumps). The analysis assumed the boron concentration within the minimum volume considered in each analyzed mode is uniform at all times because sufficient circulation exists to maintain a uniform mixture.

As stated in the TS 3.1.1.3 Bases, 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 RCS and will, therefore, be within the capability of operator recognition and control.

TS 3.1.1.3 requires maintaining RCS flow 2000 gpm whenever a reduction in RCS boron concentration is being made. Entergy requests TS 3.1.1.3 be deleted because RCS flow is assured under LCOs 3.4.1.1 (Modes 1 and 2), 3.4.1.2 (Mode 3), 3.4.1.3 (Modes 4 and 5), and 3.9.8.1 (Mode 6). LCO 3.4.1.1 and LCO 3.4.1.2 require one or more Reactor Coolant Pumps (RCPs) to be in service. The design capacity of each RCP is 80,000 gpm (ANO-2 SAR Table 1.3-1), much greater than the 2000 gpm stipulated in LCO 3.1.1.3. Because ANO-2 cannot operate at power with less than four RCPs in operation, a loss of just one RCP would result in a unit trip and entry into Mode 3. Mode 3 LCO 3.4.1.2 requires reductions in boron concentration to be suspended if no RCP is in operation.

LCO 3.4.1.3 (Modes 4 and 5) requires either an RCP or a Shutdown Cooling (SDC) pump to be in service during operation in Modes 5 and 6. Again, with no RCP or SDC pump in service, reductions in boron concentration must be suspended. Normally the Low Pressure Safety Injection (LPSI) pumps serve as the SDC pumps (a Containment Spray pump may be aligned for SDC when RCS pressure is < 50 psig). Which acronym is used to describe one of these pumps depends on the system alignment. When aligned for accident mitigation, the pumps are referred to as LPSI or Containment Spray pumps. When aligned for shutdown decay heat removal, the pumps are referred to as SDC pumps.

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.

LCO 3.4.1.3 does not specify a minimum flowrate; however, procedures caution against operating a LPSI pump at flows < 2000 gpm to avoid cavitation, vibration, and bearing wear.

When a Containment Spray pump is used for SDC, procedures require maintaining a flow rate between 2300 and 2500 gpm.

LCO 3.9.8.1 requires at least one SDC loop to be in operation with a flow rate of 2000 gpm while operating in Mode 6. Again, with no SDC pump in service, reductions in boron concentration must be suspended.

Based on the controls describe above, Entergy has concluded that a separate TS governing boron dilution is not required. In addition, NUREG-1432 does not include a separate specification for boron dilutions. Therefore, Entergy proposes that LCO 3.1.1.3 be deleted.

This proposed change is consistent with NUREG-1432.

Enclosure to 2CAN121901 Page 8 of 42 TS 3.3.3.1 - Radiation Monitoring Instrumentation The MSL Radiation Monitor System (RMS) is an Eberline Model RMS. The detectors are remote detector assemblies, one attached on each MSL between the containment penetration and the Main Steam Safety Valves (MSSVs). The detectors provide a signal to an indicator in the Control Room and to the Safety Parameter Display System (SPDS). The signal from each indicator is also supplied to a dual pen recorder in the Control Room.

NUREG-0737, Item II.F.1, Attachment 1, (Reference 10) required that noble gas effluent monitors with an extended range be installed to function during accident conditions as well as during normal operating conditions. NUREG-0737 specifically listed PWR steam safety valve discharge / atmospheric steam dump valve discharge as an effluent pathway which should be monitored. In response to this requirement, MSL RMS was installed at ANO-2 and subsequently added to the ANO-2 TSs under Amendment 145 in March of 1993 (Reference 11).

ANO-2 TS 3.3.3.1 requires a MSL radiation monitor to be operable on each MSL with alarm/trip setpoints within the specified limits.

NUREG-0737, Item II.F.1, Attachment 1, Noble Gas Effluent Monitors referenced NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (Reference 12), Recommendation 2.1.8, "Instrumentation to Follow the Course of an Accident,"

Item b, "Increased Range of Radiation Monitors." The Item b recommendation was two-fold:

(1) to provide high range radiation monitors for noble gases in plant effluent lines and a high range radiation monitor in the containment, and (2) to provide instrumentation for monitoring effluent release lines capable of measuring and identifying radioiodine and particulate radioactive effluents under accident conditions. Along with this recommendation the NRC stated:

"The Offices of Standards Development and Nuclear Reactor Regulation have agreed to expedite revision of Regulatory Guide 1.97, which deals with this subject area, and its early implementation for all operating plants and plants under construction. It is expected that the necessary revisions would be developed within a few months and implementation would follow soon afterward. In the meantime, the following provisions are recommended for early implementation on all plants to provide a uniform, minimum capability in this area."

In December of 2017, Entergy submitted a license amendment request (LAR) intended to provide greater consistency with NUREG 1432, Revision 4, and ensure that both Category 1 and Type A RG 1.97 instrumentation is included in the ANO-2 TSs (unless already addressed within another specification) (Reference 13). The LAR stated that ANO-2 is designed with installed post-accident monitoring (PAM) instrumentation of the various types and categories referred to in Revision 3 of RG 1.97 (Reference 14). The NRC approved the LAR in December of 2018 as ANO-2 TS Amendment 313 (Reference 15).

The ANO-2 TS Amendment 313 NRC Safety Evaluation (SE) discussed that the changes were intended to ensure both Category 1 and Type A RG 1.97 instrumentation are included in the PAM specification (unless already addressed within another specification) and gains greater consistency with NUREG-1432, Revision 4. In Section 3.2.1.7 of the SE the NRC listed the variables that were considered under the scope of RG 1.97, but were addressed by other specifications. The other variables are the neutron flux and containment radiation monitors.

MSL radiation monitors were not listed in Section 3.2.1.7 of the SE and are not listed in PAM

Enclosure to 2CAN121901 Page 9 of 42 TS 3.3.3.6. This is because the MSL radiation monitors are not considered a Type A or Category 1 variable in accordance with the ANO-2 licensing basis related to RG 1.97 as evaluated below.

Type A variables are those which provide the primary information required to permit the control room operators to take specific manual actions for which no automatic control is provided and that are required for a safety system to accomplish its safety function for design basis accident scenarios. A Category 1 designation provides for the most stringent requirements and are intended for key variables. Type A, B, and C key variables fall into this category. In accordance with NUREG 1432, all Type A and Category 1 variables should be included in the TSs.

Table 7.5-3 of the ANO-2 SAR lists the variables committed to by Entergy as recommended by RG 1.97, Revision 3. The table includes the assigned category, range, redundancy, power supply, type of Control Room display, availability on the SPDS, and comments. Table 7.5-3 separates the variables by Type and Category. Radiation detection for the two MSLs is listed as a Type E, Category 2 variable. Type E variables are those which provide information for use in determining the magnitude of a release of radioactive materials and for use in assessing the consequences of such releases. A Category 2 designation provides less stringent requirements and applies to instrumentation designated for indicating system operating status. Type D and E key variables fall into this category. Because the MSL radiation detectors are not classified as Type A or Category 1 variables, these instruments are not required to be included in the TSs.

The four 10 CFR 50.36 criteria were reviewed in support of removing the MSL radiation monitor requirements from the TSs.

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).

The MSL radiation monitors detect radioactivity transported through either MSL during events which involve SG tube leakage. However, other indications such as RCS level and pressure are the primary means for detecting a significant degradation of the RCPB. SG level and feed rates are also used to determine the affected SG. RCS level, RCS pressure, SG level, and SG feed rate (Emergency Feedwater flow rate) instruments are included in PAM TS 3.3.3.6 (as Type A and/or Category 1 variables). As stated above, the MSL radiation detectors were installed to monitor potential post-accident radioactive release via the main steam system (a Type E variable). The intended purpose and classification of the MSL radiation monitors are consistent with other non-TS instrument capability such as the main condenser offgas radiation monitor (a Type C, Category 3 variable) and the MSSV position indicators (a Type D, Category 2 variable). The MSL radiation monitors are not installed instrumentation intended to meet Criterion 1.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The MSL radiation monitors are provided for monitoring effluent radioactive release under accident conditions. The monitors do not provide direct input to the Reactor Protection System (RPS) or Engineered Safety Features Actuation System (ESFAS) functions, nor are

Enclosure to 2CAN121901 Page 10 of 42 the monitors a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 - An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

The MSL radiation monitors are not part of a primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The instruments monitor potential radioactive release during an event that involves a SG tube leak, but are not a primary success path used to determine the response of plant structures, systems, or components (SSCs) with respect to accident mitigation. With respect to SG tube leakage, the affected SG is determined by monitoring SG level and/or SG feed rates, both of which are included in PAM TS 3.3.3.6 as Type A and/or Category 1 variables.

Criterion 4 - An SSC which operating experience or probabilistic risk assessment (PRA) has shown to be significant to public health and safety.

The MSL radiation monitors have not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS requirement does not involve an SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

This MSL radiation monitor TS requirements do not fulfill any of the 10 CFR 50.36c(2)(ii) criteria on items for which TSs must be established and are not RG 1.97 Type A or Category 1 variables. Therefore, these requirements may be relocated from TSs to the TRM.

Based on the above, Entergy proposes that the TS operability, action, and surveillance requirements contained in TS 3.3.3.1 for the MSL radiation monitors, along with the associated TS Bases, be relocate to the TRM. Because NUREG 1432 only contains Type A and Category 1 variables, this change is consistent with the STS.

TS 3.3.3.5 - Remote Shutdown Instrumentation The Remote Shutdown (RSD) instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of hot standby (Mode 3) conditions from locations outside of the Control Room and in conjunction with procedures, reach cold shutdown (Mode 5) conditions if necessary. This capability is required in the event Control Room habitability is lost and is consistent with General Design Criteria (GDC) 19 of 10 CFR 50.

ANO-2 LCO 3.3.3.5 requires the RSD instrumentation channels shown in TS Table 3.3-9 to be operable with readouts displayed external to the Control Room. Operability is verified through required Channel Checks and Channel Calibrations in accordance with TS Table 4.3-6. Entergy requests that TS Table 3.3-9 and TS Table 4.3-6 be removed from the TS. As an Operator aid, this information will be placed in the associated TS 3.3.3.5 Bases. Reference to either of these tables in LCO 3.3.3.5, the associated Action Statement, and SR 4.3.3.5 is removed. Changes to the tables will be controlled under TS 6.5.14, "Technical Specifications (TS) Bases Control Program," and 10 CFR 50.59, as applicable.

Enclosure to 2CAN121901 Page 11 of 42 TS Table 4.3-6 indicates that the frequency of the required Channel Checks and Channel Calibrations is in accordance with the Surveillance Frequency Control Program (SFCP). The table also excludes neutron indications (logarithmic neutron channel and startup channel) and the reactor trip breaker indication from Channel Calibration. With the removal of this table from the TSs, Entergy proposes SR 3.3.3.5 be modified to address both the test frequencies and the stated exclusions as follows:

Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of athe CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6in accordance with the Surveillance Frequency Control Program. The logarithmic neutron instrumentation, the startup channel instrumentation, and the reactor trip breaker indication are excluded from CHANNEL CALIBRATION.

RSD capability is described in ANO-2 SAR Sections 3.1.2, 7.4.1.5, and 15.1.26. An illustration of the RSD panel is included in SAR Figure 7.4-3.

The definition of "operable" in ANO-2 TS states that a system shall be operable or have operability when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system to perform its specified safety function(s) are also capable of performing their related support function. This definition provides adequate guidance for determining what instrumentation and controls are necessary for a given RSD function. Therefore, listing specific instrumentation and controls within the TS proper is unnecessary and may lead to needless expenditure of licensee and NRC resources processing license amendments to revise the RSD instrumentation details contained in the TS. These details are not necessary to describe the actual regulatory requirements and, therefore, the details may be removed from the TSs without a significant impact to nuclear or public safety. The list of variables (instruments) will be placed in the associated TS Bases as an Operator aid. This is acceptable because the TS Bases are consistent with the design basis as described in the SAR. Changes to the TS Bases are evaluated in accordance with TS 6.5.14 and 10 CFR 50.59, as applicable. In addition, precedent for the removal of the list of RSD instrumentation from the TSs has been established via TSTF-266 (Reference 7). The absence of any listing of RSD instruments within the TS proper is also consistent with NUREG-1432.

The TS definition of operability and the TS Bases control program of TS 6.5.14 provides adequate assurance that credited RSD functions will be appropriately controlled.

TS 3.4.6.2 (SR 4.4.6.2.1.b) - Reactor Coolant System Operational Leakage The ANO-2 reactor vessel is illustrated in SAR Figure 5.4-1. The reactor vessel is a right circular cylinder with two hemispherical heads. The lower head is welded to the vessel shell; the upper closure head can be removed to provide access to the reactor internals. Sealing is accomplished by using two silver-plated, NiCrFe alloy, self-energized O-rings. The space between the two rings is monitored to detect any inner-ring coolant leakage. An increase in the flange leakoff temperature indicates that reactor coolant is leaking past the inner O-ring to the Reactor Drain Tank (RDT).

Enclosure to 2CAN121901 Page 12 of 42 LCO 3.4.6.2 places limits on RCS operational leakage. SR 4.4.6.2.1.b requires monitoring the reactor head flange leakoff temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Note that there is no TS Action associated with the failure to monitor this leak path (i.e., instrument failure). This could lead to the potential entry into LCO 3.0.3 because in accordance with SR 4.0.1, failure to perform a SR equates to failure to meet the subject LCO. LCO 3.0.3 is a catch-all action to address plant conditions that are not specifically described in any other TS action statement.

LCO 3.0.3 requires a shutdown of the unit.

RCS leakage is continuously monitored by leak detection instruments defined in ANO-2 LCO 3.4.6.1, "Leakage Detection Systems." These systems include a containment atmosphere particulate radioactivity monitor, a containment atmosphere gaseous radioactivity monitor, and the containment sump level monitor. Non-TS leak detection indications displayed in the Control Room include containment humidity, temperature, and pressure, along with Quench Tank and RDT level indications. In addition to leak detection instruments, LCO 3.4.6.2 requires RCS leakage to be verified by performance of an RCS inventory balance. The RCS inventory balance is normally performed once per day and is required to be performed once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in accordance with the current ANO-2 SFCP.

Operation's logs record the RVH leakoff temperature daily along with readings from the TS-required leak detection instruments and other non-TS installed instrumentation. Any potential increase in RCS leakage detected by one or more of these devices is evaluated by performance of an RCS inventory balance calculation (Reference SR 4.4.6.2.1.a). Performance of an RCS inventory balance is the most accurate means of detecting and quantifying RCS leakage.

The subject SR does not exist in NUREG-1432. The RVH flange leakoff temperature instrument does not meet any of the 10 CFR 50.36(c)(2)(ii) criteria for inclusion in the TSs:

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The RVH flange leakoff temperature indication does not detect leakage from the RCPB. The subject instrument detects only leakage from the RVH inner O-ring. The RCPB consists of both the inner and outer RVH O-rings. Therefore, this specification does not meet Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Neither the subject instrument or its output is an initial condition to any accident or transient analysis described in the SAR. Therefore, this specification does not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The subject instrument does not function or actuate to perform any accident or transient mitigation function described in the SAR. Therefore, this specification does not meet Criterion 3.

Enclosure to 2CAN121901 Page 13 of 42 Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The subject instrument monitors only the inner RVH O-ring and not the RCS pressure boundary. Therefore, this instrument is not significant to the health and safety of the public.

This SR does not involve an SSC requiring risk review/unavailability monitoring. Therefore, this specification does not meet Criterion 4.

Surveillance requirements are required to be included in the TSs in accordance with 10 CFR 50.36(c)(3), which states that SRs 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 limiting conditions for operation will be met. The subject SR does not provide a significant improvement in the assessment of RCS leakage in comparison to the RCS inventory balance and other leak detection instrumentation.

The leakage detection function is maintained without the use of the RVH flange leakoff temperature indication.

Deletion of the subject SR will not have a significant adverse impact on the ability of leakage detection instrumentation to detect RCS leakage and, therefore, has no significant adverse impact to nuclear or public safety. Based on the above, Entergy requests deletion of SR 4.4.6.2.1.b from the TSs. This change is consistent with NUREG 1432.

TS 3.7.2.1 - Steam Generator Pressure/Temperature Limitation In a PWR, the SGs provide the interface between the RCS (primary) and the main steam system (secondary). The SGs are vertical U-tube heat exchangers in which heat is transferred from the RCS to the main steam system. The RCS is a closed system preventing the release of radioactive materials into the Containment Building.

The limitation on SG pressure and temperature ensures that the pressure-induced stresses in the SGs do not exceed the maximum allowable fracture toughness stress limits. The limitation that both the primary and secondary water temperatures must be above 90 °F when the pressure of either coolant in the SG is > 275 psig is based on a SG reference temperature for nil ductility transition (RTNDT) of 30 F, which is intended to prevent brittle fracture.

Entergy requests TS 3.7.2.1 be relocated to the ANO-2 TRM. Associated TS bases will also be relocated to the TRM.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The limitations imposed by TS 3.7.2.1 on SG temperature and pressure are provided to prevent brittle fracture in the SGs and do not describe a limitation on instrumentation that is used to detect, and indicate in the Control Room, an abnormal degradation of the RCPB.

Therefore, this specification does not meet Criterion 1.

Enclosure to 2CAN121901 Page 14 of 42 Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The temperature and pressure values described in TS 3.7.2.1 do not provide direct input to RPS or ESFAS functions, nor do the limits represent process variable, design feature, or operating restriction that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Therefore, this specification does not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.7.2.1 does not involve an SSC that is part of the primary success path (heat removal) which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The limits apply to conditions when the RCS temperature is low (Mode 5 or lower). Under these conditions, the SGs are not required to function to mitigate any DBA or transient. Therefore, this specification does not meet Criterion 3.

Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The TS 3.7.2.1 limits have not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS does not establish limits requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

TS 3.7.2.1 does not meet any of the criterion in 10 CFR 50.36 for retention. The proposed relocation of the subject requirements to the TRM will facilitate future changes to these requirements without obtaining NRC approval. Any changes to the requirements contained in the TRM will require a 10 CFR 50.59 review. This proposed change is consistent with NUREG-1432.

TS 3.7.5.1 - Flood Protection ANO-2 safety-related facilities are designed to withstand the effects of hydrostatic pressures, buoyancy, and wave action under Probable Maximum Flood (PMF) conditions. Category 1 systems and equipment susceptible to the effects of flooding are either located a specified number of feet above mean sea level to protect against splash, or are protected up to the PMF by: (1) wall thicknesses in Seismic Category 1 structures, (2) potential flood paths through construction joints are sealed by use of waterstops or other seal materials, (3) the number of openings in walls and slabs below flood level was kept to a minimum, and (4) watertight doors, watertight equipment hatches, and flood seals.

As discussed in ANO-2 SAR Section 2.4.14, a highly unusual occurrence of flood levels approaching the ground elevation of 354 feet mean sea level would take from two days to several weeks to develop. This would allow ample lead time to perform necessary emergency actions for all accesses which need to be protected. These actions are implemented in accordance with abnormal operating procedures for natural emergencies.

Enclosure to 2CAN121901 Page 15 of 42 LCO 3.7.5.1, "Flood Protection," requires flood protection to be provided for all safety-related SSCs when the water level of the Dardanelle Reservoir exceeds 350 feet mean sea level United States Geological Survey datum, at the intake structure. Entergy requests that the requirements of TS 3.7.5.1 be relocated to the ANO-2 TRM. Associated TS bases will also be relocated.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The limitations imposed by TS requirements related to flood protection are provided to ensure that facility protective actions will be taken in the event of forecast flood conditions and do not describe a limitation on instrumentation that is used to detect, and indicate in the Control Room, an abnormal degradation of the RCPB. This specification does not meet Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The limitations imposed are related only to flood protection and do not provide direct input to RPS or ESFAS functions, nor do the limits represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

This specification does not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.7.5.1 does not involve an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 3.

Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The limits established by TS 3.7.5.1 have not been shown to be risk significant to public health and safety by either operating experience or PRA. TS 3.7.5.1 only specifies a flood elevation in which preemptive actions are to be implemented such that flood protection is established prior to levels reaching the ground level of 354 feet mean sea level. While specific flood protection SSCs included in the ANO-2 design may be assessed in the ANO-2 PRA model, the subject TS does not directly involve an SSC requiring risk review/unavailability monitoring. Therefore, this specification does not meet Criterion 4.

Enclosure to 2CAN121901 Page 16 of 42 TS 3.7.5.1 does not fulfill any of the 10 CFR 50.36c(2)(ii) criteria for which TSs must be established. Relocating this requirement to the ANO-2 TRM simplifies the ANO-2 TS and will facilitate future changes to these requirements without obtaining NRC approval. Any changes to the requirements contained in the TRM will require a 10 CFR 50.59 review. This proposed change is consistent with NUREG-1432.

TS 3.7.9.1 - Sealed Source Contamination Sealed source contamination occurs when the source material sealed inside a container leaks out of the container and contaminates the surface of the container. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c). This limitation is to ensure that leakage from byproduct, source, and special nuclear material will not exceed allowable intake values.

LCO 3.7.9.1, "Sealed Source Contamination," requires each sealed source containing radioactive material either in excess of 100 microcuries (µCi) of beta and/or gamma emitting material or 5 µCi of alpha emitting material shall be free of 0.005 µCi of removable contamination. Because these limits are controlled by other regulation, Entergy requests deletion of TS 3.7.9.1.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The limitations imposed by TS requirements related to sealed source contamination are provided to ensure that leakage from byproduct, source, and special nuclear material will not exceed allowable intake values and do not describe a limitation on instrumentation that is used to detect, and indicate in the Control Room, an abnormal degradation of the RCPB.

This specification does not meet Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The limitations imposed are related only to sealed source contamination and do not provide direct input to RPS or ESFAS functions, nor do the limits represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 2.

Enclosure to 2CAN121901 Page 17 of 42 Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.7.9.1 does not involve an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 3.

Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The limits established by TS 3.7.9.1 have not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS does not involve an SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

TS 3.7.9.1 does not fulfill any of the 10 CFR 50.36(c)(2)(ii) criteria on items for which TSs must be established. The subject limits are controlled by other regulation (10 CFR 70.39 and 10 CFR 31.5). Entergy radiation protection procedures provide appropriate controls on all potential sources of contamination, including sealed sources. Entergy procedure EN-RP-143, "Source Control," establishes the controls necessary for compliance with TS 3.7.9.1 with respect to the 0.005 µCi limit along with leak detection methods and frequency. Although TS 3.7.9.1 requires only an annual report, Entergy procedure EN-RP-102, "Radiological Control," requires reporting of removable contamination of 0.005 µCi within 30 days in accordance with 10 CFR 31.5(c)(5). This proposed change is consistent with NUREG-1432.

Therefore, Entergy proposes the deletion of TS 3.7.9.1.

TS 3.7.12 - Spent Fuel Pool Structural Integrity Safe storage is provided for approximately one core and eleven reload batches of spent fuel assemblies (including Control Element Assemblies (CEAs)), new fuel stored under water during core reloading, and the spent fuel shipping cask. Maintaining structural integrity in the Spent Fuel Pool (SFP) is needed to ensure safe storage. The stainless steel lined, reinforced concrete SFP provides storage for 988 fuel assemblies. Spent fuel assemblies are stored in vertical racks. Adequate spacing is provided to preclude criticality with no credit for the borated pool water. The massive reinforced concrete SFP is monolithically tied to the remaining part of the Auxiliary Building. The SFP walls include reinforced steel which limits concrete cracking during expected thermal stresses.

LCO 3.7.12, "Spent Fuel Pool Structural Integrity," requires that the structural integrity of the SFP be maintained in accordance with SR 4.7.12. SR 4.7.12 has two requirements:

1) inspection frequencies and, 2) inspection acceptance criteria. If the requirements of this specification are not met, then a special report is to be submitted to the NRC within 30 days.

Entergy proposes that this TS and associated TS Bases be relocated to the TRM.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Enclosure to 2CAN121901 Page 18 of 42 Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The limitations imposed by TS requirements related to the structural integrity of the SFP are provided to ensure that the SFP remains safe for use and will adequately resist the imposed loadings. The SFP structural integrity limitations do not describe a limitation on instrumentation that is used to detect, and indicate in the Control Room, an abnormal degradation of the RCPB. This specification does not meet Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The limitations imposed are related to SFP structural inspections and acceptance criteria.

The limits do not provide direct input to RPS or ESFAS functions, nor do the limits represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.7.12 does not involve an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 3.

Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The limits established by TS 3.7.12 have not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS does not involve an SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

TS 3.7.5.1 does not fulfill any of the 10 CFR 50.36c(2)(ii) criteria for which TSs must be established. Relocating this requirement to the ANO-2 TRM simplifies the ANO-2 TS and will facilitate future changes to these requirements without obtaining NRC approval. Any changes to the requirements contained in the TRM will require a 10 CFR 50.59 review. This proposed change is consistent with NUREG-1432.

TS 3.9.3.a - Decay Time and Spent Fuel Storage The minimum time requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) 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 and the resulting dose calculations using the Alternative Source Term described in RG 1.183, "Alternative Radiological Source Terms for

Enclosure to 2CAN121901 Page 19 of 42 Evaluating Design Basis Accidents at Nuclear Power Reactors," (Reference 16). The likelihood of a fuel handling accident is minimized by administrative controls and physical limitations imposed on fuel handling operations. All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a qualified supervisor.

LCO 3.9.3.a requires the reactor to be subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months before the commencement of core offload. Entergy proposes to relocate TS 3.9.3.a to the TRM.

Associated TS bases will also be relocated.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The subject decay time limit does not involve installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

Therefore, this specification does not satisfy Criterion 1 for retention in the TS.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

During the development of NUREG-1432 it was determined that this specification could be relocated. The basis for this determination was that existing scheduling restraints associated with moving irradiated fuel following a plant shutdown will prevent the decay time limit from being exceeded. These activities include containment entry, removal of the reactor vessel head and upper internals, and filling the refueling canal.

The decay time assumed for the ANO-2 fuel handling accident (FHA) is 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . The scheduling constraints discussed above are applicable to ANO-2 and ensure that fuel decay time assumed in the FHA is met. 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 six full days of required 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.

Therefore, it is unnecessary to maintain the decay time limit in the TSs to ensure the plant is operated within the bounds of the FHA design basis analysis due to physical plant limitations.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.9.3.a does not involve an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this specification does not satisfy Criterion 3 for retention in the TS.

Enclosure to 2CAN121901 Page 20 of 42 Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

TS 3.9.3.a does not involve an SSC which operating experience or PRA has shown to be significant to the public health and safety. Therefore, this specification does not satisfy Criterion 4 for retention in the TS.

TS 3.9.3.a is not required to ensure the plant is operated within the bounds of the FHA design basis analysis. This conclusion is supported by the absence of operability and surveillance requirements for the fuel decay time in the STS. Accordingly, this proposed change conforms to the STS, and fuel decay time requirements may be relocated to the TRM. Relocating the decay time requirements from the TS to the TRM will eliminate the burden of processing license amendments when future changes are made to the decay time requirements and will facilitate the more effective utilization of NRC and Entergy resources. Future changes to fuel decay time requirements in the TRM will be subject to the controls of 10 CFR 50.59.

TS 3.9.5 - Communications Voice and non-voice systems are provided for reliable communication during plant startup, operation, shutdown, and maintenance under normal conditions. Direct communication between the control room and the refueling machine console is available whenever changes in core geometry are taking place. This provision allows the Control Room Operator and refueling machine operator in the Containment Building to inform one another of any impending unsafe condition during fuel movement. 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 core alterations.

LCO 3.9.5, "Communications," requires direct communications to be maintained between the Control Room and personnel at the refueling station during core alterations. Entergy proposes that this TS be deleted. Establishing communications is an administrative requirement governed by procedures established in accordance with 10 CFR 50, Appendix B. The ANO-2 SAR does not specifically address communication as described above and does not include communication in the monitoring instrumentation needed to alert personnel of conditions that would be of concern such as degraded ability to remove decay heat or excessive radiation levels.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The communications requirement of TS 3.9.5 is provided to ensure that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during core alterations and does not describe a limitation on instrumentation that is used to detect, and indicate in the Control Room, an abnormal degradation of the RCPB. This specification does not meet Criterion 1.

Enclosure to 2CAN121901 Page 21 of 42 Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Communications do not provide direct input to RPS or ESFAS functions, nor does this requirement represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.9.5 does not involve an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 3.

Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The communications requirement of TS 3.9.5 has not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS does not involve an SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

This TS does not fulfill any of the 10 CFR 50.36c(2)(ii) criteria for which TSs must be established. Therefore, Entergy proposes deletion of TS 4.9.5. This proposed change is consistent with NUREG-1432.

TS 3.9.6 - Refueling Machine Operability The refueling machine is a traveling bridge and trolley which is located above the refueling canal and rides on rails set in the concrete on each side of the refueling canal. Motors on the bridge and trolley position the machine over each fuel assembly location within the reactor core or fuel transfer carrier. The hoist assembly contains a grappling device which, when rotated by the actuator mechanism, engages the fuel assembly to be removed. The hoist assembly and grappling device are raised and lowered by a cable attached to the hoist winch. After the fuel assembly has been raised into the refueling machine, the refueling machine transports the fuel assembly to its designated location.

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 an SSC is inadvertently engaged during lifting operations.

Enclosure to 2CAN121901 Page 22 of 42 LCO 3.9.6, "Refueling Machine Operability," requires that the refueling machine has:

1) a minimum capacity of 3750 pounds, 2) an overload cut off limit of 100 pounds plus the combined weight of one fuel assembly, one CEA, and the grapple in the "fuel only" region, and

3) an overload cut off limit of 100 pounds plus the combined weight of one fuel assembly, one CEA, the grapple, and the hoist box in the "fuel plus hoist box" region.

TS 3.9.6 does not fulfill any of the 10 CFR 50.36(c)(2)(ii) criteria for which TSs must be established. Entergy requests TS 3.9.6 requirements be relocated to the TRM.

Although the refueling machine design includes interlocks that can prevent damage to the refueling equipment and fuel assemblies, they are not assumed to function to mitigate the consequences of a design basis accident. No actual plant equipment or accident analyses will be affected by the proposed change and no failure modes not bounded by previously evaluated accidents will be created.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The limitations imposed by this specification on the refueling machine are provided to ensure that that the refueling machine has sufficient load capacity to lift a fuel assembly and that the core internals and pressure vessel are protected from excessive lifting force in the event an SSC is inadvertently engaged during lifting operations. These TS requirements do not describe a limitation on instrumentation that is used to detect, and indicate in the Control Room, an abnormal degradation of the RCPB. These specifications do not meet Criterion 1 Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The subject refueling machine limits do not provide direct input to RPS or ESFAS functions, nor does this requirement represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.9.6 does not involve an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 3.

Enclosure to 2CAN121901 Page 23 of 42 Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The limits subject to TS 3.9.6 have not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS does not involve an SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

This TS does not fulfill any of the 10 CFR 50.36c(2)(ii) criteria for which TSs must be established. Relocating these requirements to the TRM simplifies the ANO-2 TSs and will facilitate future changes to these requirements without obtaining NRC approval. Any changes to these requirements will require a 10 CFR 50.59 review. This proposed change is consistent with NUREG-1432.

TS 3.9.7 - Crane Travel - Spent Fuel Pool Building A 130-ton capacity single failure proof bridge crane is provided to transport the new and spent fuel shipping casks from the station train bay into the ANO-1 and ANO-2 Auxiliary Buildings.

Auxiliary 15-ton and 4-ton capacity hoists are also provided on the crane.

LCO 3.9.7, "Crane Travel - Spent Fuel Pool Building," states that loads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the SFP with fuel assemblies in the SFP. TS 3.9.7 does not fulfill any of the 10 CFR 50.36(c)(2)(ii) criteria on items for which TSs must be established. Entergy proposes to relocate TS 3.9.7 to the TRM.

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.

RG 1.13, "Spent Fuel Storage Facility Design Basis," Revision 1 (Reference 17), Section C.3 states, in part:

"Interlocks should be provided to prevent cranes from passing over stored fuel (or near stored fuel in a manner such that if a crane failed, the load could tip over on stored fuel) when fuel handling is not in progress. During fuel handling operations, the interlocks may be bypassed and administrative control used to prevent the crane from carrying loads that are not necessary for fuel handling over the stored fuel or other prohibited areas."

The layout of the fuel handling area in the Auxiliary Building is such that a spent fuel cask is never required to transverse the SFP during removal of spent fuel elements. Diverse electrical interlocks (a limit switch and a power disconnect from the main contact rails) are provided to the fuel handling crane to prevent an inadvertent transverse of the pool.

The Final Policy Statement encouraged licensees to implement a voluntary update program of station TSs to be consistent with the STS. The four 10 CFR 50.36 criteria provide a basis for relocating requirements from the TSs to other licensee-controlled documents, provided the requirements meet none of the four criteria.

Enclosure to 2CAN121901 Page 24 of 42 Criterion 1 Installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the RCPB.

The restrictions on the weight and travel of crane loads over irradiated spent fuel are not related to any installed instrumentation that is used to detect RCPB degradation or to indicate such degradation in the Control Room. Therefore, TS 3.9.7 does not meet Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The subject refueling machine limits do not provide direct input to RPS or ESFAS functions, nor does this requirement represent a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not meet Criterion 2.

TS 3.9.7 is associated with the fuel handling accident (FHA) described in SAR Section 15.1.23. The TS 3.9.7 crane load limit is an operating restriction that provides defense-in-depth against a heavy load drop over irradiated fuel. The FHA analysis does not assume a load drop of > 2000 lbs over fuel stored in the SFP (Reference 18). The FHA analysis assumes a load drop equivalent to the weight of one fuel assembly plus the weight of the hoist grapple and one CEA. The FHA analysis considered a non-fuel load drop of 2000 lbs in order to illustrate that the drop of a fuel assembly remained bounding with respect to the amount of fuel damaged and any subsequent radioactive release. Therefore, the crane load limit is not considered an initial condition of the FHA.

The TS load limit restriction, in conjunction with other non-TS requirements that restrict crane operation (such as interlocks and physical stops, operator training, and load handling procedures) provide a defense-in-depth approach to handling heavy loads in the vicinity of the SFP. The crane load limit and the other established defense-in-depth features are not considered an initial condition for the FHA and, therefore, do not meet Criterion 2.

Criterion 3 An SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

As discussed under Criterion 2, the crane load limit addressed by TS 3.9.7 is not an initial condition of the FHA analysis and is not required to mitigate a FHA or any other design basis accident or transient relating to fission product barrier integrity. Therefore, TS 3.9.7 does not meet Criterion 3.

Criterion 4 An SSC which operating experience or PRA has shown to be significant to public health and safety.

The limits subject to TS 3.9.7 have not been shown to be risk significant to public health and safety by either operating experience or PRA. This TS does not involve an SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4.

Enclosure to 2CAN121901 Page 25 of 42 This TS does not fulfill any of the 10 CFR 50.36c(2)(ii) criteria for which TSs must be established. Relocating this requirement to the TRM simplifies ANO-2 TSs and will facilitate future changes to these requirements without obtaining NRC approval. Any changes to these requirements will require a 10CFR50.59 review. This proposed change is consistent with NUREG-1432.

TS 3.11.1 - Liquid Holdup Tanks TS 3.11.2 - Gas Storage Tanks TS 3.11.3 - Explosive Gas Mixture TS 6.5.8 - Explosive Gas and Storage Tank Radioactivity Monitoring Program (new)

Due to being related to limiting potential radiological exposures to the public, the technical evaluations of the above subject TSs are grouped for discussion. TSs 3.11.1 and 3.11.2 involve limits on the quantity of radioactive material in a storage tank. TS 3.11.3 places limits on the oxygen and hydrogen content within a gaseous radwaste (GRW) storage tank to preclude potential explosive mixtures that could result in an offsite release. Entergy is also proposing adoption of a new TS program that replaces the former three TSs.

The following is a summary description of each TS proposed for deletion in favor of the new program:

1.

TS 3.11.1, "Liquid Holdup Tanks," limits the radioactive content of temporary outdoor liquid storage tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents or those which do not have overflows and surrounding area drains connected to the liquid radwaste (LRW) treatment system to 10 curies.

2.

TS 3.11.2, "Gas Storage Tanks," limits the radioactive content of GRW storage tanks to no more than 82,400 curies noble gases (Xe-133).

3.

If either of the above limits are exceeded, both TS 3.11.1 and 3.11.2 require a discussion to be included in the next annual Radioactive Effluent Release Report (RERR).

4.

Verification that TS 3.11.1 and 3.11.2 tank contents remain within limits is in accordance with the SFCP.

5.

TS 3.11.3, "Explosive Gas Mixture," requires the oxygen/hydrogen content within a GRW tank to be less than the illustrated in TS Figure 3.11-3. Monitoring of these parameters is required whenever the GRW system is in service (i.e., when gases are being added to a GRW tank).

The limit placed on temporary outdoor liquid storage tanks is established such that, in the event of an uncontrolled release of the contents of the tanks, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area. Note that ANO-2 does not currently have any outdoor LRW tanks, whether permanent or temporary.

Enclosure to 2CAN121901 Page 26 of 42 The limit placed on GRW storage tanks is established such that, in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a member of the public at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Branch Technical Position ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure" (Reference 19). TS 3.11.3 provides defense-in-depth by controlling the oxygen/hydrogen mixture in a GRW tank, thus limiting the potential of a radioactive gaseous release due to tank rupture.

The overall scope of the above requirements can be met by adoption of a new proposed TS entitled, "Explosive Gas and Storage Tank Radioactivity Monitoring Program" (EGSTRMP).

Entergy proposes this program to be added as TS 6.5.8, which is currently an unused number within the Administrative Controls section of the ANO-2 TSs. The program is contained in the STS under TS 5.5.12. This program is already established at ANO because the program currently exists in ANO-1 TS 5.5.12. Implementation of the program is provided through application of the ANO Offsite Dose Calculation Manual (ODCM) and supporting procedures and reports.

The EGSTRMP provides controls for potentially explosive gas mixtures contained in the ANO GRW tanks, the quantity of radioactivity contained in the GRW storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The program requires that the gaseous radioactivity quantities in the GRW storage tanks to be determined following the methodology in Branch Technical Position ETSB 11-5. ANO-2 SAR Section 15.1.16.2 describes the analysis of GRW storage:

Assumptions and methods used in this analysis are consistent with those of Branch Technical Position ETSB 11-5. The quantity of radioactivity contained in a single tank has been limited to a curie value which will prevent a member of the public at the exclusion area boundary from receiving a total body exposure exceeding 0.5 Rem in a 2-hour period. This is consistent with NUREG-0800, BTP ETSB 11-5, July 1981.

In addition to the above, the new program establishes three requirements. Each is discussed below.

1.

The limits for concentrations of hydrogen and oxygen in the Waste Gas System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion).

TS 3.11.3 currently includes controls that meet Item 1. Generic Letter (GL) 95-10, "Relocation of Selected Technical Specifications Requirements Related to Instrumentation" (Reference 20), the NRC stated with respect to explosive gas monitoring instrumentation:

The actions required by existing TSs typically require alternate sampling, limited operation of the gaseous waste system, and submittal of a special report if the explosive gas monitoring instrumentation does not conform to the limiting condition for operation. The explosive gas monitoring instrumentation requirements address detection of possible precursors to the failure of a waste gas system but do not prevent or mitigate design basis accidents or transients which assume a failure of or

Enclosure to 2CAN121901 Page 27 of 42 present a challenge to a fission product barrier. Acceptable concentrations of explosive gases are actually controlled by other limiting conditions for operation (e.g.,

Gaseous Effluents, Explosive Gas Mixture) or by programs described in the "Administrative Controls" section of TSs. The requirements related to explosive gas monitoring instrumentation do not conform to the 10 CFR 50.36 criteria for inclusion in the TSs. Therefore, licensees may propose to relocate the explosive gas monitoring instrumentation requirements to the UFSAR and control changes to those provisions in accordance with 10 CFR 50.59.

The installed waste gas analyzer system and backup sampling methods are described in SAR Section 11.3.3. Explosive mixture control is established in procedure OP-2607.018, "Waste Gas Sampling and Analyzer Operation," which includes, with added detail, the TS Figure 3.11-1 hydrogen-oxygen limits graph. Changes to both the SAR and this procedure are governed by 10 CFR 50.59. Based on the above, the deletion of TS 3.11.3 in favor of the proposed new TS 6.5.8 EGSTRMP meets the intent of GL 95-10 and is consistent with the STS.

2.

A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

LCO 3.11.2 requires that the quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 82,400 curies noble gases (considered as Xe-133). The second EGSTRMP requirement stated above does not contain a specific curie limit for the affected tanks. Instead, the potential dose impact to a member of the public is limited to < 0.5 rem which meets the intent of the LCO 3.11.2 curie limit. This is supported by the associated ANO-2 TS 3.11.2 Bases which states:

The limit placed on GRW storage tanks is established such that, in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a member of the public at the nearest exclusion area boundary will not exceed 0.5 rem.

TS 3.11.2 includes an Action to restore the tank contents to within the limit should the limit be exceeded. ODCM Limitation (L) 2.4.1, Required Action A.1, requires action to be initiated immediately to restore tank contents to within limits. A description of any event that involved the limit being exceeded must also be included in the Radioactive Effluent Release Report (RERR) required annually by ANO-2 TS 6.6.3. Step 5.2 of the ODCM includes a reference for submitting this information within the annual RERR.

TS 3.11.2 also contains a surveillance requirement to periodically verify that the contents of the tanks are within the limit. As stated previously, the ODCM and related implementing procedures along with automated analysis reports provide for sampling and ensuring tank contents are verified and limits maintained.

The ODCM is required by ANO-2 TS 6.5.1 and is submitted annually as part of the TS 6.6.3 required RERR (Reference 21). ANO-2 TS 6.5.4, "Radioactive Effluent Controls Program," also establishes related requirements, including the monitoring, sampling, and analysis of radioactive liquid and gaseous effluents. These and other TS programmatic controls were established following issuance of GL 89-01,

Enclosure to 2CAN121901 Page 28 of 42 "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program" (Reference 22).

The ANO-2 TS programs noted above along with the proposed adoption of the EGSTRMP provide the necessary administrative controls to ensure plant procedures and programs are maintained to monitor and control the curie content of GRW storage.

In turn, TS 3.11.2 is proposed for deletion in favor of the adoption of the EGSTRMP.

This change is consistent with the STS.

3.

A surveillance program to ensure that the quantity of radioactivity contained in all temporary outdoor liquid radwaste tanks: 1) that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents; and 2) that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations equal to the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

LCO 3.11.1 requires that the quantity of radioactivity contained in each unprotected outside temporary LRW storage tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. The third EGSTRMP requirement stated above does not contain a specific curie limit for the affected tanks.

Instead, the potential dose impact to a member of the public is limited to that described in 10 CFR 20, Appendix B, Table 2, Column 2, which meets the intent of the LCO 3.11.1 curie limit. This is supported by the associated ANO-2 TS 3.11.2 Bases which states:

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that, in the event of an uncontrolled release of the contents of the tanks, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

TS 3.11.1 includes an Action to discontinue additions to the affected tank and to restore the tank contents to within the limit should the limit be exceeded. ODCM L 2.3.1, Required Action A.1, requires action to be initiated immediately to restore tank contents to within limits. A description of any event that involved the limit being exceeded must also be included in the Radioactive Effluent Release Report (RERR) required annually by ANO-2 TS 6.6.3. Step 5.2 of the ODCM includes reference for submitting this information within the annual RERR.

TS 3.11.1 also contains a surveillance requirement to periodically verify that the contents of the tanks are within the limit. As stated previously, the ODCM and related implementing procedures along with automated analysis reports provide for sampling and ensuring tank contents are verified and limits maintained. The ODCM states, in part:

This Limitation is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II.

Enclosure to 2CAN121901 Page 29 of 42 Procedure OP-1042.003 includes the sampling requirements and frequency for the subject tanks; note, however, that ANO does not currently have any outdoor unprotected LRW storage tanks on site.

This program provision provides adequate regulatory assurance that the quantity of radioactivity in the affected tanks will be controlled consistent with NRC approved standards and the dose consequences of an unplanned release restricted to within limits dictated by federal regulations. Consistent with the conclusions in GL 95-10, the limit for acceptable concentrations of radioactive material will be controlled by the proposed EGSTRMP along with other aforementioned TS programs associated with radiological controls. Therefore, Entergy proposes that the TS 3.11.1 requirements be deleted in favor of the proposed new EGSTRMP.

The net effect of the proposed changes is to provide adequate regulatory control in the TS while making the content of the ANO-2 TS more consistent with the STS and simplifying the ANO-2 TS consistent with the goals of the NRC's Final Policy Statement.

In adopting the new EGSTRMP, the following differences are proposed for the ANO-2 program when compared to the program wording contained in the STS:

The bracketed information [Waste Gas Holdup System], is changed to "Waste Gas Decay Tanks," consistent with ANO-2 terminology. This difference does not affect the intent or requirements of the EGSTRMP.

The bracketed information [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks], is changed to "the quantity of radioactivity contained in the Waste Gas Decay Tanks, and the quantity of radioactivity contained in unprotected temporary outdoor liquid storage tanks," consistent with ANO-2 terminology.

This difference does not affect the intent or requirements of the EGSTRMP. This includes the bracketed statement in sub-bullet "b.", [each gas storage tank and fed into the offgas treatment system], which is changed to "each Waste Gas Decay Tank."

The bracketed information [Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures"] is changed to "the ODCM." As stated previously, the ODCM and related implementing procedures control the quantity of a radioactive material in LRW tanks. The ODCM ensures that radioactive quantities of the associated liquid storage tanks will be limited to that stated 10 CFR 20, Appendix B, Table 2, Column 2, consistent with Section 15.7.3 of NUREG-0800, "Standard Review Plan," with states:

Tanks and associated components containing radioactive liquids outside containment are acceptable if failure does not result in radionuclide concentrations in excess of the limits in 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply,* in an unrestricted area, or if special design features are provided to mitigate the effects of postulated failures for systems not meeting these limits.

This is also stated in the third provision of the proposed EGSTRMP. Therefore, this difference maintains consistency with the EGSTRMP, the STS, and NUREG-0800.

Enclosure to 2CAN121901 Page 30 of 42 The references to STS SR 3.0.2 and SR 3.0.3 were revised to the corresponding ANO-2 specifications 4.0.2 and 4.0.3. These ANO-2 specifications are consistent with the wording or intent of the SR 3.0.2 and SR 3.0.3 versions contained in the STS.

These described differences do not adversely affect the intent of the STS EGSTRMP program.

The described differences provide the correct ANO-2 component nomenclature and references for the specifications. As such, the resulting ANO-2 specific program effectively retains the current licensing and design bases requirements for the affected ANO-2 TS, and provides the requirements necessary to assure that the explosive gas and radioactivity content of the subject tanks are monitored and controlled within acceptable limits.

Conclusion Entergy has concluded that the subject changes are appropriately justified and that adequate assurance that compliance with regulatory requirements will be effectively maintained.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance documents are applicable to the proposed change.

10 CFR 50, Appendix A, General Design Criterion (GDC) 2, " Design Bases for Protection Against Natural Phenomena," states:

Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) appropriate consideration of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

GDC 13, "Instrumentation and Control," states:

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

Enclosure to 2CAN121901 Page 31 of 42 GDC 14, "Reactor Coolant Pressure Boundary," states:

The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure and of gross rupture.

GDC 19, "Control Room," states:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including LOCAs. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposure in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided: (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

GDC 25, "Protection System Requirements for Reactivity Control Malfunctions," states:

The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

GDC 26, "Reactivity Control System Redundancy and Capability," states:

Two independent reactivity control systems of different design principles shall be provided.

One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operation occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

GDC 32, "Inspection of Reactor Coolant Pressure Boundary (RCPB)," states:

Components which are part of the RCPB shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leak tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

GDC 60, "Control of Releases of Radioactivity Materials to the Environment," states:

The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences.

Enclosure to 2CAN121901 Page 32 of 42 Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.

GDC 64, "Monitoring Radioactivity Releases," states:

Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of LOCA fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

Regulatory Guide (RG) 1.13, "Spent Fuel Storage Facility Design Basis," Revision 1, states:

Interlocks should be provided to prevent cranes from passing over stored fuel (or near stored fuel in a manner such that if a crane failed, the load could tip over on stored fuel) when fuel handling is not in progress. During fuel handling operations, the interlocks may be bypassed and administrative control used to prevent the crane from carrying loads that are not necessary for fuel handling over the stored fuel or other prohibited areas RG 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plants. This includes indications of plant variables required by the Control Room operating personnel during accident situations to (1) provide information required to permit the operator to take pre-planned manual actions to accomplish safe plant shutdown; (2) determine whether the reactor trip, engineered safety feature systems, and manually initiated safety systems and other systems important to safety are performing their intended functions (i.e.. reactivity control, core cooling, maintaining reactor coolant system.

integrity, and maintaining containment integrity); and (3) provide information to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary, and containment) and to determine if a gross breach of a barrier has occurred. RG 1.97 also includes indications of plant variables that provide information on operation of plant safety systems and other systems important to safety that are required by the control room operating personnel during an accident to (1) furnish data regarding the operation of plant systems in order that the operator can make appropriate decisions as to their use and (2) provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.

NRC Branch Technical Position ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure," states:

The purpose of this BTP is to provide guidelines on postulated radioactive releases due to a radioactive waste gas system leak or failure. The goal is to minimize potential radiation exposures to workers and the public, and to provide reasonable assurance that the radiological consequences of a single failure of an active component in the waste gas system would not result in exceeding the guidelines of 10 CFR Part 20 for a unique unplanned release and would, therefore, be substantially below the guidelines of 10 CFR Part 100 for a postulated event.

Enclosure to 2CAN121901 Page 33 of 42 The proposed changes do not affect compliance with these regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met. 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 STS [(e.g.,

TSTF-567)], even if the licensee does not adopt all STS improvements. The Commission encourages all licensees who submit Technical Specification related submittals based on this Policy Statement to emphasize human factors principles.

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. [I]t is the Commission intent that the wording and Bases of the improved STS be usedto the extent practicable.

Entergy has developed the proposed changes consistent with the STS.

10 CFR 50.36(b) Technical Specifications, states that TSs shall be derived from the analyses and evaluation included in the SAR. 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.

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 removes requirements from the TSs which do not meet the TS inclusion criteria of 10 CFR 50.36(C)(2)(ii). This may include SRs which are not necessary to ensure facility operation will be maintained within stipulated safety limits or that ensure LCOs will be met. In addition, the proposed change relocates requirements associated with the monitoring of explosive gas mixtures and radioactive fluid storage to a new TS-controlled program.

Based on the above, the proposed change does not affect compliance with the regulations or relevant regulatory guidance, is consistent with the STS, and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

Enclosure to 2CAN121901 Page 34 of 42 4.2 Precedent This amendment request is based most closely on various NRC approved amendment requests. Identified precedents are listed by subject matter below.

TS 3.1.1.3 - Boron Dilution The removal of this requirement from the TSs has been previously approved under license amendments for conversions to the STS for Palisades Plant (Reference 23) and for D.C.

Cook Nuclear Plant (Reference 24).

TS 3.3.3.1 - Radiation Monitoring Instrumentation No specific precedent was found for this proposed change. However, the criteria used in the justification for the relocation of the Main Steam Line monitor requirements is that provided for accident monitoring instrument by RG 1.97.

TS 3.3.3.5 - Remote Shutdown Instrumentation The removal of the list of Remote Shutdown instruments from the TSs has been generically approved under TSTF-266-A, "Eliminate the Remote Shutdown System Table of Instrumentation and Controls," Revision 3 (Reference 7). The NRC approved this TSTF by letter dated September 10, 1999 (Reference 8). This change has been approved by the NRC for South Texas Project - Units 1 and 2 (Reference 25) and Hope Creek Generating Station (Reference 26).

TS 3.4.6.2 (SR 4.4.6.2.1.b) - Reactor Coolant System Operational Leakage The removal of the requirement to monitor Reactor Vessel flange leakoff temperature from the TSs has been previously approved under several license amendments: Waterford Steam Electric Station, Unit 3 (Reference 27), Sequoyah Nuclear Plant - Units 1 and 2 (Reference 28), and the Callaway conversion to STS (Reference 29).

TS 3.7.2.1 - Steam Generator Pressure/Temperature Limitation The removal of this requirement from the TSs has been previously approved by the NRC under license amendment for Millstone Power Station, Unit Nos. 2 and 3 (Reference 30) and South Texas Project - Units 1 and 2 (Reference 31).

TS 3.7.5.1 - Flood Protection The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Millstone Power Station, Unit Nos. 2 and 3 (Reference 30) and Waterford Steam Electric Station, Unit 3 (Reference 32).

TS 3.7.9.1 - Sealed Source Contamination The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Millstone Power Station, Unit Nos. 2 and 3 (Reference 30),

Waterford Steam Electric Station, Unit 3 (Reference 32), South Texas Project - Units 1 and 2 (Reference 31), and the ANO-1 conversion to STS (Reference 33).

Enclosure to 2CAN121901 Page 35 of 42 TS 3.7.12 - Spent Fuel Pool Structural Integrity No precedent was identified related to this proposed change. However, a similar specification does not exist in the STS.

TS 3.9.3.a - Decay Time and Spent Fuel Storage The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Millstone Power Station, Unit No. 2 (Reference 34), South Texas Project - Units 1 and 2 (Reference 31), and the ANO-1 conversion to STS (Reference 33).

TS 3.9.5 - Communications The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Millstone Power Station, Unit No. 2 (Reference 34),

St Lucie Plant, Units 1 and 2 (Reference 35), South Texas Project - Units 1 and 2 (Reference 31), Waterford Steam Electric Station, Unit 3 (Reference 32), and the ANO-1 conversion to STS (Reference 33).

TS 3.9.6 - Refueling Machine Operability The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Millstone Power Station, Unit No. 2 (Reference 34),

St Lucie Plant, Units 1 and 2 (Reference 35), South Texas Project - Units 1 and 2 (Reference 36), and Waterford Steam Electric Station, Unit 3 (Reference 32).

TS 3.9.7 - Crane Travel - Spent Fuel Pool Building The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Millstone Power Station, Unit No. 2 (Reference 34),

St Lucie Plant, Units 1 and 2 (Reference 37), South Texas Project - Units 1 and 2 (Reference 35), Waterford Steam Electric Station, Unit 3 (Reference 32), and the ANO-1 conversion to STS (Reference 33).

TS 3.11.1 - Liquid Holdup Tanks The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 38), Shearon Harris Nuclear Power Plant, Unit 1(Reference 39), and the ANO-1 conversion to STS (Reference 33).

TS 3.11.2 - Gas Storage Tanks The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 38), Shearon Harris Nuclear Power Plant, Unit 1(Reference 39), and the ANO-1 conversion to STS (Reference 33).

Enclosure to 2CAN121901 Page 36 of 42 TS 3.11.3 - Explosive Gas Mixture The removal of this requirement from the TSs has been previously approved by the NRC under a license amendment for Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 38), South Texas Project - Units 1 and 2 (Reference 31), Shearon Harris Nuclear Power Plant, Unit 1(Reference 39), and the ANO-1 conversion to STS (Reference 33).

TS 6.5.8 - Explosive Gas and Storage Tank Radioactivity Monitoring Program The addition of this program to the TSs has been previously approved by the NRC under a license amendment for Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 38),

Shearon Harris Nuclear Power Plant, Unit 1(Reference 39), and the ANO-1 conversion to STS (Reference 33).

4.3 No Significant Hazards Consideration Analysis Entergy Operations, Inc. (Entergy) has evaluated the proposed changes to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (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 the following changes to the ANO-2 TSs that would permit relocation of information associated with some TSs to other locations within the ANO-2 TS, the TS Bases, the Technical Requirements Manual (TRM), or to a new Explosive Gas and Storage Tank Radioactivity Monitoring Program. Requirements relocated to licensee-controlled documents do not meet the TS inclusion criteria of 10 CFR 50.36(c)(2)(ii). Other requirements are removed from the TSs that do not meet the TS inclusion criteria of 10 CFR 50.36(c)(2)(ii) or are controlled by other regulation.

TS 3.1.1.3, "Boron Dilution," and associated TS Bases are removed since proper mixing of the Reactor Coolant System (RCS) is ensured by other TSs.

TS 3.3.3.1, "Radiation Monitoring Instrumentation," requirements associated with the Main Steam Line Radiation Monitors and respective TS Bases are relocated to the TRM.

TS 3.3.3.5, "Remote Shutdown Instrumentation," list of instruments in Table 3.3-9 and Table 4.3-6 are removed from the TS. These tables will be placed in the associated TS Bases as an operator aid.

TS 3.4.6.2, "Reactor Coolant System Operational Leakage," requirement associated with the monitoring of reactor vessel head flange leakoff temperature is removed. This indication is not indicative of RCS pressure boundary leakage.

TS 3.7.2.1, "Steam Generator Pressure/Temperature Limitation," requirements and associated TS Bases are relocated to the TRM.

TS 3.7.5.1, "Flood Protection," requirements and associated TS Bases are relocated to the TRM.

Enclosure to 2CAN121901 Page 37 of 42 TS 3.7.9.1, "Sealed Source Contamination," and associated TS Bases are removed. This subject matter is governed by 10 CFR 70.39.

TS 3.7.12, "Spent Fuel Pool Structural Integrity," requirements and associated TS Bases are relocated to the TRM.

TS 3.9.3.a, "Decay Time and Spent Fuel Storage," requirements and associated TS Bases are relocated to the TRM.

TS 3.9.5, "Communications," and associated TS Bases are removed. 10 CFR 50, Appendix B provides requirements for establishing procedures to control plant evolutions.

TS 3.9.6, "Refueling Machine Operability," requirements and associated TS Bases are relocated to the TRM.

TS 3.9.7, "Crane Travel - Spent Fuel Pool Building," requirements and associated TS Bases are relocated to the TRM.

TS 3.11.1, "Liquid Holdup Tanks," requirements are relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program. Associated TS Bases are deleted.

TS 3.11.2, "Gas Storage Tanks," requirements are relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program. Associated TS Bases are deleted.

TS 3.11.3, "Explosive Gas Mixture," requirements are relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program. Associated TS Bases are deleted.

New TS 6.5.8, "Explosive Gas and Storage Tank Radioactivity Monitoring Program", is added to Section 6.0 of the TSs consistent with NUREG 1432, Revision 4, "Standard Technical Specifications Combustion Engineering Plants" (STS).

Basis for no significant hazards consideration determination: As required by 10 CFR 50.91(a),

Entergy's analysis of the issue of no significant hazards consideration (NSHC) is presented below.

1.

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

Response: No The proposed change removes or relocates requirements and surveillances for structures, systems, components (SSCs), or variables that do not meet the criteria of 10 CFR 50.36(c)(2)(ii) for inclusion in TSs, are included in other TSs, governed by other regulation, or consists of information not necessary to be included in TSs to ensure safe operation of the facility in accordance with 10 CFR 50.36(b). The affected SSCs or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events.

Enclosure to 2CAN121901 Page 38 of 42 TS requirements associated with explosive gas mixtures and the radiological content of liquid and gas storage tanks are relocated to a new TS Explosive Gas and Storage Tank Radioactivity Monitoring Program, consistent with the STS, which provides controls that meet the intent of these specifications. Therefore, the relocation of these TSs to the new program will not result in a significant increase in the consequences of an accident previously evaluated.

In all remaining cases, information removed or relocated from the TS are maintained in licensee-controlled documents such as the TRM (administered as part of the Safety Analysis Report), the TS Bases, or procedures required in accordance with 10 CFR 50, Appendix B. These documents are maintained pursuant to 10 CFR 50.59, plant administrative procedures which apply applicable regulations and standards, or in accordance with the Bases Control Program, as applicable. Based on these controls and because the proposed change does not involve a physical alteration to any SSC, the consequences of an accident previously evaluated in the SAR are not altered.

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

2.

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

Response: No The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or change in the methods governing normal plant operation. The proposed change will not impose or eliminate any requirements, and adequate control of existing requirements are maintained.

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

3.

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

Response: No The proposed change will not reduce a margin of safety because the change has no significant effect on any safety analyses assumptions, as indicated by the fact that the requirements do not meet the 10 CFR 50.36 criteria for retention, are retained in TS, or are not necessary to ensure the safe operation of the facility. In addition, those requirements that are relocated remain unchanged and any future changes to these requirements are evaluated in accordance with 10 CFR 50.59, plant administrative procedures which apply applicable regulations and standards, or the TS Bases Control Program.

NRC prior review and approval of changes to relocated requirements, in accordance with 10 CFR 50.92, will no longer be required. This review and approval does not provide a specific margin of safety that can be evaluated. However, the proposed change is consistent with NUREG-1432, issued by the NRC, which allows revising the TSs to relocate these requirements and surveillances to licensee-controlled documents.

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

Enclosure to 2CAN121901 Page 39 of 42 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 or relocate 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 or relocate 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.

6.0 REFERENCES

1.

Federal Register Notice 58 FRN 39132, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," dated July 22, 1993.

2.

U.S. Nuclear Regulatory Commission (NRC) NUREG-1432, Revision 4, "Standard Technical Specifications Combustion Engineering Plants," (ML12102A165), published April 2012.

3.

NRC NUREG-0212, Revision 0, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors," (ML17266A003), dated March 15, 1977.

4.

NRC letter to Mr. William Cavanaugh III, "Issuance of Facility Operating License No.

NPF-6 (Arkansas Nuclear One, Unit 2)," (2CNA077805) (ML021490087), dated July 18, 1978.

5.

NRC NUREG-0212, Revision1, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors," (ML17266A004), dated March 1979.

6.

NRC NUREG 1432, Revision 0, "Standard Technical Specifications Combustion Engineering Plants," (ML13196A352), dated September 1992.

7.

Nuclear Energy Institute (NEI) letter to NRC TSTF-266-A, Revision 3, "Eliminate the Remote Shutdown System Table of Instrumentation and Controls," (ML040620072), dated August 1999.

Enclosure to 2CAN121901 Page 40 of 42

8.

NRC letter to NEI, "Approval of TSTFs-212, R.1; -266, R.3; -324, R.1; -327; -388; and -399," dated September 10, 1999.

9.

Federal Register Notice 60 FRN 36953, "Technical Specifications," dated July 19, 1995.

10. NRC NUREG-0737, "Clarification of TMI Action Plan Requirements," (ML051400209),

published November 1980.

11. NRC letter to Entergy Operations, Inc. (Entergy), "Issuance of Amendment Nos 163 and 145," (ML021270024), dated March 6, 1993.

12. NRC NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," (ML090060030), published July 1979.

13. Entergy letter to NRC, "License Amendment Request Post-Accident Instrumentation Technical Specification Revision Arkansas Nuclear One, Unit 2," (2CAN121702)

(ML17348A150), dated December 14, 2017.

14. NRC Regulatory Guide (RG) 1.97, Revision 3, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," (ML003740282), dated May 1983.

15. NRC letter to Entergy, "Arkansas Nuclear One, Unit 2 - Issuance of Amendment No. 313 Re: Post-Accident Instrumentation Technical Specification Revision (EPID L-2017-LLA-0432)," (2CNA121801) (ML18317A382), dated December 19, 2018.

16. NRC RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000.

17. NRC RG 1.13, Revision 1, "Spent Fuel Storage Facility Design Basis," dated December 1975.

18. Entergy letter to NRC, "License Amendment Request - Technical Specification Changes and Analyses Relating to Use of Alternative Source Term - Supplemental Information,"

(2CAN061004) (ML102000199), dated June 23, 2010.

19. NRC Branch Technical Position ETSB 11-5, Revision 3, "Postulated Radioactive Release due to Waste Gas System Leak or Failure," (ML052350110), July 1981.

20. NRC Generic Letter (GL) 95-10, "Relocation of Selected Technical Specifications Requirements Related to Instrumentation," dated December 15, 1995.

21. Entergy letter to NRC, "Annual Radioactive Effluent Release Report for 2018,"

(0CAN041903) (ML19115A122), dated April 25, 2019.

22. NRC GL 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program," dated January 31, 1989.

Enclosure to 2CAN121901 Page 41 of 42

23. NRC letter to Palisades Plant, "Issuance of Amendment Re: Conversion of Current Technical Specifications to Improved Technical Specifications (TAC No. MA0805),"

(ML993510317), dated November 30, 1999.

24. NRC letter to D.C. Cook Nuclear Plant, Units 1 and 2, "Issuance of Amendments for the Conversion to the Improved Technical Specifications with Beyond Scope Issues (TAC Nos. MC2629, MC2630, MC2653 through MC2687, MC2690 through MC2695, MC3152 through MC3157, MC3432 through MC3453)," (ML050620034), dated June 1, 2005.

25. NRC letter to South Texas Project, Units 1 and 2, "Issuance of Amendments Re: Remote Shutdown System (TAC Nos. MC1246 and MC1247)," (ML042370841), dated August 20, 2004.

26. NRC letter to Hope Creek Generating Station, "Issuance of Amendment No. 217 Re: Remote Shutdown System (EPID L-2018-LLA-0295)," (ML19186A205), dated August 6, 2019.

27. NRC letter to Waterford Steam Electric Station, Unit 3, "Issuance of Amendment Re:

Reactor Coolant System Leakage Detection (TAC No. MC3085)," (ML042150057), dated July 30, 2004.

28. NRC letter to Sequoyah Nuclear Plant, Units 1 and 2, "Issuance of Amendments Regarding Enhancement of Reactor Coolant Leakage Detection and Operational Leakage Consistent with Standard Technical Specifications (TAC Nos. MA6760 and MA6761)

(TS 98-10)," (ML003738637), dated August 4, 2000.

29. NRC letter to Callaway Plant, Unit 1, "Conversion to Improved Technical Specifications for Callaway Plant, Unit 1 - Amendment No. 133 to Facility Operating License No. NPF-30 (TAC No. M98803)," (ML021640446), dated May 28, 1999.

30. NRC letter to Millstone Power Station, Unit Nos. 2 and 3, "Issuance of Amendment Re:

Relocate Selected Millstone Unit 2 and 3 Technical Specification Related to The Reactor Coolant System and Plant Systems to the Technical Requirements Manual (TAC Nos.

MB4066 and MB4067)," (ML030030636), dated January 2, 2003.

31. NRC letter to South Texas Project, Units 1 and 2, "Issuance of Amendments on Relocation of Various Technical Specifications (TSs) to the Technical Specification Requirements Manual (TRM) (TAC Nos. MB3588 AND MB3592)," (ML022800581), dated December 17, 2002.

32. NRC letter to Waterford Steam Electric Station, Unit 3, 'Issuance of Amendment Re:

Relocating Technical Specifications to the Technical Requirements Manual (TAC No.

ME7614)," (ML12278A331), dated December 20, 2012.

33. NRC letter to Arkansas Nuclear One, Unit No. 1, "Issuance of Amendment Re: The Conversion to Improved Technical Specifications (TAC No. MA8082)," (ML013090437),

dated October 29, 2001.

Enclosure to 2CAN121901 Page 42 of 42

34. NRC letter to Millstone Nuclear Power Station, Unit No. 2, "Issuance of Amendment Re:

Relocation of Technical Specifications (TAC No. MA6081)," (ML003684825), dated February 10, 2000.

35. NRC letter to St. Lucie Plant, Unit Nos. 1 and 2, "Issuance of Amendments Regarding Technical Specification Change to Remove Communications and Manipulator Crane Requirements and Relocate to Licensee-Controlled Documents (CAC Nos. MF5835 and MF5836)," (ML16034A080), dated March 7, 2016.

36. NRC letter to South Texas Project, Units 1 and 2, "Issuance of Amendments Regarding Relocation of Technical Specifications to the Technical Requirements Manual (TAC Nos.

MB2122 and MB2123)," (ML012260128), dated September 13, 2001.

37. NRC letter to St. Lucie Units 1 and 2, "Issuance of Amendments Regarding the Relocation of Spent Fuel Pool Crane Technical Specification Requirements (TAC Nos. MB5667 and MB5668)," (ML040440119), dated April 28, 2004.

38. NRC letter to Beaver Valley Power Station, Unit Nos. 1 and 2, "Issuance of Amendment Re: Relocating Designated Technical Specifications to the Licensing Requirements Manual and the Offsite Dose Calculation Manual (TAC Nos. MB2048 and MB2049),"

(ML020530410), dated May 21, 2002.

39. NRC letter to Shearon Harris Nuclear Power Plant, Unit 1, "Issuance of Amendment Re:

Relocation of Explosive Gas Monitoring Program Technical Specifications (CAC No. MF8067)," (ML17074A672), dated May 25, 2017.

ATTACHMENTS

1.

Technical Specification Page Markups

2.

Technical Specification Bases Page Markups

3.

Retyped Technical Specification Pages

Enclosure Attachment 1 to 2CAN121901 Technical Specification Page Markups (32 pages)

ARKANSAS - UNIT 2 3/4 1-3 Amendment No. 24,82,157,315, Next page is 3/4 1-5 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg 200 °F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than that required above, immediately initiate and continue boration at 40 gpm of 2500 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT:

a.

Within one hour after detection of an inoperable CEA(S) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(S) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

b.

In accordance with the Surveillance Frequency Control Program by consideration of at least the following factors:

1.

Reactor coolant system boron concentration,

2.

CEA position,

3.

Reactor coolant system average temperature,

4.

Fuel burnup based on gross thermal energy generation,

5.

Xenon concentration, and

6.

Samarium concentration.

ARKANSAS - UNIT 2 3/4 1-4 Amendment No. 126,255,315 REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall be 2000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.

APPLICABILITY:

ALL MODES.

ACTION:

With the flow rate of reactor coolant through the reactor coolant system < 2000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

SURVEILLANCE REQUIREMENTS 4.1.1.3 The flow rate of reactor coolant through the reactor 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 reactor coolant system.

ARKANSAS - UNIT 2 3/4 3-25 Amendment No. 63,130,145,206,231,255, TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE APPLICABLE MODES ALARM/TRIP SETPOINT MEASUREMENT 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 & 6 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 handling of irradiated fuel.

ARKANSAS - UNIT 2 3/4 3-26 Amendment No. 63,130,145,206,231, 255,301, 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 performing CORE ALTERATIONS or moving irradiated fuel within the reactor building, secure the containment purge system or suspend CORE ALTERATIONS and movement of irradiated fuel within the reactor building.

b. If a containment PURGE is in progress, secure the containment purge system.

c. If continuously ventilating, verify the SPING monitor operable or perform the ACTIONS of the Offsite Dose Calculation Manual, Appendix 2, Table 2.2-1, or 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.DELETED

ARKANSAS - UNIT 2 3/4 3-26a Amendment No. 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 handling of irradiated fuel with one or two channels inoperable, immediately place one OPERABLE CREVS train in the emergency recirculation mode or immediately suspend handling of irradiated fuel.

Move to previous page

ARKANSAS - UNIT 2 3/4 3-27 Amendment No. 63,130,145,206,231,255,315, Next page is 3/4 3-36 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE 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 Exhaust Isolation Note 2 SFCP Note 3 5 & 6

b.

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

c.

Main Steam Line Radiation Monitors SFCP SFCP SFCP 1, 2, 3, & 4 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 operations.

Note 3 - Within 31 days prior to initiating containment purge operations and in accordance with the Surveillance Frequency Control Program during containment purge operations.

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 handling of irradiated fuel.

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 3-28 Amendment No. 53,134,151,163,191 THIS PAGE INTENTIONALLY LEFT BLANK (Next page is 3/4 3-36)

ARKANSAS - UNIT 2 3/4 3-36 Amendment No. 281, Next page is 3/4 3-39 INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either restore the inoperable channel to OPERABLE status within 30 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of athe CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6 in accordance with the Surveillance Frequency Control Program. The logarithmic neutron instrumentation, the startup channel instrumentation, and the reactor trip breaker indication are excluded from CHANNEL CALIBRATION.

ARKANSAS - UNIT 2 3/4 3-37 Amendment No. 255 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION INSTRUMENT READOUT LOCATION MEASUREMENT RANGE MINIMUM CHANNELS OPERABLE

1.

Logarithmic Neutron Channel 2C80 10 200%

1

2.

Startup Channel 2C80 1 - 106 cps 1

3.

Reactor Trip Breaker Indication OPEN-CLOSE 1/trip breaker

4.

Reactor Coolant Cold Leg Temperature 2C80 0 - 600 °F 1

5.

Pressurizer Pressure 2C80 0 - 3000 psia 1

6.

Pressurizer Level 2C80 0 - 100%

1

7.

Steam Generator Pressure 2C80 0 - 1200 psia 1/steam generator

8.

Steam Generator Level 2C80 and Local (at EFW Valves Control) 0 - 100%

1/steam generator

9.

Shutdown Cooling Flow Rate 2C80 0 - 8000 gpm 1

10.

Condensate Storage Tank Level 2C80 0 - 100%

1

ARKANSAS - UNIT 2 3/4 3-38 Amendment No. 255,315 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION

1.

Logarithmic Neutron Channel SFCP N.A.

2.

Startup Channel SFCP N.A.

3.

Reactor Trip Breaker Indication SFCP N.A.

4.

Reactor Coolant Cold Leg Temperature SFCP SFCP

5.

Pressurizer Pressure SFCP SFCP

6.

Pressurizer Level SFCP SFCP

7.

Steam Generator Level SFCP SFCP

8.

Steam Generator Pressure SFCP SFCP

9.

Shutdown Cooling Flow Rate SFCP SFCP

10.

Condensate Storage Tank Level SFCP SFCP

ARKANSAS - UNIT 2 3/4 4-14a Order date 4/20/81 Amendment No. 231,266,315, REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakage, except for primary to secondary leakage, shall be demonstrated to be within each of the above limits by:

a.

pPerformance of a Reactor Coolant System water inventory balance in accordance with the Surveillance Frequency Control Program during steady state operation except when operating in the shutdown cooling mode*.

b.

Monitoring the reactor head flange leakoff temperature in accordance with the Surveillance Frequency Control Program.

4.4.6.2.2 Primary to secondary leakage shall be verified to be 150 gallons per day through any one SG in accordance with the Surveillance Frequency Control Program*.

4.4.6.2.3 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4.6-1 shall be demonstrated OPERABLE by individually verifying leakage to be within its limit:

a.

Prior to entering MODE 2 after each refueling outage,

b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, and

c.

Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

ARKANSAS - UNIT 2 3/4 7-10 Amendment No. 233,281,305, Next page is 3/4 7-15 PLANT SYSTEMS MAIN STEAM ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam isolation valve shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

MODE 1 With one main steam isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

MODES 2 With one main steam isolation valve inoperable, subsequent operation in and 3 MODES 1, 2 or 3 may proceed provided the isolation valve is maintained closed; otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam isolation valve shall be demonstrated OPERABLE by verifying full closure within 3 seconds when tested pursuant to the INSERVICE TESTING PROGRAM.

3/4 7-11 3/4 7-12 ARKANSAS - UNIT 2 3/4 7-13 Amendment No. 30 These pages intentionally left blank. Section 3.7.1.6 deleted by issuance of Amendment No.

ARKANSAS - UNIT 2 3/4 7-14 Amendment No. 315 PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 90 °F when the pressure of either coolant in the steam generator is > 275 psig.

APPLICABILITY:

At all times.

ACTION:

With the requirements of the above specification not satisfied:

a.

Reduce the steam generator pressure of the applicable side to 275 psig within 30 minutes, and

b.

Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200 °F.

SURVEILLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generators shall be determined to be

< 275 psig in accordance with the Surveillance Frequency Control Program when the temperature of either the primary or secondary coolant is < 90 °F.

ARKANSAS - UNIT 2 3/4 7-16a Amendment No. 315 PLANT SYSTEMS 3/4.7.5 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.5.1 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Dardanelle Reservoir exceeds 350 feet Mean Sea Level USGS datum, at the intake structure.

APPLICABILITY:

When a flood warning exists at the facility site.

ACTION:

With the water level at the intake structure above elevation 350 feet Mean Sea Level USGS datum, initiate and complete within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , closure of the openings and penetrations listed in Table 3.7-6 using the equipment listed in Table 3.7-6.

SURVEILLANCE REQUIREMENTS 4.7.5.1 The water level at the intake structure shall be determined to be within the limits by:

a.

Measurement in accordance with the Surveillance Frequency Control Program when the water level is below elevation 350 feet Mean Sea Level USGS datum, and

b.

Measurement in accordance with the Surveillance Frequency Control Program when the water level is equal to or above elevation 350 feet Mean Sea Level USGS datum.

ARKANSAS - UNIT 2 3/4 7-16b TABLE 3.7-6 FLOOD PROTECTION PROVISIONS Structure Opening or Penetration Type of Flooding Protection Containment Equipment hatch Double seal in Hatch cover Escape lock Double seal in Lock doors Tendon gallery exits Water Tight Scuttle Auxiliary Building Door openings Watertight Doors Floor openings Watertight hatch covers Roof openings over underground vaults Concrete plugs with neoprene seals Pipe penetrations Rubber seals or closure plates Emergency Diesel Fuel Storage Vaults Door opening Watertight Door Roof opening Concrete Plug with neoprene seals Pipe Penetrations Rubber seals or closure plates

ARKANSAS - UNIT 2 3/4 7-18 Amendment No. 191,206,219,255, Next Page is 3/4 7-27 283,288,315, PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.6.1.1 Each control room emergency air conditioning system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by:

1.

Starting each unit from the control room, and

2.

Verifying that each unit operates for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and maintains the control room air temperature 84 °F D.B.

b.

In accordance with the Surveillance Frequency Control Program by verifying a system flow rate of 9900 cfm +/- 10%.

4.7.6.1.2 Each control room emergency air filtration system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the system operates for at least 15 minutes.

b.

In accordance with the Surveillance Frequency Control Program by verifying that on a control room high radiation signal, either actual or simulated, the system automatically isolates the control room and switches into a recirculation mode of operation.

c.

By performing the required Control Room Emergency Ventilation filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

d.

Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

ARKANSAS - UNIT 2 3/4 7-27 Amendment No. 134,315 PLANT SYSTEMS 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.9.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of 0.005 microcuries of removable contamination.

APPLICABILITY:

At all times.

ACTION:

a.

Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:

1.

Either decontaminated and repaired, or

2.

Disposed of in accordance with Commission Regulations.

b.

The provisions of Specifications 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.9.1.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a.

The licensee, or

b.

Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.9.1.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequencies described below.

a.

Sources in use - In accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive material:

ARKANSAS - UNIT 2 3/4 7-28 Amendment No. 58,212 Next Page is 3/4 7-38 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1.

With a half-life greater than 30 days (excluding Hydrogen 3), and

2.

In any form other than gas.

b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c.

Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source or detector.

4.7.9.1.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of 0.005 microcuries of removable contamination.

ARKANSAS - UNIT 2 3/4 7-38 Amendment No. 91,117,191,255 PLANT SYSTEMS 3/4.7.12 SPENT FUEL POOL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.7.12 The structural integrity of the spent fuel pool shall be maintained in accordance with Specification 4.7.12.

APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

a.

With the structural integrity of the spent fuel pool not conforming to the above requirements, in lieu of any other report, prepare and submit a Special Report to the NRC within 30 days of a determination of such non-conformity.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.12.1 Inspection Frequencies - The structural integrity of the spent fuel pool shall be determined per the acceptance criteria of Specification 4.7.12.2 at the following frequencies:

a.

At least once per 92 days after the pool is filled with water. If no abnormal degradation or other indications of structural distress are detected during five consecutive inspections, the inspection interval may be extended to at least once per 5 years.

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following any seismic event which actuates or should have actuated the seismic monitoring instrumentation.

4.7.12.2 Acceptance Criteria - The structural integrity of the spent fuel pool shall be determined by a visual inspection of at least the interior and exterior surfaces of the pool, the struts in the tilt pit, the surfaces of the separation walls, and the structural slabs adjoining the pool walls. This visual inspection shall verify no changes in the concrete crack patterns, no abnormal degradation or other signs of structural distress (i.e, cracks, bulges, out of plumbness, leakage, discolorations, efflorescence, etc.).

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

ACTION:

a.

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

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 CORE ALTERATIONS.

ARKANSAS - UNIT 2 3/4 9-3 Amendment No. 43,166,240 REFUELING OPERATIONS DECAY TIME AND SPENT FUEL STORAGE LIMITING CONDITION FOR OPERATION 3.9.3.a The reactor shall be subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

APPLICABILITY:

During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3.a The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

ARKANSAS - UNIT 2 3/4 9-4 Amendment No. 166,203,230,315, Next page is 3/4 9-69 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 irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment.

The provisions of Specification 3.0.3 are not applicable.

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 irradiated fuel in the containment.

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 9-6 Amendment No. 315 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.

ACTION:

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE 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 CORE ALTERATIONS.

ARKANSAS - UNIT 2 3/4 9-7 Amendment No. 24,169 Correction Letter dated 10/24/95 REFUELING OPERATIONS REFUELING MACHINE OPERABILITY LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine shall be used for movement of fuel assemblies and shall be OPERABLE with:

a.

A minimum capacity of 3750 pounds,

b.

An overload cut off limit of 100 pounds plus the combined weight of one fuel assembly, one CEA, and the grapple in the "fuel only" region, and

c.

An overload cut off limit of 100 pounds plus the combined weight of one fuel assembly, one CEA, the grapple, and the hoist box in the "fuel plus hoist box" region.

APPLICABILITY:

During movement of CEAs or fuel assemblies within the reactor pressure vessel.

ACTION:

With the requirements for refueling machine OPERABILITY not satisfied, suspend its use from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.6 The refueling machine shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of movement of fuel assemblies within the reactor pressure vessel by performing a load test of at least 3750 pounds and demonstrating automatic load cut offs when the crane loads exceed 100 pounds plus the applicable loads.

ARKANSAS - UNIT 2 3/4 9-8 Amendment No. 315 REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool.

APPLICABILITY:

With fuel assemblies in the spent fuel pool.

ACTION:

With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.7 The crane electrical power disconnect which prevents crane travel over the spent fuel pool shall be verified open under administrative control in accordance with the Surveillance Frequency Control Program, or the crane travel interlock which prevents crane travel over the spent fuel pool shall be demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to each use of the crane for lifting loads in excess of 2000 pounds.

ARKANSAS - UNIT 2 3/4 11-1 Amendment No. 60,134,149,193,315 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID HOLDUP TANKS*

LIMITING CONDITION FOR OPERATION 3.11.1 The quantity of radioactive material contained in each unprotected outside temporary radioactive liquid storage tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material exceeding the above limit, immediately suspend all additions of radioactive material to the affected tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit and describe the events leading to the condition in the next Radioactive Effluents Release Report pursuant to Specification 6.9.3.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1 The quantity of radioactive material contained in each unprotected outside temporary radioactive liquid storage tank shall be determined to be within the above limit by analyzing a representative sample of the contents of the tank in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank.

Tanks included in this specification are those outdoor temporary tanks that 1) are not surrounded by liners, dikes, or walls capable of holding the tank contents, and 2) do not have overflows and surrounding area drains connected to the liquid radwaste treatment system.

ARKANSAS - UNIT 2 3/4 11-2 Amendment No. 60,134,193,211,315 RADIOACTIVE EFFLUENTS 3/4.11.2 GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 82,400 curies noble gases (considered as Xe-133).

APPLICABILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit and describe the events leading to the condition in the next Radioactive Effluent Release Report pursuant to Specification 6.9.3.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank and the reactor coolant activity exceeds the limits of Specification 3.4.8.

ARKANSAS - UNIT 2 3/4 11-3 Amendment No. 60,61,134,193 RADIOACTIVE EFFLUENTS 3/4.11.3 EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.3 The concentration of the hydrogen/oxygen shall be limited in the waste gas storage tanks to Region "A" of Figure 3.11-1.

APPLICABILITY:

At all times.

ACTION:

a.

When the concentration of hydrogen/oxygen in the waste gas storage tanks enters Region "B" of Figure 3.11-1, corrective action shall be taken to return the concentration values to Region "A" within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3 The concentration of hydrogen/oxygen in the waste gas holdup system shall be determined to be within the above limits, with the waste gas system in operation, by continuously monitoring with the hydrogen/oxygen monitors required OPERABLE by Table 3.11-3.

ARKANSAS - UNIT 2 3/4 11-4 Amendment No. 60,61,91,134,193 TABLE 3.11-3 EXPLOSIVE GAS MONITORING INSTRUMENTATION Instrument Minimum Channels Operable Applicability Action

1. Waste Gas Holdup System Explosive Gas Monitoring System

a. Hydrogen monitor

b. Oxygen monitor 1

1 1

1

During waste gas compressing operation (treatment for primary system off gases.)

ACTION 1 - With both channels inoperable, operation may continue provided grab samples are taken 1) every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations, and 2) daily during other operations. The analysis of these samples shall be completed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of taking the sample.

ARKANSAS - UNIT 2 3/4 11-5 Amendment No. 60,61,91,134,193 HYDROGEN - OXYGEN LIMITS FOR ANO-2 WASTE GAS SYSTEM Figure 3.11-1 0

2 4

6 8

10 12 14 16 18 20 22 0

20 40 60 80 100 120 OXYGEN %

HYDROGEN %

REGION B (UNACCEPTABLE OPERATION)

REGION A (ACCEPTABLE OPERATION)

(4,4)

ARKANSAS - UNIT 2 6-7 Amendment No. 255,262,291,305, ADMINISTRATIVE CONTROLS 6.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendation of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. The volumetric examination per Regulatory Position C.4.b.1 will be performed on approximately 10-year intervals.

6.5.8 DELETEDExplosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Decay Tanks, the quantity of radioactivity contained in the Waste Gas Decay Tanks, and the quantity of radioactivity contained in unprotected temporary outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be determined in accordance with the ODCM.

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen in the Waste Gas Decay Tanks and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion),

b.

A surveillance program to ensure that the quantity of radioactivity contained in each Waste Gas Decay Tank is less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents, and

c.

A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

Enclosure Attachment 2 to 2CAN121901 Technical Specification Bases Page Markups (10 pages)

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

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 3-4 Amendment No. 22,29,60,123,132,191 Rev. 8,17,40,56,60,70, INSTRUMENTATION BASES 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.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the Control Room. This capability is required in the event Control Room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. With regard to CST level, the required Remote Shutdown (RSD) panel indication is that CST level indication associated with the CST aligned to the EFW system. The following RSD instruments are required to be OPERABLE:

INSTRUMENT READOUT LOCATION MEASUREMENT RANGE MINIMUM CHANNELS OPERABLE

1.

Logarithmic Neutron Channel 2C80 10 200%

1

2.

Startup Channel 2C80 1 - 106 cps 1

3.

Reactor Trip Breaker Indication OPEN-CLOSE 1/trip breaker

4.

Reactor Coolant Cold Leg Temperature 2C80 0 - 600 °F 1

5.

Pressurizer Pressure 2C80 0 - 3000 psia 1

6.

Pressurizer Level 2C80 0 - 100%

1

7.

Steam Generator Pressure 2C80 0 - 1200 psia 1/SG

8.

Steam Generator Level 2C80 and Local (at EFW Valves Control) 0 - 100%

1/SG

9.

Shutdown Cooling Flow Rate 2C80 0 - 8000 gpm 1

10.

Condensate Storage Tank Level 2C80 0 - 100%

1

ARKANSAS - UNIT 2 B 3/4 3-4 Amendment No. 22,29,60,123,132,191 Rev. 8,17,40,56,60,70, INSTRUMENTATION BASES 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION (continued)

The following CHANNEL CHECKS and CHANNEL CALIBRATIONS are applicable to each required RSD instrument:

INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION

1.

Logarithmic Neutron Channel SFCP N.A.

2.

Startup Channel SFCP N.A.

3.

Reactor Trip Breaker Indication SFCP N.A.

4.

Reactor Coolant Cold Leg Temperature SFCP SFCP

5.

Pressurizer Pressure SFCP SFCP

6.

Pressurizer Level SFCP SFCP

7.

Steam Generator Level SFCP SFCP

8.

Steam Generator Pressure SFCP SFCP

9.

Shutdown Cooling Flow Rate SFCP SFCP

10.

Condensate Storage Tank Level SFCP SFCP 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION BACKGROUND The primary purpose of the Post Accident Monitoring (PAM) instrumentation is to display plant variables that provide information required by the Control Room Operators during accident situations. This information provides the necessary support for the Operator to take the manual actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions for Design Basis Events. The OPERABILITY of the PAM instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess plant status and behavior following an accident. The availability of PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. This capability is consistent with the recommendations of Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 as required by Supplement 1 to NUREG-0737, "TMI Action Items," and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations."

ARKANSAS - UNIT 2 B 3/4 7-4 Amendment No. 153 Rev. 26,33,56, PLANT SYSTEMS BASES 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations to 90 °F and 275 psig are based on a steam generator RTNDT of 30 °F and are sufficient to prevent brittle fracture.

3/4.7.3 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

If the inoperable Service Water Pump cannot be restored to an OPERABLE status within the allowable outage time, 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 7-5 Amendment No. 153 Revised by letter dated 9/8/95 Rev. 26,33,56, PLANT SYSTEMS BASES 3/4.7.4 EMERGENCY COOLING POND The limitations on the emergency cooling pond volume and temperature are based on worst case initial conditions which could be present considering a simultaneous normal shutdown of Unit 1 and emergency shutdown of Unit 2 following a LOCA in Unit 2, using the ECP as a heat sink. The minimum indicated ECP level of 5.2 feet is based on soundings and includes measurement, calculation, and other uncertainties (equivalent to 0.15 feet) to ensure a minimum contained water volume of 70 acre-feet (equivalent to an indicated level of 5.05 feet), crediting operator action to initiate makeup to the ECP upon a loss of Dardanelle Reservoir event as discussed below. These soundings ensure degradation is within acceptable limits such that the indicated level is consistent with the required volume and the pond meets its design basis. The measured ECP temperature at the discharge from the pond is considered a conservative average of total pond conditions since solar gain, wind speed, and thermal current effects throughout the pond will essentially be at equilibrium conditions under initial stagnant conditions.

Visual inspections are performed to ensure erosion, undercut caused by wave action, or any physical degradation is within acceptable limits to enable the ECP to fulfill its safety function. An engineering evaluation shall be performed by a qualified engineer of any apparent changes in visual appearance or other abnormal degradation within 7 days to determine operability.

The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to safety-related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Plants, March 1974. Operator action is credited in the inventory analysis during the transfer of the service water system to the pond. Specifically, pump returns are transferred to the pond shortly after a loss of lake event and pump suctions are transferred later in the event depending on pump bay level. In the time frame between the transfer of the returns and suctions to the pond, lake water is pumped into the pond, increasing level by at least 4.5 inches. This additional water is required, along with that maintained by Technical Specifications, to ensure a 66.9 inch pond depth, which ensures a 30 day supply of cooling water.

ACTION a permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(s) while relying on the ACTION.

3/4.7.5 FLOOD PROTECTION The limitation on flood protection ensures that facility protective actions will be taken in the event of flood conditions.

ARKANSAS - UNIT 2 B 3/4 7-14 Amendment No. 99,132,206 Rev. 1,11,36,41,56,63, PLANT SYSTEMS BASES 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION (CREVS) AND AIR CONDITIONING SYSTEM (CREACS) (continued)

SURVEILLANCE REQUIREMENTS (continued)

SR 4.7.6.1.2.d (continued)

The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, ACTION d must be entered. ACTION d allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident.

Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 3) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 4). These compensatory measures may also be used as mitigating actions. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 5).

Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.

REFERENCES

1.

SAR, Section 6.4 and 9.4.

2.

SAR, Chapter 15.

3.

Regulatory Guide 1.196.

4.

NEI 99-03, "Control Room Habitability Assessment," June 2001.

5.

Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) dated January 30, 2004, "NEI Draft White Paper, Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability." (ADAMS Accession No. ML040300694) 3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

ARKANSAS - UNIT 2 B 3/4 7-15 Amendment No. 99,132 Rev. 63, PLANT SYSTEMS BASES 3/4.7.11 FIRE BARRIERS DELETED 3/4.7.12 SPENT FUEL POOL STRUCTURAL INTEGRITY The reinforcing steel in the walls of the spent fuel pool was erroneously terminated into the front face instead of the rear face of the adjoining walls during construction of the spent fuel pool.

Therefore, the specified structural integrity inspections of the spent fuel pool are required to be performed to ensure that the pool remains safe for use and that it will adequately resist the imposed loadings. If no abnormal degradation is observed during the first five inspections, the inspection interval for subsequent routine inspections may be extended to at least once per 18 months or longer if justified by observed performance of the pool.

ARKANSAS - UNIT 2 B 3/4 9-1 Amendment No. 43,166,203,230,240 Rev. 48,58, 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: 1) the reactor will remain subcritical during CORE 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.

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).

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.

3/4.9.4 CONTAINMENT PENETRATIONS 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 element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

Containment penetrations, the personnel airlock doors, and/or the equipment door may be open during movement of irradiated fuel in the containment and during CORE ALTERATIONS provided a minimum of one closure method (manual or automatic valve, blind flange, or equivalent) in each penetration, one door in each airlock, and the equipment door are capable of being closed in the event of a fuel handling accident. This allowance assumes that 23 feet of water is maintained above the fuel seated within the reactor vessel to ensure any offsite dose consequence remains within 10 CFR 50.67 limits in the event of a fuel handling accident.

Equivalent 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.

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

3/4.9.6 REFUELING MACHINE OPERABILITY 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 11-1 Amendment No. 60,193 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that, in the event of an uncontrolled release of the contents of the tanks, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3/4.11.2 GAS STORAGE TANKS Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that, in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest EXCLUSION AREA boundary will not exceed 0.5 rem. This is consistent with Branch Technical Position ETSB 11-5 in NUREG-0800, July 1981.

3/4.11.3 EXPLOSIVE GAS MIXTURE It is expected that the hydrogen/oxygen concentration will be kept within the limits and therefore not enter the flammable or detonable region concentrations within the waste gas storage tanks.

These levels provide reasonable assurance that no hydrogen/oxygen explosion could occur to allow rupture of the waste gas storage tanks. The hydrogen and oxygen limits are based on information in NUREG/CR-2726, "Light Water Reactor Hydrogen Manual."

Grab samples are to be taken every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations when both hydrogen/oxygen analyzers are out of service. These samples are to be analyzed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to assure that the hydrogen/oxygen concentration is within the limits in Figure 3.11-1. During other Waste Gas Compressor operations, the hydrogen/oxygen concentration is not as subject to change, therefore grab samples are to be taken every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Enclosure Attachment 3 to 2CAN121901 Retyped Technical Specification Pages (11 pages)

ARKANSAS - UNIT 2 3/4 1-3 Amendment No. 24,82,157,315, Next page is 3/4 1-5 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg 200 °F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than that required above, immediately initiate and continue boration at 40 gpm of 2500 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT:

a.

Within one hour after detection of an inoperable CEA(S) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(S) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

b.

In accordance with the Surveillance Frequency Control Program by consideration of at least the following factors:

1.

Reactor coolant system boron concentration,

2.

CEA position,

3.

Reactor coolant system average temperature,

4.

Fuel burnup based on gross thermal energy generation,

5.

Xenon concentration, and

6.

Samarium concentration.

ARKANSAS - UNIT 2 3/4 3-25 Amendment No. 63,130,145,206,231,255, TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE APPLICABLE MODES ALARM/TRIP SETPOINT MEASUREMENT 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 & 6 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 Note 1 - With fuel in the spent fuel pool or building.