ML082540799

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Issuance of Amendment No. 239, Modify Technical Specifications to Adopt TSTF-448, Revision 3, Control Room Habitability Using Consolidated Line Item Improvement Process
ML082540799
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 10/29/2009
From: Wang A
Plant Licensing Branch IV
To:
Entergy Operations
Wang, A B, NRR/DORL/LPLIV, 415-1445
Shared Package
ML082540777 List:
References
TAC MD7177
Download: ML082540799 (44)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 29, 2009 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT NO.1 - ISSUANCE OF AMENDMENT RE:

ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE (TSTF) CHANGE TRAVELER TSTF-448, REVISION 3, "CONTROL ROOM ENVELOPE HABITABILITY" (TAC NO. MD7177)

Dear Sir or Madam:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 239 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit NO.1 (ANO-1). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 22, 2007, as supplemented by letters dated January 12 and October 22, 2009.

The amendment adds a new license condition 2.c.(10) on the control room envelope habitability program, revises TS requirements related to control room envelope habitability in TS 3.7.9, "Control Room Emergency Ventilation System (CREVS)," and adds a new administrative controls program, TS 5.5.5, "Control Room Envelope Habitability Program." These changes are consistent with the NRC-approved IndustryfTS Task Force (TSTF) change traveler TSTF-448, Revision 3, "Control Room Envelope Habitability." The availability of this TS improvement was published in the Federal Register on January 17, 2007 (72 FR 2022), as part of the consolidated line item improvement process.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313

Enclosures:

1. Amendment No. 239 to DPR-51
2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 239 Renewed License No. DPR-51

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated October 22,2007, as supplemented by letters dated January 12 and October 22, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

-2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.c.(2) of Renewed Facility Operating License No. OPR-51 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 239, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.

3. In addition, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and a new license condition under Paragraph 2.c.(1 0) of Renewed Facility Operating License No. OPR-51 is hereby amended to read as follows:

(10) Upon implementation of Amendment 239 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.9.4, in accordance with Specifications 5.5.5.c.(i), 5.5.5.c.(ii), and 5.5.5.d, shall be considered met. Following implementation:

1. The first performance of SR 3.7.9.4, in accordance with Specification 5.5.5.c.(i), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
2. The first performance of the periodic assessment of CRE habitability, Specification 5.5.5.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
3. The first performance of the periodic measurement of CRE pressure, Specification 5.5.5.d, shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.

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4. The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the implementation of the Alternate Source Term license Amendment No. 238.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Attachment Changes to the Renewed Facility Operating License No. DPR-51 and Technical Specifications Date of Issuance: October 29, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 239 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 Replace the following pages of the Renewed Facility Operating License No. DPR-51 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Operating License REMOVE INSERT 3 3 6 6 Technical Specifications REMOVE INSERT 3.7.9-1 3.7.9-1 3.7.9-2 3.7.9-2 3.7.9-3 5.0-9 5.0-9 5.0-10 5.0-10 5.0-11 5.0-11 5.0-12 5.0-12 5.0-20 5.0-13 5.0-21 5.0-14 5.0-22 5.0-15 5.0-23 5.0-16 5.0-24 5.0-17 5.0-25 5.0-18 5.0-25a 5.0-19 5.0-26 5.0-20 5.0-27 5.0-21 5.0-28 5.0-22 5.0-29 5.0-23 5.0-30 5.0-24 5.0-31 5.0-25 5.0-32 5.0-26 5.0-33

-3 (5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6) EOI, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

c. This renewed license shall be deemed to contain and is sUbject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 239, are hereby incorporated in the renewed license.

EOI shall operate the facility in accordance with the Technical Specifications.

(3) Safety Analysis Report The licensee's SAR supplement submitted pursuant to 10 CFR 54.21(d),

as revised on March 14,2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.

(4) Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan," as submitted on May 4,2006.

Renewed License No. OPR-51 Amendment No. 239 Revised by letter dated July 18, 2007

-6 (10) Upon implementation of Amendment 239 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.9.4, in accordance with Specifications 5.5.5.c.(i),

5.5.5.c.(ii), and 5.5.5.d, shall be considered met. Following implementation:

1. The first performance of SR 3.7.9.4, in accordance with Specification 5.5.5.c.(i), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
2. The first performance of the periodic assessment of CRE habitability, Specification 5.5.5.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
3. The first performance of the periodic measurement of CRE pressure, Specification 5.5.5.d, shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
3. This renewed license is effective as of the date of issuance and shall expire at midnight, May 20, 2034.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by:

Jon R. Johnson Jon R. Johnson, Acting Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A - Technical Specifications and Technical Specifications Bases (ML011710071 and ML011710100)

Date of Issuance: June 20, 2001 Renewed License No. DPR-51 Amendment No. 239 Revised by letter dated July 18, 2007

CREVS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Control Room Emergency Ventilation System (CREVS)

LCO 3.7.9 Two CREVS trains shall be OPERABLE.

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

1. The control room envelope (CRE) boundary may be opened intermittently under administrative controls.
2. One CREVS train shall be capable of automatic actuation.

APPLICABI L1TY: MODES 1, 2, 3, 4, During movement of irradiated fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A, One CREVS train A,1 Restore CREVS train to 7 days inoperable for reasons OPERABLE status.

other than Condition B.

B. One or more CREVS trains B.1 Initiate action to implement Immediately inoperable due to mitigating actions.

inoperable CRE boundary in MODES 1, 2, 3, or 4. AND B.2 Verify mitigating actions 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

AND B.3 Restore control room 90 days boundary to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ANO-1 3.7.9-1 Amendment No. ~, 239

CREVS 3.7.9 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and 0.1 Place OPERABLE CREVS Immediately associated Completion train in emergency Time of Condition A not recirculation mode.

met during movement of irradiated fuel assemblies. OR 0.2 Suspend movement of Immediately irradiated fuel assemblies.

E. Two CREVS trains E.1 Suspend movement of Immediately inoperable during irradiated fuel assemblies.

movement of irradiated fuel assemblies.

OR One or more CREVS trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies.

F. Two CREVS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CREVS train for 2 15 minutes. 31 days SR 3.7.9.2 Perform required CREVS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.9.3 Verify the CREVS automatically isolates the Control 18 months Room and switches into a recirculation mode of operation on an actual or simulated actuation signal.

ANO-1 3.7.9-2 Amendment No. 2-1-e, 239

CREVS 3.7.9 SURVEILLANCE FREQUENCY SR 3.7.9.4 Perform required CRE unfiltered air inleakage testing In accordance with in accordance with the Control Room Envelope the Control Room Habitability Program. Envelope Habitability Program.

ANO-1 3.7.9-3 Amendment No. ~,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.5 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.
b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision O.
d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.
e. The quantitative limits on unfiltered air in leakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air in leakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered in leakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

ANO-1 5.0-9 Amendment No.~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.6 (Not Used).

5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. Surface and volumetric examination of the reactor coolant pump flywheels will be conducted coincident with refueling or maintenance shutdowns such that during 10 year intervals all four reactor coolant pump flywheels will be examined. Such examinations will be performed to the extent possible through the access ports, i.e., those areas of the flywheel accessible without motor disassembly. The surface and volumetric examination may be accomplished by Acoustic Emission Examination as an initial examination method. Should the results of the Acoustic Emission Examination indicate that additional examination is necessary to ensure the structural integrity of the flywheel, then other appropriate NDE methods will be performed on the area of concern.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program inspection frequencies.

5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Code terminology for Required Frequencies for performing inservice testing activities inservice testing activities Monthly At least once per 31 days Every 6 weeks At least once per 42 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;
c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and
d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

ANO-1 5.0-10 Amendment No. 2-1-5, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm.
3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

ANO-1 5.0-11 Amendment No. 2-+a,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
2. Inspect 100% of the tubes at sequential periods of 144, 108,72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.
3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e. Provisions for monitoring operational primary to secondary LEAKAGE.

ANO-1 5.0-12 Amendment No. ~,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.10 Secondary Water Chemistry This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;
b. Identification of the procedures used to measure the values of the critical variables;
c. Identification of process sampling points;
d. Procedures for the recording and management of data;
e. Procedures defining corrective actions for all off control point chemistry conditions; and
f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.

5.5.11 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safeguards (ES) ventilation systems filters at the frequencies specified in Regulatory Guide 1.52, Revision 2. The VFTP is applicable to the Penetration Room Ventilation System (PRVS) and the Control Room Emergency Ventilation System (CREVS).

a. Demonstrate that an inplace cold DOP test of the high efficiency particUlate (HEPA) filters shows:
1. ~ 99% DOP removal for the PRVS when tested at the system design f10wrate of 1800 scfm +/- 10%; and
2. ~ 99.95% DOP removal for the CREVS when tested in accordance with Regulatory Guide 1.52, Revision 2, at the system design flowrate of 2000 cfm +/- 10%.

ANO-1 5.0-13 Amendment No. 2-+a,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals

b. Demonstrate that an inplace halogenated hydrocarbon test of the charcoal adsorbers shows:
1. ~ 99% halogenated hydrocarbon removal for the PRVS when tested at the system design flowrate of 1800 cfm +/- 10%; and
2. ~ 99.95% halogenated hydrocarbon removal for the CREVS when tested in accordance with Regulatory Guide 1.52, Revision 2, at the system design flowrate of 2000 cfm +/- 10%.
c. Demonstrate that a laboratory test of a sample of the charcoal adsorber meets the laboratory testing criteria of ASTM D3803-1989 when tested at 30°C and 95% relative humidity for a methyl iodide penetration of:
1. < 5% for the PRVS;
2. when obtained as described in Regulatory Guide 1.52, Revision 2, for CREVS
i. :s; 2.5% for 2 inch charcoal adsorber beds; and ii. :s; 0.5% for 4 inch charcoal adsorber beds.
d. Demonstrate for the PRVS and CREVS, that the pressure drop across the combined HEPA filters, other filters in the system, and the charcoal adsorbers is < 6 inches of water when tested at the following system design flowrates +/- 10%:

PRVS 1800 cfm CREVS 2000 cfm The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

ANO-1 5.0-14 Amendment No. ~,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas System, the quantity of radioactivity contained in gas storage 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 System and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (Le., 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 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;
c. 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.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

ANO-1 5.0-15 Amendment No. ~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
1. an API gravity or an absolute specific gravity within limits,
2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and
3. water and sediment within limits;
b. Within 31 days following addition of new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 20 fuel oil;
c. Total particulate concentration of the fuel oil is ~ 10 mgll when tested every 31 days based on ASTM 0-2276, Method A-2 or A-3; and
d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program surveillance Frequencies.

5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
1. A change in the TS incorporated in the license; or
2. A change to the updated SAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

Proposed changes that do meet these criteria shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the SAR.

ANO-1 5.0-16 Amendment No. 2-+5,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, and assuming no concurrent loss of offsite power or loss of onsite diesel generator(s),

a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

ANO-1 5.0-17 Amendment No. ~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.16 Reactor Building Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the reactor building as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, except that the next Type A test performed after the April 16, 1992 Type A test shall be performed no later than April 15, 2007.

In addition, the reactor building purge supply and exhaust isolation valves shall be leakage rate tested once prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days.

The peak calculated reactor building internal pressure for the design basis loss of coolant accident, Pa, is 54 psig.

The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.

Reactor Building leakage rate acceptance criteria is ~ 1.0La. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60La for the Type B and Type C tests and < 0.75La for Type A tests.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.

ANO-1 5.0-18 Amendment No. 24-5,~,~, 239

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.17 Metamic Coupon Sampling Program A coupon surveillance program will be implemented to maintain surveillance of the Metamic absorber material under the radiation, chemical, and thermal environment of the SFP. The purpose of the program is to establish the following:

  • Coupons will be examined on a two year basis for the first three intervals with the first coupon retrieved for inspection being on or before February 2009 and thereafter at increasing intervals over the service life of the inserts.
  • Measurements to be performed at each inspection will be as follows:

A) Physical observations of the surface appearance to detect pitting, swelling or other degradation, B) Length, width, and thickness measurements to monitor for bulging and swelling C) Weight and density to monitor for material loss, and D) Neutron attenuation to confirm the B-10 concentration or destructive chemical testing to determine the boron content.

  • The provisions of SR 3.0.2 are applicable to the Metamic Coupon Sampling Program.
  • The provisions of SR 3.0.3 are not applicable to the Metamic Coupon Sampling Program.

ANO-1 5.0-19 Amendment No. 22-3, 239

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements 5.6.1 DELETED 5.6.2 Annual Radiological Environmental Operating Report

1\1 0 TE----------------------------------------------

A single submittal may be made for ANO. The submittal should combine sections common to both units.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.2 Annual Radiological Environmental Operating Report (continued)

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report

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

A single submittal may be made for ANO. The submittal shall combine sections common to both units. The submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

5.6.4 DELETED ANO-1 5.0-20 Amendment No. 24-a,~, 239

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

2.1.1 Variable Low RCS Pressure - Temperature Protective Limits 3.1.1 SHUTDOWN MARGIN (SDM) 3.1.8 PHYSICS TESTS Exceptions - MODE 1 3.1.9 PHYSICS TEST Exceptions - MODE 2 3.2.1 Regulating Rod Insertion Limits 3.2.2 AXIAL POWER SHAPING RODS (APSR) Insertion Limits 3.2.3 AXIAL POWER IMBALANCE Operating Limits 3.2.4 QUADRANT POWER TILT (QPT) 3.2.5 Power Peaking 3.3.1 Reactor Protection System (RPS) Instrumentation 3.4.1 RCS Pressure, Temperature, and Flow DNB limits 3.4.4 RCS Loops - MODES 1 and 2 3.9.1 Boron Concentration

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

Babcock & Wilcox Topical Report BAW-10179-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses" (the approved revision at the time the reload analyses are performed). The approved revision number shall be identified in the COLR.

Entergy Topical Report ENEAD-01-P, "Qualification of Reactor Physics Methods for the Pressurized Water Reactors of the Entergy System" (the approved revision at the time the reload analyses are performed). The approved revision number shall be identified in the COLR.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

ANO-1 5.0-21 Amendment No. 2-ta ,24&, 239

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements 5.6.6 Reactor Building Inspection Report Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Inspection Program shall undergo an engineering evaluation within 60 days of the completion of the inspection surveillance. The results of the engineering evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

5.6.7 Steam Generator Tube Inspection Reports A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,
b. Active degradation mechanisms found,
c. Nondestructive examination techniques utilized for each degradation mechanism,
d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
e. Number of tubes plugged during the inspection outage for each active degradation mechanism,
f. Total number and percentage of tubes plugged to date,
g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

ANO-1 5.0-22 Amendment No. ~,~, 239

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601 (c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation .

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP), or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously displays radiation dose rates in the area; or
2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or ANO-1 5.0-23 Amendment No. ~, 239

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:
1. All such door and gate keys shall be maintained under the administrative control of the shift manager, radiation protection manager, or his or her designee.
2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

ANO-1 5.0-24 Amendment No. ~,2-1-3, 239

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or
3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP, or equivalent, while in the area by means of closed circuit television, or personnel qualified in radiation protection procedures responsible for controlling personnel radiation exposure in the area and with the means to communicate with individuals in the area who are covered by such surveillance.
4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

ANO-1 5.0-25 Amendment No. 24-5, 239

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

ANO-1 5.0-26 Amendment No. ~, 239

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 239 TO FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT NO.1 DOCKET NO. 50-313

1.0 INTRODUCTION

By application dated October 22, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML073030541), as supplemented by letters dated January 12 and October 22, 2009 (ADAMS Accession No. ML090130124 and ML092990459, respectively),

Entergy Operations, Inc. (Entergy, the licensee), requested changes to the Technical Specifications (TS) for Arkansas Nuclear One, Unit 1 (ANO-1). The proposed changes adopt the U.S. Nuclear Regulatory Commission (NRC)-approved industryfTS Task Force (TSTF) change traveler TSTF-448, Revision 3, "Control Room Envelope Habitability." The availability of this TS improvement was published in the Federal Register on January 17,2007 (72 FR 2022),

as part of the consolidated line item improvement process.

The supplemental letters dated January 12 and October 22, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on December 18, 2007 (72 FR 71708).

The proposed amendment would (1) add a new license condition 2.c.(10) on the control room envelope habitability program, (2) revise the TS, 3.7.9, "Control Room Emergency Ventilation System (CREVS)," requirements related to the control room envelope habitability and (3) add the new TS 5.5.5, "Control Room Envelope Habitability Program," to the administrative controls section of the TSs.

On August 8, 2006, the commercial nuclear electrical power generation owners group TSTF submitted a proposed change, TSTF-448, Revision 3, to the improved standard technical specifications (STS) (NUREGs 1430-1434) on behalf of the commercial nuclear electrical power Enclosure 2

-2 Revisions 0, 1, and 2 were prior draft iterations). The approved TSTF-448, Revision 3, established more effective and appropriate TSs actions, surveillances, and administrative requirements to ensure habitability of the control room envelope (CRE) is maintained. ANO-1 is a Babcock and Wilcox designed pressurized-water reactor (PWR), the TSs for the plant are consistent with NUREG-1430, "Standard Technical Specifications - Babcock and Wilcox Plants:

Specifications."

The NRC alerted licensees in NRC Generic Letter (GL) 2003-01, "Control Room Habitability" (Reference 1), to NRC staff findings that facility TS surveillance requirements (SRs) for the Control Room Envelope Emergency Ventilation System (CREEVS) may not be adequate.

Specifically, results of American Society for Testing and Materials (ASTM) E741 (Reference 2) tracer gas tests to measure CRE unfiltered inleakage indicated that differential pressure surveillance testing is not a reliable method for demonstrating CRE boundary operability.

Licensees were requested to address their existing TSs as follows:

Provide confirmation that your technical specifications verify the integrity [Le.,

operability] of the CRE [boundary], and the assumed [unfiltered] inleakage rates of potentially contaminated air. If you currently have a differential pressure surveillance requirement to demonstrate CRE [boundary] integrity, provide the basis for your conclusion that it remains adequate to demonstrate CRE integrity in light of the ASTM E741 testing results. If you conclude that your differential pressure surveillance requirement is no longer adequate, provide a schedule for:

1) revising the surveillance requirement in your technical specification to reference an acceptable surveillance methodology (e.g., ASTM E741), and
2) making any necessary modifications to your CRE [boundary] so that compliance with your new surveillance requirement can be demonstrated.

If your facility does not currently have a technical specification surveillance requirement for your CRE integrity, explain how and at what frequency you confirm your CRE integrity and why this is adequate to demonstrate CRE integrity.

To promote standardization and to minimize the resources that would be needed to create and process plant-specific amendment applications in response to the concerns described in the generic letter, the industry and the NRC proposed revisions to CRE habitability system requirements contained in the STS, using the STS change traveler process. This effort culminated in Revision 3 to change traveler TSTF-448, "Control Room Envelope Habitability,"

which the NRC staff approved on January 17, 2007.

Consistent with the traveler as incorporated into NUREG-1430, the licensee proposed revising actions and SRs in TS 3.7.9, "Control Room Emergency Ventilation System (CREVS)," and adding a new administrative controls program, TS 5.5.5, "Control Room Envelope Habitability Program." The purpose of the changes is to ensure that CRE boundary operability is maintained and verified through effective surveillance and programmatic requirements, and that appropriate remedial actions are taken in the event of an inoperable CRE boundary.

Some editorial and plant-specific changes were incorporated into this safety evaluation resulting in minor deviations from the model safety evaluation text in TSTF-448, Revision 3.

- 3 The licensee identified changes to the TS Bases in Attachment 4 to its application. These changes were submitted for information only because changes to the TS Bases are controlled and approved by TS 5.5.14, "Technical Specifications (TS) Bases Control Program." However, because there are TS Bases accompanying the TSs in the NRC-approved TSTF-448, Revision 3, the NRC staff needs to review the identified TS Bases changes for this amendment to adopt TSTF-448, Revision 3, to make sure no problems are created by these changes.

2.0 REGULATORY EVALUATION

2.1 Control Room and Control Room Envelope NRC Regulatory Guide 1.196, "Control Room Habitability at Light-water Nuclear Power Reactors," Revision 0, May 2003 (Reference 4), uses the term "control room envelope (CRE)" in addition to the term "control room" and defines each term as follows:

Control Room: The plant area, defined in the facility licensing basis, in which actions can be taken to operate the plant safely under normal conditions and to maintain the reactor in a safe condition during accident situations. It encompasses the instrumentation and controls necessary for a safe shutdown of the plant and typically includes the critical document reference file, computer room (if used as an integral part of the emergency response plan), shift supervisor's office, operator wash room and kitchen, and other critical areas to which frequent personnel access or continuous occupancy may be necessary in the event of an accident.

Control Room Envelope: The plant area, defined in the facility licensing basis, that in the event of an emergency, can be isolated from the plant areas and the environment external to the CRE. This area is served by an emergency ventilation system, with the intent of maintaining the habitability of the control room. This area encompasses the control room, and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident.

NRC Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity At Nuclear Power Reactors," Revision 0, May 2003 (Reference 5), also contains these definitions, but uses the term CRE to mean both. This is because the protected environment provided for operators varies with the nuclear power facility. At some facilities, this environment is limited to the control room; at others, it is the CRE. In this safety evaluation, consistent with the proposed changes to the STS, the CRE will be used to designate both.

2.2 Control Room Envelope Ventilation System (CREVS)

The CREVS (the term used at ANO-1 for the CREEVS) provides a protected environment from which operators can control the unit, during airborne challenges from radioactivity, hazardous chemicals, and fire byproducts, such as fire suppression agents and smoke, during both normal and accident conditions.

- 4 The CREVS is designed to maintain a habitable environment in the control room envelope for 30 days of continuous occupancy after a Design Basis Accident (DBA) without exceeding a 5 rem total effective dose equivalent (TEDE).

The CREVS consists of two redundant trains, each capable of maintaining the habitability of the CRE. The CREVS is considered operable when the individual components necessary to limit operator exposure are operable in both trains. A CREVS train is considered operable when the associated:

  • Fan is operable;
  • High efficiency particulate air (HEPA) filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions;
  • Ductwork, valves, and dampers are operable, and air circulation can be maintained; and
  • CRE boundary is operable (the single boundary supports both trains).

The CRE boundary is considered operable when the measured unfiltered air inleakage is less than or equal to the inleakage value assumed by the licensing basis analyses of design-basis accident (DBA) consequences to CRE occupants.

2.3 Regulations Applicable to Control Room Habitability In Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," General Design Criteria (GDC) 1, 2, 3, 4, 5, and 19 apply to CRE habitability.

In Attachment 1 of a letter from Entergy dated August 28,2003 (ADAMS Accession No. ML032450205), "Response to Generic Letter 2003-01 Arkansas Nuclear One, Units 1 &2,"

Entergy confirmed that ANO-1 conforms to the following GDCs:

  • GDC 1, "Quality Standards and Records," requires that structures, systems, and components (SSCs) important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions performed.
  • GDC 2, "Design Bases for Protection Against Natural Phenomena," requires that SSCs important to safety be designed to withstand the effects of earthquakes and other natural hazards without loss of capability to perform their safety functions.
  • GDC 3, "Fire Protection," requires SSCs important to safety be designed and located to minimize the effects of fires and explosions.

- 5

  • GDC 4, "Environmental and Dynamic Effects Design Bases," requires SSCs important to safety to be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents (LOCAs).
  • GDC 5, "Sharing of Structures, Systems, and Components," requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, the orderly shutdown and cooldown of the remaining units.
  • GDC 19, "Control Room," requires that a control room be provided from which actions can be taken to operate the nuclear reactor safely under normal conditions and to maintain the reactor in a safe condition under accident conditions, including a LOCA. With regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE), as defined in 10 CFR 50.2 for the duration of the accident.

Prior to incorporation of TSTF-448, Revision 3, the STS requirements addressing CRE boundary operability resided only in TS 3.7.10, "Control Room Emergency Ventilation System (CREVS)." In this TS, the SR associated with demonstrating the operability of the CRE boundary requires verifying that one CREVS train can maintain a positive pressure relative to the areas adjacent to the CRE during the pressurization mode of operation at a makeup flow rate. Facilities that pressurize the CRE during the emergency mode of operation of the CREVS have similar SRs. Regardless, the results of ASTM E741 (Reference 2) tracer gas tests to measure CRE unfiltered inleakage at facilities indicated that the differential pressure surveillance (or the alternative surveillance at non-pressurization facilities) is not a reliable method for demonstrating CRE boundary operability. That is, licensees were able to obtain differential pressure and flow measurements satisfying the SR limits even though unfiltered inleakage was determined to exceed the value assumed in the safety analyses.

In addition to an inadequate SR, the action requirements of these specifications were ambiguous regarding CRE boundary operability in the event CRE unfiltered inleakage is found to exceed the analysis assumption. The ambiguity stemmed from the view that the CRE boundary may be considered operable but degraded in this condition, and that it would be deemed inoperable only if calculated radiological exposure limits for CRE occupants exceeded a licensing basis limit (i.e., as stated in GDC-19, even while crediting compensatory measures).

NRC Administrative Letter (AL) 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," (AL 98-10) states that "[t]he discovery of an improper or inadequate TS value or required action is considered a degraded or nonconforming condition,"

which is defined in NRC Inspection Manual Chapter 9900; see latest guidance in Regulatory Issue Summary (RIS) 2005-20 (Reference 3). AL 98-10 also states that "imposing administrative controls in response to an improper or inadequate TS is considered an acceptable short-term corrective action. The staff expects that, following the imposition of

-6 administrative controls, an amendment to the TS, with appropriate justification and schedule, will be submitted in a timely fashion."

Licensees that have found unfiltered inleakage in excess of the limit assumed in the safety analyses and have yet to either reduce the inleakage below the limit or establish a higher bounding limit through re-analysis, have implemented compensatory actions to ensure the safety of CRE occupants, pending final resolution of the condition, consistent with RIS 2005-20, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability." However, based on GL 2003-01 and AL 98-10, the NRC staff expects each licensee to propose TS changes that include a surveillance to periodically measure CRE unfiltered inleakage in order to satisfy Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(3), which requires a facility's TS to include SRs, which it defines as "[r]equirements 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 limiting conditions for operation will be met." (Emphasis added.)

The NRC staff also expects facilities to propose unambiguous remedial actions, consistent with 10 CFR 50.36(c)(2), for the condition of not meeting the limiting condition for operation (LCO) due to an inoperable CRE boundary. The action requirements should specify a reasonable completion time to restore conformance to the LCO before requiring a facility to be shut down.

This completion time should be based on the benefits of implementing mitigating actions to ensure CRE occupant safety and sufficient time to resolve most problems anticipated with the CRE boundary, while minimizing the chance that operators in the CRE will need to respond to abnormal operating occurrences or accident conditions.

Because the design of the plant is not being changed by the proposed amendment, the plant continues to meet GDCs 1, 2, 3, 4, and 5. Because the proposed addition to TS 5.5.11 requires the dose requirements in GDC 19 to be met, GDC 19 will also continue to be met by the plant.

2.4 Regulations Applicable to TS Changes In Section 50.36 of 10 CFR, "Technical specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TS. As stated in 10 CFR 50.36(c)(2)(i), the "[I]imiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility." Also the regulations establish surveillance requirements in 10 CFR 50.36(c)(3) state that "[s]urveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components will be maintained within safety limits, and that the limiting conditions for operation will be met."

2.5 Adoption of TSTF-448, Revision 3 Adoption of TSTF-448, Revision 3, will assure that the facility TS LCO for the CREVS is met by demonstrating unfiltered leakage into the CRE is within limits, Le., establish requirements for

-7 operability of the CRE boundary. In support of this surveillance, which specifies a test interval (frequency) based on Regulatory Guide 1.197, TSTF-448 also adds TS administrative controls to assure the habitability of the CRE between performances of the ASTM E741 test. In addition, adoption of TSTF-448 will establish specific required actions to be met in the event CRE unfiltered inleakage is found to exceed the analysis assumption.

The changes made by TSTF-448 to the STS requirements for the CREVS and the CRE boundary conform to 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(3). Their adoption will better assure that the ANO-1 CRE will remain habitable during normal operation and DBA conditions.

These changes are, therefore, acceptable from a regulatory standpoint.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed changes against the corresponding changes made to the STS by TSTF-448, Revision 3 (which the NRC staff has found to satisfy applicable regulatory requirements) as described above in Section 2.0. The emergency operational mode of the CREVS at ANO-1 pressurizes the CRE to minimize unfiltered air inleakage. The proposed changes are consistent with this design.

3.1 Proposed Changes to the Operating License 3.1.1 Condition Added to the License The following new condition is to be added to the license as condition number 2.c.(10):

(10) Upon implementation of Amendment 239 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.9.4, in accordance with Specifications 5.5.5.c.(i),

5.5.5.c.(ii), and 5.5.5.d, shall be considered met. Following implementation:

1. The first performance of SR 3.7.9.4, in accordance with Specification 5.5.5.c.(i), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
2. The first performance of the periodic assessment of CRE habitability, Specification 5.5.5.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
3. The first performance of the periodic measurement of CRE pressure, Specification 5.5.5.d, shall be within 15 months of the approval of TSTF 448. SR 3.0.2 will not be applicable to this first performance.

-8 SR 3.0.2 states that:

The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per ... " basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

Providing the statement "SR 3.0.2 will not be applicable to this first performance" in the license condition eliminates any confusion regarding whether SR 3.0.2 is applicable.

3.1.2 Proposed Changes to the TSs The proposed amendment would strengthen CRE habitability TS requirements by changing TS 3.7.9, and adding a new TS administrative controls program on CRE habitability in TS 5.5.5.

Accompanying the proposed TS changes are appropriate conforming technical changes to the TS Bases. The proposed revision to the Bases also includes editorial and administrative changes to reflect applicable STS Bases changes, which were made to improve clarity, conform to the latest information and references, correct factual errors, and achieve more consistency among the STS NUREGs. Except for plant-specific differences, all of these changes are consistent with STS as revised by TSTF-448, Revision 3.

The NRC staff compared the proposed TS changes to the STS and the STS markups and evaluations in TSTF-448. The NRC staff verified that differences from the STS were adequately justified on the basis of plant-specific design or retention of current licensing basis. The NRC staff also reviewed the proposed changes to the TS Bases for consistency with the STS Bases and the plant-specific design and licensing bases. Although approval of the Bases is not a condition for accepting the proposed amendment, approval of the identified TS Bases changes is through TS 5.5.14, "Technical Specifications (TS) Bases Control Program." This provides assurance that the licensee has established correct Bases for the TSs, is not changing the licensing basis of the plant without NRC review and approval, and will maintain the adequacy of the Bases. The proposed Bases for TS 3.7.9 refer to specific guidance in Nuclear Energy Institute (NEI) 99-03, "Control Room Habitability Assessment Guidance," Revision 0, dated June 2001 (Reference 6), which the NRC staff has formally endorsed, with exceptions, through Regulatory Guide 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors,"

dated May 2003 (Reference 4). Based on its review of the changes to the Bases, the NRC staff has not identified any problems with the changes.

-9 3.2 Editorial Changes The licensee proposed editorial changes to TS 3.7.9 to establish standard terminology, such as "control room envelope (CRE)" in place of "control room," except for the plant-specific name for the CREVS (plant-specific name for CREEVS), and "radiological, chemical, and smoke hazards (or challenges)" in place of various phrases to describe the hazards that CRE occupants are protected from by the CREVS. The licensee also proposed other editorial changes:

Verify [each CREVS train actuates] [or the control room isolates]

on an actual or simulated actuation signal.

the licensee has proposed:

Verify the CREVS automatically isolates the Control Room and switches into a recirculation mode of operation on an actual or simulated actuation signal.

The NRC staff has reviewed this deviation and concludes it meets the intent of the STS and is editorial in nature.

  • A typographical error in TS 3.7.9 Condition E was corrected by replacing "irradiate" with "irradiated."
  • The title of TS 5.5, "Administrative Controls," was misspelled and corrected.

These changes improve the usability and quality of the presentation of the TSs, have no impact on safety, and meet 10 CFR 50.36. Therefore, the NRC staff concludes that these changes are acceptable.

3.3 TS 3.7.9, CREVS The licensee proposed to revise the action requirements of TS 3.7.9 to establish that an inoperable CRE boundary, depending upon the location of the associated degradation, could cause just one, instead of both, CREVS trains to be inoperable. This is accomplished by (1) revising Condition A to exclude Condition B by adding the phrase "for reasons other than Condition B," and (2) revising Condition B to address one or more CREVS trains. Therefore, LCO 3.7.9 Conditions A and B have been revised to read as follows:

A. One CREVS train inoperable for reasons other than Condition B.

B. One or more CREVS trains inoperable due to inoperable CRE boundary in MODES 1, 2, 3, or 4.

- 10 This change clarifies TS action requirements to be taken in the event just one CREVS train is unable to ensure CRE occupant safety within licensing basis limits due to an inoperable CRE boundary. It enhances the usability of Conditions A and B with a presentation that is more consistent with the intent of the existing requirements. This change is an administrative change because it neither reduces nor increases the existing action requirements, and, therefore, is acceptable.

The licensee proposed to replace existing Required Action B.1, "Restore control room boundary to OPERABLE status," which has a 24-hour Completion Time, with Required Action B.1, to immediately "Initiate action to implement mitigating actions"; Required Action B.2, to "Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits," within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and Required Action B.3, to "Restore control room boundary to OPERABLE status," within 90 days.

The 24-hour Completion Time of new Required Action B.2 is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions as directed by Required Action B.1. The 90-day Completion Time of new Required Action B.3 is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. The 90-day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most anticipated problems with the CRE boundary. Therefore, proposed Required Actions B.1, B.2, and B.3 are acceptable.

In the emergency radiation mode of operation, the CREVS isolates unfiltered ventilation air supply intakes, filters the emergency ventilation air supply to the CRE, and pressurizes the CRE to minimize unfiltered air inleakage past the CRE boundary. The licensee proposed to delete the CRE pressurization SRs 3.7.9.4 and 3.7.9.5. These SRs require verifying that one CREVS train, operating in the emergency radiation mode, can maintain a positive pressure in the CRE relative to the adjacent areas during the pressurization mode of operation a pressure. The deletion of these SRs is proposed because measurements of unfiltered air leakage into the CRE at numerous reactor facilities demonstrated that a basic assumption of these SRs, an essentially leak-tight CRE boundary, was incorrect for most facilities. Therefore, meeting these SRs by achieving the required CRE pressure is not necessarily a conclusive indication of CRE boundary leak-tightness (Le., CRE boundary operability). In its response to GL 2003-01, dated August 28, 2003, the licensee reported that it had determined that the ANO-2 CRE pressurization surveillances, SRs 3.7.9.4 and 3.7.9.5, were inadequate to demonstrate the operability of the CRE boundary, and proposed to replace the SR with an inleakage measurement SR and a CRE Habitability Program in TS Section 5.5, in accordance with the approved version of TSTF-448. Based on the adoption of TSTF-448, Revision 3, the licensee's proposal to delete current SRs 3.7.9.4 and 3.7.9.5 is acceptable.

The proposed new CRE inleakage measurement SR would state, "Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program." Proposed TS 5.5.5, "Control Room Envelope Habitability Program," requires that the program include "Requirements for ... determining the unfiltered air inleakage past the CRE

- 11 boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197 .... n (Reference 5). This guidance references ASTM E741 (Reference 2) as an acceptable method for ascertaining the unfiltered leakage into the CRE. The licensee has proposed to follow this method. Therefore, the proposed CRE inleakage measurement SR is acceptable.

3.4 TS 5.5.5, Control Room Envelope Habitability Program The proposed administrative controls program TS is consistent with the model program TS in TSTF~448, Revision 3. In combination with SR 3.7.9.4, this program is intended to ensure the operability of the CRE boundary, which as part of an operable CREVS, will ensure that CRE habitability is maintained such that CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

A CRE Habitability Program TS acceptable to the NRC staff requires the program to contain the following elements:

  • Definitions of CRE and CRE boundary. This element is intended to ensure that these definitions accurately describe the plant areas that are within the CRE, and also the interfaces that form the CRE boundary, and are consistent with the general definitions discussed in Section 2.1 of this safety evaluation.

Establishing what is meant by the CRE and the CRE boundary will preclude ambiguity in the implementation of the program.

  • Configuration control and preventive maintenance of the CRE boundary. This element is intended to ensure the CRE boundary is maintained in its design condition. Guidance for implementing this element is contained in Regulatory Guide 1.196 (Reference 4), which endorsed, with exceptions, NEI 99-03 (Reference 6). Maintaining the CRE boundary in its design condition provides assurance that its leak-tightness will not significantly degrade between CRE inleakage determinations.
  • Assessment of CRE habitability at the frequencies stated in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0 (Reference 5), and measurement of unfiltered air leakage into the CRE in accordance with the testing methods and at the freq uencies stated in Sections C.1 and C.2 of Regulatory Guide 1.197. This element is intended to ensure that the plant assesses CRE habitability consistent with Sections C.1 and C.2 of Regulatory Guide 1.1.97. Assessing CRE habitability at the NRC accepted frequencies provides assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations. Determination of CRE inleakage using test methods acceptable to the NRC staff assures that test results are reliable for ascertaining CRE boundary operability. Determination of CRE inleakage at the NRC staff accepted

- 12 frequencies provides assurance that significant degradation of the CRE boundary will not occur between CRE inleakage determinations.

  • Measurement of CRE pressure with respect to all areas adjacent to the CRE boundary at designated locations for use in assessing the CRE boundary at a frequency of 18 months on a staggered test basis (with respect to the CREVS trains). This element is intended to ensure that CRE differential pressure is regularly measured to identify changes in pressure warranting evaluation of the condition of the CRE boundary. Obtaining and trending pressure data provides additional assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations.
  • Quantitative limits on unfiltered inleakage. This element is intended to establish the CRE inleakage limit as the CRE unfiltered infiltration rate assumed in the CRE occupant radiological consequence analyses of DBAs. Having an unambiguous criterion for the CRE boundary to be considered operable in order to meet LCO 3.7.9, will ensure that associated action requirements will be consistently applied in the event of CRE degradation resulting in inleakage exceeding the limit.

Consistent with TSTF-448, Revision 3, the program states that the provisions of SR 3.0.2 are applicable to the program frequencies for performing the activities required by program paragraph number c, parts (i) and (ii) (assessment of CRE habitability and measurement of CRE inleakage), and paragraph number d (measurement of CRE differential pressure). This statement is needed to avoid confusion. SR 3.0.2 is applicable to the surveillance that references the testing in the CRE Habitability Program. However, SR 3.0.2 is not applicable to Administrative Controls unless specifically invoked. Providing this statement in the program eliminates any confusion regarding whether SR 3.0.2 is applicable and meets 10 CFR 50.36.

Based on the above, the proposed change is acceptable.

Consistent with TSTF-448, Revision 3, proposed TS 5.5.5 states that (1) a CRE Habitability Program shall be established and implemented, (2) the program shall include all of the NRC staff required elements, as described above, and (3) the provisions of SR 3.0.2 shall apply to program frequencies. Therefore, TS 5.5.5, which is consistent with the model program TS approved by the NRC staff in TSTF-448, Revision 3, is acceptable and meets 10 CFR 50.36.

Based on the above, the proposed change is acceptable.

3.5 License Conditions In its October 22,2007, submittal, as revised by letters dated January 12 and October 22,2009, the licensee proposed the following new license condition 2.c.(10) regarding the initial performance of the new surveillance and assessment requirements:

(10) Upon implementation of Amendment 239 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage

- 13 as required by SR 3.7.9.4, in accordance with Specifications 5.5.5.c.(i),

5.5.5.c.(ii), and 5.5.5.d, shall be considered met. Following implementation:

1. The first performance of SR 3.7.9.4, in accordance with Specification 5.5.5.c.(i), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
2. The first performance of the periodic assessment of CRE habitability, Specification 5.5.5.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 3.0.2 will not be applicable to this first performance.
3. The first performance of the periodic measurement of CRE pressure, Specification 5.5.5.d, shall be within 15 months of the approval of TSTF 448. SR 3.0.2 will not be applicable to this first performance.

The licensee has proposed license conditions regarding the initial performance of the new surveillance and assessment requirements. The new license conditions adopted the conditions in Section 2.3 of the model application published in the Federal Register on January 17, 2007 (72 FR 2022). Plant-specific changes were made to these proposed license conditions. The proposed plant-specific license conditions are consistent with the model application and are, therefore, acceptable.

3.6 Summary Based on the above evaluation, the NRC staff concludes the proposed changes meet the technical requirements in the regulations that are discussed in Sections 2.0 and 3.0 of this safety evaluation. Based on this, the NRC staff further concludes that the proposed TS changes in the proposed amendment meet 10 CFR 50.36 and, therefore, the proposed amendment is acceptable.

The licensee identified changes to be made to the TS Bases that are associated with the TSs that are being changed in Attachment 4 to its application. The NRC does not approve these changes. The changes to the TS Bases to the TS Bases are made by the licensee through TS 5.5.14, "Technical Specifications (TS) Bases Control Program." However, the NRC has reviewed the identified changes to the TS Bases for this amendment and has no objection with these proposed changes.

- 14 4.0 REGULATORY COMMITMENTS In its October 22, 2007, submittal, as revised by letter dated January 12, 2009, the licensee made the following regulatory commitment:

  • Entergy will establish the Technical Specification (TS) Bases for TS 3.7.9, consistent with TSTF 448, Revision 3, as adopted with the applicable license amendment.

The commitment is scheduled to be implemented with issuance of this amendment. The NRC concludes this commitment is acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards considerations, and there has been no public comment on the finding published in the Federal Register on December 18, 2007 (72 FR 71708). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, on the basis of the considerations discussed above, that (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 Commission's 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.

8.0 REFERENCES

1. NRC Generic Letter 2003-01, "Control Room Habitability," dated June 12, 2003 (ADAMS Accession No. ML031620248).
2. American Society for Testing and Materials (ASTM) E741-00, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution," 2000.

- 15

3. NRC Regulatory Issue Summary 2005-20: Revision to Guidance Formerly Contained in NRC Generic Letter 91-18," Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability," dated September 26,2005 (RIS 2005-20).
4. Regulatory Guide 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," Revision 0, dated May 2003 (ADAMS Accession No. ML031490611).
5. Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003 (ADAMS Accession No. ML031490664).
6. Nuclear Energy Institute (NEI) 99-03, "Control Room Habitability Assessment Guidance," Revision 0, dated June 2001.
7. G. R. Ashley, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to Generic Letter 2003-01, Control Room Habitability," dated August 28, 2003 (ADAMS Accession No. ML032450205).

Principal Contributor: A. Wang Date: October 29, 2009

October 29, 2009 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 SR. 333 Russellville, AR 72802 SUB..IECT: ARKANSAS NUCLEAR ONE, UNIT NO.1 - ISSUANCE OF AMENDMENT RE:

ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE (TSTF) CHANGE TRAVELER TSTF-448, REVISION 3, "CONTROL ROOM ENVELOPE HABITABILITY" (TAC NO. MD7177)

Dear Sir or Madam:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 239 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit NO.1 (ANO-1). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 22, 2007, as supplemented by letters dated January 12 and October 22, 2009.

The amendment adds a new license condition 2.c.(10) on the control room envelope habitability program, revises TS requirements related to control room envelope habitability in TS 3.7.9, "Control Room Emergency Ventilation System (CREVS)," and adds a new administrative controls program, TS 5.5.5, "Control Room Envelope Habitability Program." These changes are consistent with the NRC-approved IndustryfTS Task Force (TSTF) change traveler TSTF-448, Revision 3, "Control Room Envelope Habitability." The availability of this TS improvement was published in the Federal Register on January 17, 2007 (72 FR 2022), as part of the consolidated line item improvement process.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, IRA by Kaly Kalyanam fori Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313

Enclosures:

1. Amendment No. 239 to DPR-51
2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrDorlDpr Resource RidsOgcRp Resource LPLIV RlF RidsNrrDorlLpl4 Resource RidsRgn4MailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsNrrLAJBurkhardt Resource RidsNrrDirsltsb Resource RidsNrrPMANO Resource ADAMS Accession No ML082540799 OFFICE NRR/LPL4/PM NRRlLPL4/LA DIRS/ITSB/BC OGC NRR/LPL4/BC NRRlLPL4/PM MMarkley AWang NAME AWang JBurkhardt RElliott Not Required KKalyanam for DATE 9/30/09 9/22/09 10/27/09 - 10/29/09 10/29/09 OFFICIAL RECORD COpy

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