Text

Arkansas, Unit 1 - Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, and to Request an Alternative to the ASME Code
	ML16088A181
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/25/2016
From:	Jeremy G. Browning; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	1CAN031604
	Download: ML16088A181 (53)
	v • d • e

s Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Vice President - Operations Arkansas Nuclear One 1CAN031604 10 CFR 50.90 10 CFR 50.55a March 25, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, and to Request an Alternative to the ASME Code Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TS) for Arkansas Nuclear One, Unit 1. The proposed change revises the TSs to eliminate Section 5.5, Inservice Testing Program. A new defined term, Inservice Testing Program, is added to the TS Definitions section. This request is consistent with TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing.

Pursuant to 10 CFR 50.55a(z), the application also proposes an alternative to the testing frequencies in the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, by adoption of approved Code Case OMN-20, Inservice Test Frequency, for the current 10 year Inservice Testing (IST) interval.

Attachment 1 provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides TS Bases pages marked up to show the associated TS Bases changes and is provided for information only.

Approval of the proposed amendment and relief request is requested by October 1, 2016. Once approved, the amendment shall be implemented within 90 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Arkansas state official.

1CAN031604 Page 2 of 2 No new commitments have been identified in this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 25, 2016.

If you have any questions or require additional information, please contact Stephenie Pyle at 479-858-4704.

Sincerely, ORIGINAL SIGNED BY JEREMY G. BROWNING JGB/dbb Attachments:

1. Description and Assessment of Technical Specification Changes

2. Proposed Technical Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

4. Proposed Technical Specification Bases Changes (Mark-Up) - Information Only

5. Description and Assessment of the Proposed Alternative to the ASME Code cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Ms. Andrea E. George MS O-8B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment 1 to 1CAN031604 Description and Assessment of Technical Specification Changes to 1CAN031604 Page 1 of 4 Description and Assessment of Technical Specification Changes

1.0 DESCRIPTION

The proposed change eliminates the Technical Specifications (TS), Section 5.5, Inservice Testing (IST) Program, to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code),

Case OMN-20, Inservice Test Frequency." A new defined term, "Inservice Testing Program," is added to TS Section 1.1, "Definitions. The proposed change to the TS is consistent with TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Entergy Operations, Inc. (Entergy), has reviewed the model safety evaluation (SE) addressed to the Technical Specifications Task Force in a letter dated December 11, 2015 (ADAMS Accession No. ML15317A071). This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-545. Entergy concluded that the justifications presented in TSTF-545, and the model safety evaluation prepared by the NRC staff are applicable to Arkansas Nuclear One, Unit 1 (ANO-1) and justify this amendment for the incorporation of the changes to the ANO-1 TS.

ANO-1 was issued a construction permit on December 6, 1968, and the provisions of 10 CFR 50.55a(f)(1) are applicable.

2.2 Variations No technical variances are proposed in this amendment request. The following bulleted items identify administrative-type variations. These differences to not result in any technical conflict with the intent of TSTF-545 or the associated model SE. Therefore, Entergy has concluded the following variations are both acceptable and permit NRC review of the proposed adoption of TSTF-545 under the Consolidated Line Item Improvement Process (CLIIP).

Some minor formatting clean-up is performed, where needed, to adjust available space on a page or in a table cell. With reference to TS 3.7.1, Table 3.7.1-1 is moved from TS Page 3.7.1-3 to Page 3.7.1-2 and the former is deleted in its entirety.

In a few examples, the ANO-1 Surveillance Requirement (SR) numbering does not match the numbering included in the TSTF-545 markup pages; however, Entergy verified the SRs are equivalent. TSTF-545 contains markup pages based on Revision 4 of NUREG 1430, Standard Technical Specifications - Babcock & Wilcox Plants, while the ANO-1 TSs are based on Revision 1 of NUREG 1430. With reference to SR 3.7.2.1, the IST reference also appears in the SR column of the table and, therefore, is capitalized consistent with the intent of TSTF-545. NUREG 1430, Revision 4, contains the IST reference in the Completion Time column only.

to 1CAN031604 Page 2 of 4 Entergy performed a search of the entire ANO-1 TSs for the key phrase inservice testing program and IST. In addition to that already described above, the TS Bases may have included other examples of this phrase, all of which are now capitalized as a defined term, consistent with TSTF-545. Also, the following reference (or similar) is added to applicable TS Bases (note that the TS Bases markups included in Attachment 4 are submitted for information only).

ASME OM Code - 2001 Edition through 2003 Addenda and Code Case OMN-20 (Inservice Test Frequency).

TSTF-545 deletes the IST program TS 5.5.8 and re-numbers all subsequent TS programs. This also impacts several TS Bases references. Entergy proposes to retain the TS 5.5.8 reference, now shown as DELETED, and not change the subsequent TS program numbers. These program numbers are referenced in a wealth of station procedures. By maintaining the current program numbering, excessive administrative burden is avoided. Based on this approach, several TSTF-545 TS Bases markup pages associated with the TSTF-545 program numbering are not included in Attachment 4 of this application.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Entergy Operations, Inc. (Entergy), requests adoption of the Technical Specification (TS) changes described in TSTF-545, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Arkansas Nuclear One, Unit 1 (ANO-1) TS.

The proposed change revises the TS Chapter 5, "Administrative Controls," Section 5.5, "Programs and Manuals," to delete the "Inservice Testing Program" specification.

Requirements in the Inservice Testing (IST) Program are removed, as they are duplicative of requirements in the American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code, as clarified by Code Case OMN-20, Inservice Test Frequency.

Other requirements in Section 5.5 are eliminated because the Nuclear Regulatory Commission (NRC) has determined their appearance in the TS is contrary to regulations. A new defined term, Inservice Testing Program, is added, which references the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.55a(f). Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed change revises TS Chapter 5, "Administrative Controls," Section 5.5, "Programs and Manuals," by eliminating the "Inservice Testing Program" specification.

Most requirements in the IST Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, Inservice Test to 1CAN031604 Page 3 of 4 Frequency. The remaining requirements in the Section 5.5 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, Inservice Testing Program, is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

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

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

Response: No.

The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged.

However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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

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

Response: No.

The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS Surveillance Requirement (SR) 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an to 1CAN031604 Page 4 of 4 assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.

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

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

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

5.0 PRECEDENCE Because TSTF-545, Revision 3, was approved in December 2015, time has not permitted licensees to apply, and gain approval for, adoption of TSTF-545. Therefore, other than TSTF 545 itself, no other precedence exists at this time.

Attachment 2 1CAN031604 Proposed Technical Specification Changes (mark-up)

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

ANO-1 1.1-3 Amendment No. 215,224,243,

Pressurizer Safety Valves 3.4.10 CONDITION REQUIRED ACTION COMPLETION TIME C. Required pressurizer C.1 Be in MODE 4 with RCS 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months safety valve inoperable in temperature 259 °F.

MODE 3 or MODE 4 with RCS temperature > 259 °F.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each required pressurizer safety valve is In accordance with OPERABLE in accordance with the INSERVICE the INSERVICE TESTING PROGRAMInservice Testing Program. TESTING Following testing, as-left lift settings shall be within PROGRAMInservic

+/- 1%. e Testing Program ANO-1 3.4.10-2 Amendment No. 215,254,

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ----------------------------NOTE---------------------------------

Not required to be performed in MODES 3 and 4.

Verify leakage from each RCS pressure isolation In accordance with check valve, or pair of check valves, as applicable, is the INSERVICE less than or equal to an equivalent of the Allowable TESTING Leakage Limit identified below at a differential test PROGRAMInservic pressure 150 psid. e Testing Program Pressure Isolation Allowable AND Check Valve(s) Leakage Limit Once prior to DH-14A 5 gpm entering MODE 2 DH-13A and DH-17 5 gpm total whenever the unit DH-14B 5 gpm has been in DH-13B and DH-18 5 gpm total MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2 Verify DHR System autoclosure interlock prevents 18 months the valves from being opened with a simulated or actual high RCS pressure signal.

SR 3.4.14.3 Verify DHR System autoclosure interlock causes the 18 months valves to close automatically with a simulated or actual high RCS pressure signal:

a. 340 psig for one valve; and

b. 400 psig for the other valve.

SR 3.4.14.4 Verify DHR System autoclosure interlock prevents 18 months the valves from being opened with a simulated or actual Core Flood Tank isolation valve not closed signal.

SR 3.4.14.5 Verify DHR System autoclosure interlock causes the 18 months valves to close automatically with a simulated or actual Core Flood Tank isolation valve not closed signal.

ANO-1 3.4.14-2 Amendment No. 215,

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) temperature > 350 °F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Reduce RCS temperature 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 350 °F.

C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE train available.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.2 Verify each ECCS pump's developed head at the test In accordance flow point is greater than or equal to the required with the developed head. INSERVICE TESTING PROGRAMInservi ce Testing Program ANO-1 3.5.2-1 Amendment No. 215,

Reactor Building Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each reactor building purge isolation valve is 31 days closed.

SR 3.6.3.2 -------------------------------NOTE-----------------------------

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

Verify each reactor building isolation manual valve 31 days and blind flange that is located outside the reactor building and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.

SR 3.6.3.3 -------------------------------NOTE-----------------------------

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

Verify each reactor building isolation manual valve Prior to entering and blind flange that is located inside the reactor MODE 4 from building and not locked, sealed, or otherwise MODE 5 if not secured, and required to be closed during accident performed within conditions is closed, except for reactor building the previous isolation valves that are open under administrative 92 days controls.

SR 3.6.3.4 Verify the isolation time of each automatic power In accordance with operated reactor building isolation valve is within the INSERVICE limits. TESTING PROGRAMInservic e Testing Program SR 3.6.3.5 Verify each automatic reactor building isolation 18 months valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

ANO-1 3.6.3-4 Amendment No. 215,253,

Reactor Building Spray and Cooling System 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify each reactor building spray manual, power 31 days operated, and automatic valve in each required flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.5.2 Operate each required reactor building cooling train 31 days fan unit for 15 minutes.

SR 3.6.5.3 Verify each required reactor building cooling train 31 days cooling water flow rate is 1200 gpm.

SR 3.6.5.4 Verify each required reactor building spray pump's In accordance with developed head at the flow test point is greater than the INSERVICE or equal to the required developed head. TESTING PROGRAMInservic e Testing Program SR 3.6.5.5 Verify each automatic reactor building spray valve in 18 months each required flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.5.6 Verify each required reactor building spray pump 18 months starts automatically on an actual or simulated actuation signal.

SR 3.6.5.7 Verify each required reactor building cooling train 18 months starts automatically on an actual or simulated actuation signal.

SR 3.6.5.8 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage ANO-1 3.6.5-3 Amendment No. 215,233,

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 ---------------------------------NOTE----------------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAMInservice Testing Program. Following TESTING testing, as-left lift settings shall be within +/- 1%. PROGRAMInservic e Testing Program Table 3.7.1-1 Allowable Power Level and RPS Nuclear Overpower Trip Allowable Value versus OPERABLE Main Steam Safety Valves RPS NUCLEAR Moved up from Page 3.7.1-3 MINIMUM NUMBER OF MAXIMUM ALLOWABLE OVERPOWER TRIP MSSVS OPERABLE POWER LEVEL ALLOWABLE VALUE (PER SG) (% RTP)

(% RTP) 6 85.7 89.9 5 71.4 74.9 4 57.1 59.9 3 42.8 44.9 2 28.5 29.9 ANO-1 3.7.1-2 Amendment No. 215,

MSSVs 3.7.1 Table 3.7.1-1 Allowable Power Level and RPS Nuclear Overpower Trip Allowable Value versus OPERABLE Main Steam Safety Valves MINIMUM NUMBER OF MAXIMUM ALLOWABLE RPS NUCLEAR MSSVS OPERABLE POWER LEVEL OVERPOWER TRIP (PER SG) (% RTP) ALLOWABLE VALUE

(% RTP) 6 85.7 89.9 5 71.4 74.9 4 57.1 59.9 3 42.8 44.9 2 28.5 29.9 ANO-1 3.7.1-3 Amendment No. 215

MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 --------------------------------NOTE-----------------------------

Only required to be performed in MODES 1 and 2.

Verify isolation time of each MSIV is within the limits In accordance with specified in the INSERVICE TESTING the INSERVICE PROGRAMInservice Testing Program. TESTING PROGRAMInservic e Testing Program SR 3.7.2.2 --------------------------------NOTE-----------------------------

1. Only required to be performed in MODES 1 and 2.

2. Not required to be met when SG pressure is

< 750 psig.

Verify each MSIV actuates to the isolation position 18 months on an actual or simulated actuation signal.

ANO-1 3.7.2-2 Amendment No. 215,

MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves 3.7.3 CONDITION REQUIRED ACTION COMPLETION TIME D. One Startup Feedwater D.1 Close or isolate Startup 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Control Valve in one or Feedwater Control Valve.

more flow paths inoperable. AND D.2 Verify Startup Feedwater Once per 7 days Control Valve is closed or isolated.

E. Two valves in the same E.1 Isolate affected flow path. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow path inoperable for one or more flow paths.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND F.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 --------------------------------NOTE-----------------------------

Only required to be performed in MODES 1 and 2.

Verify the isolation time of each MFIV, Main In accordance with Feedwater Block Valve, Low Load Feedwater the INSERVICE Control Valve and Startup Feedwater Control Valve TESTING is within the limits provided in the INSERVICE PROGRAMInservic TESTING PROGRAMInservice Testing Program. e Testing Program ANO-1 3.7.3-2 Amendment No. 215,

EFW System 3.7.5 CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months D. Two EFW trains D.1 -------------NOTE----------------

inoperable in MODE 1, 2, LCO 3.0.3 and all other or 3. LCO Required Actions requiring MODE changes are suspended until one EFW train is restored to OPERABLE status.

Initiate action to restore Immediately one EFW train to OPERABLE status.

E. Required EFW train E.1 Initiate action to restore Immediately inoperable in MODE 4. EFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each EFW manual, power operated, and 31 days automatic valve in each water flow path and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2 -------------------------------NOTE------------------------------

Not required to be performed for the turbine driven EFW pump, until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 750 psig in the steam generators.

Verify the developed head of each EFW pump at In accordance with the flow test point is greater than or equal to the the INSERVICE required developed head. TESTING PROGRAMInservic e Testing Program ANO-1 3.7.5-2 Amendment No. 215,

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 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 ProgramDELETED 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 applicable to the ASME Code For Operation and Maintenance (OM) of Nuclear Power Plants and applicable Addenda as follows:

ASME OM Code terminology Required Frequencies for performing for 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 At least once per 92 days months Semiannually or every 6 At least once per 184 days months Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 At least once per 731 days years

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program 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 OM Code shall be construed to supersede the requirements of any TS ANO-1 5.0-11 Amendment No. 215,239,240,250,

Attachment 3 1CAN031604 Revised Technical Specification Pages

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

ANO-1 1.1-3 Amendment No. 215,224,243,

Pressurizer Safety Valves 3.4.10 CONDITION REQUIRED ACTION COMPLETION TIME C. Required pressurizer C.1 Be in MODE 4 with RCS 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months safety valve inoperable in temperature 259 °F.

MODE 3 or MODE 4 with RCS temperature > 259 °F.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each required pressurizer safety valve is In accordance with OPERABLE in accordance with the INSERVICE the INSERVICE TESTING PROGRAM. Following testing, as-left TESTING PROGRAM lift settings shall be within +/- 1%.

ANO-1 3.4.10-2 Amendment No. 215,254,

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ----------------------------NOTE---------------------------------

Not required to be performed in MODES 3 and 4.

Verify leakage from each RCS pressure isolation In accordance with check valve, or pair of check valves, as applicable, is the INSERVICE less than or equal to an equivalent of the Allowable TESTING Leakage Limit identified below at a differential test PROGRAM pressure 150 psid.

AND Pressure Isolation Allowable Check Valve(s) Leakage Limit Once prior to entering MODE 2 DH-14A 5 gpm whenever the unit DH-13A and DH-17 5 gpm total has been in DH-14B 5 gpm MODE 5 for 7 days DH-13B and DH-18 5 gpm total or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2 Verify DHR System autoclosure interlock prevents 18 months the valves from being opened with a simulated or actual high RCS pressure signal.

SR 3.4.14.3 Verify DHR System autoclosure interlock causes the 18 months valves to close automatically with a simulated or actual high RCS pressure signal:

c. 340 psig for one valve; and

d. 400 psig for the other valve.

SR 3.4.14.4 Verify DHR System autoclosure interlock prevents 18 months the valves from being opened with a simulated or actual Core Flood Tank isolation valve not closed signal.

SR 3.4.14.5 Verify DHR System autoclosure interlock causes the 18 months valves to close automatically with a simulated or actual Core Flood Tank isolation valve not closed signal.

ANO-1 3.4.14-2 Amendment No. 215,

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) temperature > 350 °F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Reduce RCS temperature 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 350 °F.

C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE train available.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.2 Verify each ECCS pump's developed head at the In accordance with test flow point is greater than or equal to the the INSERVICE required developed head. TESTING PROGRAM ANO-1 3.5.2-1 Amendment No. 215,

Reactor Building Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each reactor building purge isolation valve is 31 days closed.

SR 3.6.3.2 -------------------------------NOTE-----------------------------

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

Verify each reactor building isolation manual valve 31 days and blind flange that is located outside the reactor building and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.

SR 3.6.3.3 -------------------------------NOTE-----------------------------

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

Verify each reactor building isolation manual valve Prior to entering and blind flange that is located inside the reactor MODE 4 from building and not locked, sealed, or otherwise MODE 5 if not secured, and required to be closed during accident performed within conditions is closed, except for reactor building the previous isolation valves that are open under administrative 92 days controls.

SR 3.6.3.4 Verify the isolation time of each automatic power In accordance with operated reactor building isolation valve is within the INSERVICE limits. TESTING PROGRAM SR 3.6.3.5 Verify each automatic reactor building isolation 18 months valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

ANO-1 3.6.3-4 Amendment No. 215,253,

Reactor Building Spray and Cooling System 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify each reactor building spray manual, power 31 days operated, and automatic valve in each required flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.5.2 Operate each required reactor building cooling train 31 days fan unit for 15 minutes.

SR 3.6.5.3 Verify each required reactor building cooling train 31 days cooling water flow rate is 1200 gpm.

SR 3.6.5.4 Verify each required reactor building spray pump's In accordance with developed head at the flow test point is greater than the INSERVICE or equal to the required developed head. TESTING PROGRAM SR 3.6.5.5 Verify each automatic reactor building spray valve in 18 months each required flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.5.6 Verify each required reactor building spray pump 18 months starts automatically on an actual or simulated actuation signal.

SR 3.6.5.7 Verify each required reactor building cooling train 18 months starts automatically on an actual or simulated actuation signal.

SR 3.6.5.8 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage ANO-1 3.6.5-3 Amendment No. 215,233,

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 ---------------------------------NOTE----------------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, as-left lift settings TESTING shall be within +/- 1%. PROGRAM Table 3.7.1-1 Allowable Power Level and RPS Nuclear Overpower Trip Allowable Value versus OPERABLE Main Steam Safety Valves RPS NUCLEAR MINIMUM NUMBER OF MAXIMUM ALLOWABLE OVERPOWER TRIP MSSVS OPERABLE POWER LEVEL ALLOWABLE VALUE (PER SG) (% RTP)

(% RTP) 6 85.7 89.9 5 71.4 74.9 4 57.1 59.9 3 42.8 44.9 2 28.5 29.9 ANO-1 3.7.1-2 Amendment No. 215,

MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 --------------------------------NOTE-----------------------------

Only required to be performed in MODES 1 and 2.

Verify isolation time of each MSIV is within the limits In accordance with specified in the INSERVICE TESTING PROGRAM. the INSERVICE TESTING PROGRAM SR 3.7.2.2 --------------------------------NOTE-----------------------------

1. Only required to be performed in MODES 1 and 2.

2. Not required to be met when SG pressure is

< 750 psig.

Verify each MSIV actuates to the isolation position 18 months on an actual or simulated actuation signal.

ANO-1 3.7.2-2 Amendment No. 215,

MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves 3.7.3 CONDITION REQUIRED ACTION COMPLETION TIME D. One Startup Feedwater D.1 Close or isolate Startup 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Control Valve in one or Feedwater Control Valve.

more flow paths inoperable. AND D.2 Verify Startup Feedwater Once per 7 days Control Valve is closed or isolated.

E. Two valves in the same E.1 Isolate affected flow path. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow path inoperable for one or more flow paths.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND F.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 --------------------------------NOTE-----------------------------

Only required to be performed in MODES 1 and 2.

Verify the isolation time of each MFIV, Main In accordance with Feedwater Block Valve, Low Load Feedwater the INSERVICE Control Valve and Startup Feedwater Control Valve TESTING is within the limits provided in the INSERVICE PROGRAM TESTING PROGRAM.

ANO-1 3.7.3-2 Amendment No. 215,