1CAN060502, Arkansas, Unit 1 - License Amendment Request Response to NRC Request for Additional Information on ANO-1 Proposed Technical Specification Change to Steam Generator Tube Inservice Inspection Program
| ML051660294 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/08/2005 |
| From: | James D Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN060502 | |
| Download: ML051660294 (8) | |
Text
-r Entergy Operations, Inc.
-- Entergy 1448 S.R. 333 Russellville, AR 72802 Tel 479858-4619 Dale E. James Acting, Director, Nuclear Safety Assurance 1CAN060502 June 8, 2005 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
License Amendment Request Response to NRC Request for Additional Information on ANO-1 Proposed Technical Specification Change to Steam Generator Tube Inservice Inspection Program Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51
REFERENCES:
1 Entergy letter dated September 30, 2004, Proposed Technical Specification Change for Revision to ANO-1 Steam Generator Tube Inservice Inspection Program (1CAN090401) 2 Entergy letter dated April 26, 2005, Response to NRC Request for Additional Information on ANO-1 Proposed Technical Specification Change to Steam Generator Tube Inservice (1CAN040503)
Dear Sir or Madam:
On September 30, 2004 (Reference 1), Entergy requested NRC review and approval of a proposed Operating License amendment for Arkansas Nuclear One, Unit I (ANO-1) to replace the existing steam generator tube surveillance program with that contained in the Technical Specification Task Force (TSTF) -449. On March 23, 2005, the NRC Staff provided a Request for Additional Information (RAI) on the Entergy application. Entergy provided the response to these RAls on April 26, 2004 (Reference 2)..
In the original license amendment request (Reference 1), Entergy stated that the Bases of TSs 3.4.5, 3.4.6, and 3.4.7 (which referenced the Steam Generator Tube Surveillance Program that is being deleted) would be revised under the ANO-1 Bases Control Program. However, consistent with TSTF-449, the NRC staff has requested that these Bases pages be provided to the NRC for review. Therefore, the reference to Steam Generator Tube Surveillance Program is being deleted from these TS Bases pages and the marked-up Bases pages are being provided in Attachment 2.
6o00
1CAN060502 Page 2 of 2 In addition, an inconsistency was noted on proposed TS Page 5.0-11. There was one location where Mprimary-to-secondary" contained hyphens. In other occurrences, primary to secondary is unhyphenated. This is considered an editorial change and does not impact the original no significant hazards considerations. A corrected markup of this page is provided in Attachment 1.
This TS page markup replaces the markup of the page provided in Reference 1.
The proposed changes do not include any new commitments from that provided in Reference 1.
If you have any questions or require additional information, please contact Steve Bennett at 479-858-4626.
I declare under penalty of perjury that the foregoing is true and correct. Executed on June 8, 2005.
Sincerely, EJ/sab Attach ents:
- 1. Proposed Technical Specification Changes (mark-up)
- 2. Proposed Technical Specification Bases Changes (mark-up) cc: Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas W. Alexion MS O-7D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205
- W Attachment I 1CAN060502 Proposed Technical Specification Changes (mark-up)
p Programs and Manuals 5.5
- 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."
- 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 obiective of detecting flaws of any type (e.a., 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.
ANO-1 5,0-1 1 Amendment No. 245,
- 1 I Attachment 2 I CAN060502 Proposed Technical Specification Bases Changes (mark-up)
RCS Loops - MODE 3 B 3.4.5 LCO The purpose of this LCO is to require two loops to be available for heat removal thus providing redundancy. The LCO requires the two loops to be OPERABLE with the intent of requiring both SGs to be capable of transferring heat from the reactor coolant at a controlled rate. Forced reactor coolant flow is the preferred way to transport heat, although natural circulation flow is also acceptable under certain conditions. A minimum of one running RCP meets the LCO requirement for one loop in operation.
The Note permits a limited period of operation without RCPs. All RCPs may be removed from operation of
- 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for the transition to or from the Decay Heat Removal (DHR) System, and otherwise may be removed from operation for
- 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. During this condition, boron reduction with coolant at boron concentrations less than required to assure the SDM of LCO 3.1.1, is prohibited because an even concentration distribution throughout the RCS cannot be ensured. Core outlet temperature is to be maintained at least 100F below the saturation temperature so that: a) no vapor bubble may form and possibly cause a natural circulation flow obstruction; and b) pump restart criteria (which vary with pressure) are met.
In MODES 3, 4, and 5, it is sometimes necessary to stop all RCP or DHR pump forced circulation (e.g., change operation from one DHR train to the other, to perform surveillance or startup testing, to perform the transition to and from DHR System cooling, or to avoid operation below the RCP minimum net positive suction head limit). This is acceptable because the reactor coolant temperature can be maintained subcooled and boron stratification affecting reactivity control is not expected.
An OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is OPERABLE. To be considered OPERABLE, an RCP must be capable of being powered and able to provide forced flow if required. Similarly, an SG must be capable of transferring heat from the reactor coolant at a controlled rate and be in compliance with the Steam Generator Tube Supeillance Program.
APPLICABILITY In MODE 3, the heat load is lower than at power; therefore, one RCS loop in operation is adequate for transport and heat removal. A second RCS loop is required to be OPERABLE but not in operation for redundant heat removal capability.
Operation in other MODES is covered by:
ANO-1 B 3.4.5-2 Amendment No. 245,
RCS Loops - MODE 4 B 3.4.6 LCO (continued)
When the DHR pumps are stopped, no alternate heat removal path exists, unless the RCS and SGs have been placed in service in forced or natural circulation. The response of the RCS without heat removal through the DHR System or the SGs depends on the core decay heat load and the length of time that the DHR pumps are stopped. As decay heat diminishes, the effects on RCS temperature and pressure diminish. Without cooling by DHR, if the SGs are not capable of removing heat, higher heat loads will cause the reactor coolant temperature and pressure to increase at a rate proportional to the decay heat load. Because pressure can increase, the applicable system pressure limits (pressure and temperature (PIT) or low temperature overpressure protection (LTOP) limits) must be observed and forced DHR flow or heat removal via the SGs must be re-established prior to reaching the pressure limit.
An OPERABLE RCS loop consists of at least one OPERABLE RCP and an OPERABLE SG. To be considered OPERABLE, an SG must be capable of transferring heat from the reactor coolant at a controlled rate and-be iRGernpliane with the Steam Generator Tube Sur.ecillance Program.
Similarly for the DHR System, an OPERABLE DHR loop is comprised of the OPERABLE DHR pump(s) capable of circulating RCS fluid through the DHR heat exchanger(s) and back to the RCS. To be considered OPERABLE, a DHR pump must be capable of being powered and able to provide flow if required, and a DHR heat exchanger must be capable of transferring heat from the reactor coolant at a controlled rate.
A DHR loop may be considered OPERABLE during alignment and when aligned for low pressure injection if it is capable of being manually (locally or remotely) realigned to the DHR mode of operation and is not otherwise inoperable. This provision arises because of the dual requirements of the components that comprise the low pressure injection/decay heat removal system.
APPLICABILITY In MODE 4, this LCO applies because it is possible to remove core decay heat and to provide proper boron mixing with either the RCS loops and SGs or the DHR System.
Operation in other MODES is covered by:
ANO-1 B 3.4.6-2 Amendment No. 245,
RCS Loops - MODE 5, Loops Filled B 3.4.7 LCO (continued)
A DHR loop may be considered OPERABLE during alignment and when aligned for low pressure injection if it is capable of being manually (locally or remotely) realigned to the DHR mode of operation and is not otherwise inoperable. This provision arises because of the dual requirements of the components that comprise the low pressure injection/decay heat removal system.
An OPERABLE DHR loop is composed of an OPERABLE DHR pump and an OPERABLE DHR heat exchanger.
To be considered OPERABLE, DHR pumps must be capable of being powered and are able to provide flow if required. During performance of SR 3.8.1.7 or SR 3.8.1.8, the affected DHR pump may be considered OPERABLE even with the breaker "racked down" since placing this second pump in operation is a manual action. Similarly, an OPERABLE SG can perform as a heat sink when it has an adequate water level and is in-Gemplianrewith the Steam Generator Tube Suar'veillance PDrorem.
APPLICABILITY In MODE 5 with loops filled, forced circulation is provided by this LCO to remove decay heat from the core and to provide proper boron mixing. One loop of DHR provides sufficient circulation for these purposes.
Operation in other MODES is covered by:
- LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation-High Water Level" (MODE 6); and
- LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation-Low Water Level" (MODE 6).
ANO-1 B 3.4.7-3 Amendment No. 245,