ML15021A428
| ML15021A428 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/21/2015 |
| From: | Andrea George Plant Licensing Branch IV |
| To: | David Bice, Clark R Entergy Nuclear Operations |
| References | |
| TAC MF5292 | |
| Download: ML15021A428 (6) | |
Text
NRR-PMDAPEm Resource From: George, Andrea Sent: Wednesday, January 21, 2015 11:59 AM To: 'CLARK, ROBERT W'; 'BICE, DAVID B (ANO)'
Cc: 'PYLE, STEPHENIE L'
Subject:
Requests for Additional Information - ANO-1 P/T and LTOP Limits LAR to Extend P/T Curves to 54 EFPY - TAC NO. MF5292 Attachments: MF5292 - RAIs.pdf Mr. Clark and Mr. Bice, By letter dated November 21, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML14330A249), Entergy Operations, Inc. (Entergy the licensee), submitted a license amendment request (LAR) to revise the Technical Specifications (TS) for the Reactor Coolant System (RCS) Pressure and Temperature (P-T) Limits (TS 3.4.3), Pressurizer (TS 3.4.9), Pressurizer Safety Valves (TS 3.4.10), and Low Temperature Overpressure Protection (LTOP) System (TS3.4.11) at Arkansas Nuclear One, Unit 1 (ANO-1). The proposed revision would extend the applicability of the current limits from 31 EFPY to 54 EFPY.
In the course of its review, the NRC staff has determined that additional information is required to complete its evaluation. The attached requests for additional information (RAIs) were transmitted to D. Bice and R. Clark of your staff on January 14, 2015, and a clarification call was held on January 21, 2015. During the call, it was agreed that a response to these RAIs would be provided by Entergy by (at the latest) February 12, 2015.
Sincerely, Andrea George Project Manager Division of Operating Reactor Licensing U.S. Nuclear Regulatory Commission 301-415-1081 1
Hearing Identifier: NRR_PMDA Email Number: 1825 Mail Envelope Properties (6F9E3C9DCAB9E448AAA49B8772A448C5023A5CC1FE0F)
Subject:
Requests for Additional Information - ANO-1 P/T and LTOP Limits LAR to Extend P/T Curves to 54 EFPY - TAC NO. MF5292 Sent Date: 1/21/2015 11:58:34 AM Received Date: 1/21/2015 11:58:00 AM From: George, Andrea Created By: Andrea.George@nrc.gov Recipients:
"'PYLE, STEPHENIE L'" <SPYLE@entergy.com>
Tracking Status: None
"'CLARK, ROBERT W'" <RCLARK@entergy.com>
Tracking Status: None
"'BICE, DAVID B (ANO)'" <DBICE@entergy.com>
Tracking Status: None Post Office: HQCLSTR01.nrc.gov Files Size Date & Time MESSAGE 1269 1/21/2015 11:58:00 AM MF5292 - RAIs.pdf 96359 Options Priority: Standard Return Notification: No Reply Requested: Yes Sensitivity: Normal Expiration Date:
Recipients Received:
REQUESTS FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST TO REVISE REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE AND LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM LIMITS TO 54 EFFECTIVE FULL POWER YEARS ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 By letter dated November 21, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML14330A249), Entergy Operations, Inc. (Entergy the licensee),
submitted a license amendment request (LAR) to revise the Technical Specifications (TS) for the Reactor Coolant System (RCS) Pressure and Temperature (P-T) Limits (TS 3.4.3),
Pressurizer (TS 3.4.9), Pressurizer Safety Valves (TS 3.4.10), and Low Temperature Overpressure Protection (LTOP) System (TS3.4.11) at Arkansas Nuclear One, Unit 1 (ANO-1).
The proposed revision would extend the applicability of the current limits from 31 EFPY to 54 EFPY. The NRC staff has determined that additional information is required regarding the LAR.
EVIB-RAI-1 Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix G, requires that P-T limits be developed to bound all ferritic materials in the reactor pressure vessel (RPV).
Sections I and IV.A of 10 CFR Part 50, Appendix G specify that all ferritic reactor coolant pressure boundary (RCPB) components outside of the RPV must meet the applicable requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section III, Rules for Construction of Nuclear Facility Components.
As clarified in Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, issued October 2014 (ADAMS Accession No. ML14149A165), P-T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may define P-T curves that are more limiting because the consideration of stress levels from structural discontinuities (such as RPV inlet and outlet nozzles) may produce a lower allowable pressure.
In its LAR, the licensee stated that the ANO-1 RPV P-T limits were developed in accordance with the requirements of 10 CFR 50, Appendix G, using the analytical methods and flaw acceptance criteria ASME Code Section XI, Appendix G, and AREVA topical report BAW-10046A, Revision 2. BAW-10046A, Revision 2, includes a method for determining the P-T limits for nozzles, such as the RPV inlet and outlet nozzles. However, BAW-10046A does not provide guidance for evaluating the effects of neutron fluence on the nozzle nil-ductility reference transition temperature (RTNDT).
It is not clear, from the NRC staffs review of the LAR, whether the nozzle RTNDT was adjusted due to the effects of neutron irradiation.
a) Describe how neutron fluence was considered in the evaluation of the nozzles.
b) Provide RTNDT and fluence values for the limiting nozzle. The NRC staff requests the nozzle RTNDT and fluence in order to perform a confirmatory calculation for the nozzle.
EVIB-RAI-2 10 CFR 50 Appendix G, Paragraph IV.A.1.a requires that: reactor vessel beltline materials must have Charpy upper-shelf energy [USE] in the transverse direction for base material and along the weld for weld material according to the ASME Code, of no less than 75 ft-lb (102 J) initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68 J), unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, that lower values of Charpy upper-shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.
The licensee stated in its LAR that, with respect to USE and equivalent margins analysis (EMA), the current analysis remains bounding for the projected end of life fluence, except for the Upper Shell Plate 1 Material. However, for both heats (C5120-2 and C-5114-2) of Upper Shell Plate used, the most recent projected end of life fluence at 54 EFPY calculated per topical report BAW-2241P-A methodology following Cycles 21, 22 and 23 is less than the end of life fluence projected in topical report BAW-2251A for 48 EFPY. Also, the most recent projected end of life fluence for both Upper and Lower Shell Longitudinal Welds (Heat WF-18) is greater than the end of life fluence projected in BAW-2251A for 48 EFPY.
a) Describe how the current analysis for USE and EMA would not remain bounding for the Upper Shell Plate material considering that the end of life fluence has decreased.
b) Describe how the current analysis for USE and EMA would remain bounding for both Upper and Lower Shell Longitudinal Welds considering that the end of life fluence has increased.
c) If the current analysis for USE and EMA does not remain bounding for any material, provide an updated analysis for the material that is not bounded.
EVIB-RAI-3 10 CFR 50 Appendix G, Paragraph IV.A.2, Table 1 states that for normal operation heatup and cooldown, with the core not critical and vessel pressure greater than 20 percent of the system hydrostatic pressure, minimum temperature must be the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 120 degrees F. The NRC staffs safety evaluation related to License Amendment No. 188, dated March 14, 1997 (ADAMS Accession No. ML021270228), which approved the current P-T limits for ANO-1, indicated that the limiting flange region RTNDT is 60 degrees F. This value would result in a minimum temperature to exceed 20 percent of the preservice system
hydrostatic test (PSHT) pressure of 180 degrees F. This minimum temperature does not appear to be reflected in the licensees LAR, specifically, in the proposed revised P-T limits in of the LAR (TS Figures 3.4.3-1 ,3.4.3-2, and 3.4.3-3).
a) For limiting material in the closure flange region that is highly stressed by the bolt preload, provide the material identification, heat number, and RTNDT. If the limiting material RTNDT has changed since the current P-T limits submittal, provide the basis for the changes.
b) Describe how the heatup and cooldown curves in the licensee submittal comply with the 10 CFR 50, Appendix G, Table 1 requirement for normal operation heatup and cooldown with the core not critical and the vessel pressure greater than 20 percent of the system hydrostatic pressure.
EVIB-RAI-4 10 CFR Part 50, Appendix G requires that P-T limits be developed to bound all ferritic materials in the RPV. The P-T limits are calculated based on an initial RTNDT plus factors that account for margin and transition temperature shift due to irradiation effects. The licensees LAR includes initial RTNDT values and margin terms for plates and forgings which are substantially different from the initial RTNDT values which are presented in earlier licensee submittals (e.g. the ANO-1 License Renewal Application, which incorporates by reference topical report BAW-2251A, which contains these values).
a) Describe how the initial RTNDT and margin values were determined for the Lower Nozzle Belt Forging (heat 528360), Upper Shell Plate 1 (heat C5120-2), Upper Shell Plate 2 (heat C5114-2), Lower Shell Plate 1 (heat C5120-1), and Lower Shell Plate 2 (heat C5114-1).
b) Describe why it was determined that the method of determining initial RTNDT and margin values should be changed for this LAR.
SRXB-RAI-1 to the licensees LAR contains AREVA topical report ANP-3300, Arkansas Nuclear One Unit 1 Pressure-temperature Limits at 54 EFPY, Revision 1, dated November 2014. The plant-specific topical report includes fluence estimates for the lower nozzle belt forging. Figure 2-1 of ANP-3300 depicts this forging as located immediately below the outlet nozzle forging. Although the figure does not indicate the location of the core, it appears that the top of active fuel may be below the lower nozzle belt forging. ANP-3300, Revision 1, indicated that the fluence was calculated in accordance with topical report BAW-2241A, Revision 2, and that this method complies with Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001.
It should be noted that the guidance provided in RG 1.190 applies primarily to the region of the reactor vessel that directly surrounds the effective height of the active core.1 Furthermore, the qualification of BAW-2241A is not well established for determining fluence at or near nozzle locations.2 a) Demonstrate that the spatial modeling, synthesis, and boot-strap techniques for the transport calculations are adequate to produce reliable fluence estimates in the lower nozzle belt forging. Note the discussion in Section 3.1.1.2 of BAW-2241NP-A and address where, specifically, the lower nozzle belt forging is located in context of the (r,z) models.
b) RG 1.190 acknowledges the potential limitations of S8 angular quadrature for cavity fluence calculations; similar limitations for fluence calculations at locations where either (a) the transport pathway from the source to the target is longer, or (b) neutron streaming through the cavity could contribute a more significant portion of the total fluence, would be expected. The adequacy and potential limitations of this angular quadrature, and similar modeling difficulties are also briefly noted in BAW-2241NP-A.
Demonstrate that the angular quadrature chosen for the transport solution is adequate.
c) RTNDT and RTPTS (RTNDT based on end of life fluence) calculations include a 20% margin term in the fluence factor. The uncertainty requirements associated with RG 1.190 are consistent, in that benchmarking agreement within 20% is considered acceptable.
However, the NRC staff has reviewed the qualification database supporting BAW-2241A and determined that such agreement has not been established for the nozzle locations.
Provide a qualified estimate of the accuracy and uncertainty of the fluence methods for the nozzle locations. Demonstrate that the uncertainty in the fluence estimate is within 20% margin term included in the reference temperature calculations.
1 Note discussion in Regulatory Position 1.3.1, Discrete Ordinates Transport Calculation, which assumes a relatively weak axial variation of fluence Such relatively weak axial variation may not be the case at a region above the core/ The solution-based guidance for ex-core regions recommends, more generally, that a spatial mesh that ensures the flux in any group changes by less than a factor of ~2 between adjacent intervals should be applied 2
The uncertainty analysis presented in BAW-2241A includes a significant contribution of data from the Davis-Besse Cycle 6 ex-vessel dosimetry campaign.