ML15041A074
| ML15041A074 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/06/2015 |
| From: | AREVA |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15041A065 | List:
|
| References | |
| 1CAN021502 ANP-3300Q1NP, Rev. 0 | |
| Download: ML15041A074 (46) | |
Text
Attachment 2 to 1CAN021502 AREVA document ANP-3300QINP, "Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1,"
NON-PROPRIETARY
Controlled Document A
AREVA ANP-3300Q1 NP Response to Request for Additional Revision 0 Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 February 2015 AREVA Inc.
(c) 2015 AREVA Inc.
Controlled Document Copyright © 2015 AREVA Inc.
All Rights Reserved
Controlled 0( .-uien't AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqe i Nature of Changes Section(s)
Item or Page(s) Description and Justification 1 All Initial Issue
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page ii Contents Page 1.0 INT R O D UC T IO N .............................................................................................. 5 2.0 REQUESTS FOR ADDITIONAL INFORMATION .............................................. 5 2 .1 E V IB-R A I-1 .......................................................................................... .. 5 2.1.1 Statem ent of EVIB-RA I-1 ............................................................ 5 2.1.2 EVIB-RA I-1 Response: .............................................................. 7 2 .2 EV IB-R A I-2 ........................................................................................ . . 11 2.2.1 Statem ent of EVIB-RAI-2 .......................................................... 11 2.2.2 EVIB-RA I-2 Response: .............................................................. 12 2 .3 E V IB-RA I-3 .......................................................................................... 13 2.3.1 Statem ent of EVIB-RAI-3 .......................................................... 13 2.3.2 EVIB-RA I-3 Response: .............................................................. 14 2 .4 E V IB-R A I-4 .......................................................................................... 15 2.4.1 Statem ent of EVIB-RAI-4 .......................................................... 15 2.4.2 EVIB-RA I-4 Response: .............................................................. 15 2.5 S RX B -R A I-1 .......................................................................................... 20 2.5.1 Statement of SRXB-RAI-1 ........................................................ 20 2.5.2 SRXB-RA I-1 Response: ............................................................. 21 3.0 R E FE R E NC ES .............................................................................................. 36 APPENDIX A CLOSURE HEAD FORGING CMTR .......................................... 38 List of Tables Table 2-1 ART Values for ANO-1 Reactor Vessel Outlet Nozzle Forgings .............. 10 List of Figures Figure 2-1 Relationship between minimum Charpy impact energy (strong direction) measured at 10°F and (for the same materials) the initial RTNDT determined per ASME Section III, NB-2331 ............................ 10
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page iii Nomenclature (If applicable)
Acronym Definition ADAMS Agencywide Documents Access and Management system ANO-1/ANO1 Arkansas Nuclear One, Unit 1 ART Adjusted Reference Temperature ASME American Society of Mechanical Engineers B&W Babcock & Wilcox BWOG Babcock & Wilcox Owner's Group CFR Code of Federal Regulations CMTR Certified Material Test Report C/M Calculated versus Measured Cu Copper EMA Equivalent Margins Analysis Entergy Entergy Operations, Inc.
EOC End of Cycle EOL End of Life ft-lb foot-pound INF Inlet Nozzle Forging LNBF Lower Nozzle Belt Forging LAR License Amendment Request LR/LRA License Renewal/License Renewal Application NBF Nozzle Belt Forging 2
n/cm Neutrons/square centimeter (time-averaged neutron flux) n/cm 2/s Neutrons/square centimeter/second (neutron fluence rate/flux)
Ni Nickel NRC Nuclear Regulatory Commission ONF Outlet Nozzle Forging P-T Pressure - Temperature PTS Pressurized Thermal Shock PWR Pressurized Water Reactor RAI Request for Additional Information RIS Regulatory Issue Summary RG Regulatory Guide RG 1.99R2 Regulatory Guidel.99, Revision 2
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page iv Acronym Definition RTNDT Reference Temperature for Nil-Ductility temperature RTPTs Reference Temperature for Pressurized Thermal Shock RPV/RV Reactor Pressure Vessel/ Reactor Vessel Sn Symmetric quadrature with ordinate N (in discrete ordinate transport)
SE Safety Evaluation USE Upper Shelf Energy wt% Weight percentage
'F Degree Fahrenheit 0'i Standard deviation
Con trolled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 5
1.0 INTRODUCTION
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), Reference (1), 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 (TS 3.4.11) 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, Reference (3).
Information considered proprietary to AREVA in the following discussions is enclosed in brackets [ ].
2.0 REQUESTS FOR ADDITIONAL INFORMATION The NRC requests for additional information (RAIs) are reproduced from Reference (3) in Sections 2.1.1 through 2.5.1. Responses are in Sections 2.1.2 through 2.5.2.
2.1 EVIB-RAI-1 2.1.1 Statement of EVIB-RAI-1 Title 10 of the Code of FederalRegulations (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."
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqe 6 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 Xl, 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.
Controlled Document AREVA Inc. ANP-3300Q1 NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqe 7 2.1.2 EVIB-RAI-1 Response:
2.1.2.1 Part (a)
The projected fluence values for the inlet nozzles forgings (INF) and outlet nozzle forgings (ONF) are provided in the response to SRXB-RAI-1, Section 2.5.2 below. The INFs are projected to remain below 1E+17 n/cm 2 at 54 EFPY, so the effects of neutron irradiation on the INFs are not considered in the evaluation of the nozzles (RIS 2014-11). The RTNDT values for the ONFs were adjusted to account for the effects of neutron irradiation (see EVIB-RAI-ib Response, Section 2.1.2.2 below). The resultant adjusted reference temperature (ART) values are below the RTNDT value of 60 0F, which is used for the 60-year P-T limits analysis.
2.1.2.2 Part (b)
The predicted ART values for the outlet nozzle forgings (ONF) are shown in Table 2-1.
These values were calculated using Regulatory Guide 1.99, Revision 2 (RG 1.99R2),
unless where stated otherwise.
The peak wetted surface ONF fluence value at 54 EFPY is projected to be I ] . The applicability of the fluence attenuation equation in RG 1.99R2 at the ONF location is not defined. Therefore, the projected fluence at the crack tip of the postulated flaw originating at the nozzle corner (i.e., "1/4T") was conservatively assumed to be the peak projected wetted surface ONF fluence of [ ] .
The weight percent (wt%) copper (Cu) and nickel (Ni) values are the maximum values from chemistry ladle (one measurement per forging) and check analyses (two measurements per forging), which were documented on the certified material test reports (CMTRs). This is conservative relative to RG 1.99R2, which recommends using the mean value of the measurements.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 8 The ANO-1 ONFs were procured to an ASME Code year prior to 1971 and, therefore, sufficient Charpy data to determine the initial RTNDT value per ASME Section III NB-2331 (as required by 10 CFR 50, Appendix G) was not reported on the CMTRs. The ANO-1 ONF CMTRs do however report twelve Charpy measurements (six from each ONF) at a test temperature of +100 F, with impact energy results ranging from 90 to 122 ft-lbs. These Charpy specimens were oriented in the longitudinal (strong) direction.
Branch Technical Position 5-3 (BTP 5-3) of NUREG-0800 provides alternative methods for estimating initial RTNDT for plants with construction permits prior to 1973. Paragraph 1.1 (4) of BTP 5-3 states:
"If limited Charpy V-notch tests were performed at a single temperature to confirm that at least 30 ft-lbs was obtained, that temperature may be used as an estimate of the RTNDT provided that at least 45 ft-lbs was obtained if the specimens were longitudinally oriented. If the minimum value obtained was less than 45 ft-lbs, the RTNDT may be estimated as 20'F above the test temperature."
Since the minimum Charpy value measured at +10°F for the ANO-1 ONFs is 90 ft-lbs (which is above 45 ft-lbs), Paragraph 1.1 (4) of BTP 5-3 permits initial RTNDT to be estimated as +10°F. However, Paragraph 1.1 (4) of BTP 5-3 has been shown to not always be conservative, Reference (4). Therefore, an evaluation was performed to determine an appropriate bounding estimate of the initial RTNDT for the ANO-1 ONFs (as shown below).
The initial RTNDT value for the ANO-1 ONFs was estimated using measured Charpy values from the ANO-1 ONF CMTRs and an extensive population of ASME SA-508 Class 2 forging Charpy data, Reference (4).
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 9 The extensive population of ASME SA-508 Class 2 forging Charpy data from Reference (4) consists of forgings used by Babcock & Wilcox (B&W) to fabricated reactor pressure vessels. At the time of procurement, Charpy tests were performed on these forging at +100 F with specimens oriented in the strong (L-T) direction. After vessel fabrication, archive material from these same forgings was tested per the requirements of ASME Section III, NB-2331 (i.e., using Charpy specimens oriented in the weak (T-L) orientation in the determination of RTNDT). Since these forgings were tested using both the pre-1971 method (i.e., test temperature of +10°F with Charpy specimens oriented in the strong direction) and the current method (i.e., ASME Section III, NB-2331 with Charpy specimens oriented in the weak direction), the relationship between the two methods can be considered, as shown in Figure 2-1.
Note that after a recent review of the CMTR records supporting Reference (4), it could not be conclusively confirmed that all testing was performed using the current method (NB-2331) with Charpy specimens oriented in the weak direction. Therefore, all data with inconclusive records are not shown in Figure 2-1; all data points shown have supporting records that confirm that the Charpy tests met all requirements of ASME Section III, NB-2331 (including that the Charpy specimens were oriented in the weak direction). Also note that all data points in Figure 2-1 are from forgings located in B&W-designed reactor vessels.
The vertical "red" line on Figure 2-1 indicates the minimum Charpy value at +10°F for the ANO-1 ONFs of 90 ft-lbs. The sloped "red" line on Figure 2-1 bounds all data from Reference (4) (Note that only validated data from Reference (4) are shown in Figure 2-1, as discussed above). The intersection of these two lines (+40'F) represents a reasonable bounding estimate of the initial RTNDT for the ANO-1 ONFs based on their Charpy data tested at +100F.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 10 The calculated "1/4T" ART value of +59.4'F for the ANO-1 reactor vessel ONFs is bounded by the RTNDT value of +600 F used in the nozzle region to support the ANO-1 P-T limits analysis. Also note that the ONFs do not control any part of the submitted 60-year P-T curves.
Table 2-1 ART Values for ANO-1 Reactor Vessel Outlet Nozzle Forgings Figure 2-1 Relationship between minimum Charpy impact energy (strong direction) measured at IO°F and (for the same materials) the initial RTNDT determined per ASME Section III, NB-2331 60
!L 50 (fl40 Z 30 I* 20 S10 0
0 20 40 60 80 100 120 Minimum 10'F Charpy V-Notch Impact Energy, ft-lb Note that only validated data from Reference (4) are shown in Figure 2-1, as discussed above.
Controlled Doc~ument AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqge 11 2.2 EVIB-RAI-2 2.2.1 Statement of EVIB-RAI-2 10 CFR 50 Appendix G, Paragraph IV.A.I.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 Xl 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-2241 P-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.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 12 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.
2.2.2 EVIB-RAI-2 Response:
2.2.2.1 Part (a)
The following statement from the LAR should be removed:
"The current analysis remains bounding for the projected end of life fluence, except for the Upper Shell Plate 1 Material. The USE and EMA calculations also remain bounding for close to 54 EFPY as the fluence calculated per BAW-2241 P-A methodology following Cycles 21, 22, and 23 is lower, or only marginally higher, than the conservative fluence used in BAW-2251A. The copper content has also decreased."
The above statement from the LAR should be replaced with the following:
"The current USE and EMA analyses (BAW-2251A, Reference (6), Note: Appendix B of the Report contains BAW-2275 that addresses the EMA analyses) remain valid through 48 EFPY. For the EMA analysis, comparing the current projected 48 EFPY wetted surface fluence values of the limiting welds of ANO-1 with the EMA calculations reported in BAW-2275A, it can be shown that the EMA analyses for ANO-1 remains valid through 48 EFPY."
2.2.2.2 Part (b)
As stated in the response to EVIB-RAI-2 part (a), Section 2.2.2.1 above, the current USE and EMA analyses (BAW-2251A) remain valid through 48 EFPY.
2.2.2.3 Part (c)
Updates to the current USE and EMA calculations will be necessary in the evaluation period prior to the projected fluence exceeding the fluence on which the current USE and EMA calculations were based, as described on page 5 of Attachment 1 to Reference (1).
Controlled Document AREVA Inc. ANP-3300Q1 NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 13 2.3 EVIB-RAI-3 2.3.1 Statement of 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 licensee's LAR, specifically, in the proposed revised P-T limits in Attachment 3 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.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 14 2.3.2 EVIB-RAI-3 Response:
The current P-T limits for ANO-1, submitted in 1996 (SER issued in 1997) which utilized an RTNDT value of 60 degrees Fahrenheit (0F), was based on the original reactor vessel closure head. During the Fall 2005 outage at ANO-1, the new replacement reactor vessel closure head was installed. The RTNDT of this replacement reactor vessel closure head is [ I as discussed further below.
a) The material is ASME SA-508 Class 3, the heat number is [ ], and the RTNDT for the replacement reactor vessel head is [ ] (See
[ ]). The basis for the change is that the original reactor vessel closure head was replaced with a new reactor vessel closure head as explained above.
b) The heatup and cooldown curves in the submittal were developed considering the greater of the minimum temperature requirement per 10 CFR 50, Appendix G, Table 1 as well as the minimum required temperature considering a postulated outside surface flaw in closure head dome to flange region per BAW-10046A, Revision 2.
Specifically, to satisfy 10 CFR 50 Appendix G, Paragraph IV.A.2, Table 1, operating condition 2.b, the minimum required temperature is the RTNDT of the replacement reactor vessel closure head [ ] plus 120°F, which corresponds to [ ]. The minimum required temperature, based on postulation of an outside surface flaw, as referenced in Topical Report BAW-10046A, Revision 2, at the closure head dome-to-flange region that is highly stressed by the bolt preload is 80°F. As a result, the minimum required temperature used in the analysis to satisfy operating condition 2.b is 80°F.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqe 15 2.4 EVIB-RAI-4 2.4.1 Statement of 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 licensee's 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.
2.4.2 EVIB-RAI-4 Response:
2.4.2.1 Part (a)
The initial RTNDT value for the upper shell plate 2 (C5114-2) was determined using measured values according to ASME Section III Paragraph NB-2331 as required by 10 CFR Part 50, Appendix G. These measured values were reported in Appendix C of BAW-1440, Reference (7), which includes unirradiated Charpy data for C5114-2. Since the initial RTNDT was determined using measured values from Charpy specimens oriented in the transverse (weak) direction, the standard deviation (a*) is zero. The margin term (to which (aj is an input) was calculated in accordance with RG 1.99R2.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 16 The initial RTNDT value for the lower shell plate 2 (C5114-1) was determined using measured values according to ASME Section III Paragraph NB-2331 as required by 10 CFR Part 50, Appendix G. These measured values were reported in Appendix C of BAW-1440, Reference (7), which includes unirradiated Charpy data for C5114-1. Since the initial RTNDT was determined using measured values from Charpy specimens oriented in the transverse (weak) direction, the standard deviation (oi) is zero. The margin term (to which ai is an input) was calculated in accordance with RG 1.99R2.
The initial RTNDT value (+I 0 F) and the a( value (26.9°F) used for the upper shell plate 1 (C5120-2) are generic values determined consistent with the guidance in RG 1.99R2.
The generic values (mean and standard deviation) were established from a data set of measured initial RTNDT values from ASME SA-533 (ASTM A 533) Grade B Class 1 plates and ASME SA-302 (ASTM A 302) Grade B Modified plates. The margin term (to which ai is an input) was also calculated in accordance with RG 1.99R2.
The initial RTNDT value (+1°F) and the acvalue (26.90F) used for the lower shell plate 1 (C5120-1) are generic values determined consistent with the guidance in RG 1.99R2.
The generic values (mean and standard deviation) were established from a data set of measured initial RTNDT values from ASME SA-533 (ASTM A 533) Grade B Class 1 plates and ASME SA-302 (ASTM A 302) Grade B Modified plates. The margin term (to which oi is an input) was also calculated in accordance with RG 1.99R2.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 17 The initial RTNDT value (+27.5°F) and the ai value (12.9 0 F) used for the lower nozzle belt forging (heat 528360) are generic values determined from a data set of measured initial RTNDT values from ASTM A508 Class 2 forgings supplied by Ladish, which is the same supplier used for the ANO-1 lower nozzle belt forging (heat 528360). At the time of the calculation, it was believed that all initial RTNDT values in the data set used to support this generic value were determined per ASME Section III Paragraph NB-2331, but a recent review of the supporting CMTRs indicated that it could not be conclusively determined if the Charpy specimens were oriented in the weak (T-L) direction (as required per NB-2331). Therefore, the generic value of initial RTNDT value (+3°F) and ai value (31°F) documented in the ANO-1 License Renewal Application should be used in the 60-year analysis instead. This results in the ART values for the lower nozzle belt forging to increase slightly (<12°F), but this increase does not impact the submitted 60-year P-T curves. Also note that using the generic value of initial RTNDT value (+3°F) and 0i value (31°F) for the LNBF in the 54 EFPY Pressurized Thermal Shock (RTPTS) calculation does not impact the conclusion, which is that the LNBF RTPTS value remains below the screening criteria of 2700F for forgings.
2.4.2.2 Part (b)
For the lower nozzle belt forging (heat 528360), the initial RTNDT value (+3°F) and the ai value (31°F) used for the current 40-year P-T limit analysis should also be used for the 60-year P-T limit analysis. Using the initial RTNDT of +3°F and the ai of 31°F does not impact the submitted 60-year P-T curves (as discussed in the response to EVIB-RAI-4a, Section 2.4,2.1).
Controlled Document AREVA Inc. ANP-3300Ql NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqe 18 For the upper shell plate 2 (C5114-2), the initial RTNDT value (-10 0F) and the ai value (0°F) used for the current 40-year P-T curve analysis were based on measured Charpy values. Per ASME Section III, Paragraph NB-2331, measured Charpy values used to determine initial RTNDT are required to be obtained from Charpy specimens oriented in the transverse (weak) direction. Based on a recent review of the source document test reports, the Charpy specimen orientation used to determine the initial RTNDT (-10°F) was not conclusively determined. This did not have the potential to impact the current 40-year P-T curves because the current curves are limited by welds. Since the 60-year P-T curves will use BAW-2308, the weld ART values will decrease and the base metals become limiting. Therefore, measured Charpy data for the upper shell plate 1 (C5114-2) in the transverse direction was located to support the determination of the initial RTNDT.
For the lower shell plate 2 (C5114-1), the initial RTNDT value (0°F) and the a* value (00F) used for the current 40-year P-T curve analysis were based on measured Charpy values. Per ASME Section III, Paragraph NB-2331, measured Charpy values used to determine initial RTNDT are required to be obtained from Charpy specimens oriented in the transverse (weak) direction. Based on a recent review of the source document test reports, the Charpy specimen orientation used to determine the initial RTNDT (0°F) was not conclusively determined. This did not have the potential to impact the current 40-year P-T curves because the current curves are limited by welds. Since the 60-year P-T curves will use BAW-2308, the weld ART values will decrease and the base metals become limiting. Therefore, measured Charpy data for the lower shell plate 2 (C5114-1) in the transverse direction was located to support the determination of the initial RTNDT.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paae 19 For both the upper shell plate 1 (05120-2) and the lower shell plate 1 (C5120-1), the initial RTNDT value (-10°F) and the ai value (00 F) used for the current 40-year P-T curve analysis were based on measured Charpy values. Per ASME Section III, Paragraph NB-2331, measured Charpy values used to determine initial RTNDT are required to be obtained from Charpy specimens oriented in the transverse (weak) direction. Based on a recent review of the source document test reports, the Charpy specimen orientation used to determine the initial RTNDT values was not conclusively determined. This did not have the potential to impact the current 40-year P-T curves because the current curves are limited by welds. Since the 60-year P-T curves will use BAW-2308, the weld ART values will drop and the base metals will become limiting. Measured Charpy data conclusively in the transverse orientation were not located for the upper shell plate 1 (C5120-2) or the lower shell plate 1 (C5120-1). Therefore, the initial RTNDT value and the ai value were determined generically by a method consistent with the guidance in RG 1.99R2. The generic value for reactor vessel plate material of initial RTNDT value
(+10 F) and the a, value (26.9 0 F) have been used to support the P-T curves for other B&W unit reactor vessels with beltline plate materials.
C*J'ontrolied Doc;ument AREVA Inc. ANP-3300QlNP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 20 2.5 SRXB-RAI-1 2.5.1 Statement of SRXB-RAI-1 to the licensee's 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.
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.
Controlled Document AREVA Inc. ANP-3300QlNP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 21 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.
2.5.2 SRXB-RAI-1 Response:
2.5.2.1 Part (a)
The methodology used to determine neutron fluence is in accordance with AREVA's NRC approved fluence analysis methodology as described in BAW-2241P-A (or BAW-2241NP-A for the non-proprietary version), Reference (5). Fluence analysis performed is consistent with the guidance of Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." The spatial modeling, synthesis, and transport calculation techniques that comprise this methodology produce reliable fluence estimates in the lower nozzle belt forging (LNBF). Further description of this methodology is provided in the paragraphs below.
Controlled Document AREVA Inc. ANP-3300Q1 NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 22 The bottom portion of the LNBF is connected to the Upper Shell by circumferential weld WF-182-1 as shown on Figure 2-1 of ANP-3300P, Revision 1, Reference (2). The effective height of the active core extends just above weld WF-182-1 and is surrounded by the bottom portion of the LNBF. This location was included in the "Demonstration of Management of Aging Effects for the Reactor Vessel", B&W owners group (BWOG)
Generic License Renewal Program report BAW-2251A Reference (6) with irradiation damage considerations addressed through 48 EFPY. The LNBF thickness in the beltline region is the same as the upper shell. The forging extends up past the beltline and the thickness increases as part of the structural support of the inlet nozzle forgings (INF) and outlet nozzle forgings (ONF). [
As described in BAW-2241 P-A Reference (5), and references therein, the AREVA fluence analysis methodology synthesizes the results of two 2-dimensional radial (RT or Re) and axial (RZ) discrete ordinates transport (DORT) models which use the BUGLE cross-section library. The synthesis produces a 3-dimensional flux result from the two 2-dimensional DORT models. The flux values are integrated over time to determine the 3-dimensional neutron fluence values. Extensive benchmarking of the AREVA fluence analysis has shown that this method is unbiased and has a precision well within the RG 1.190 suggested one standard deviation of 20%.
As the NRC has noted, the methodology is generally applied to the pressure vessel, the capsule specimens attached to the internal core support structures, and the dosimetry holders in the cavity region. The benchmark of the calculated results is primarily associated with the beltline region of the reactor; that is the region surrounding the effective height of the active core. The Davis-Besse experiment did include some dosimeters above the effective height of the active core which are included in the benchmark comparisons.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 23 Boot-strap: Current models no longer use a bootstrap technique. In the past, a bootstrap model, which is a multi-part model, was used due to limitations associated with computer memory and capabilities. The bootstrap required the first model to represent the core region. The core's fission sources produce leakage sources at surfaces beyond the core. Restart cases used the leakage sources to extend the modeling to other areas of interest. Current computers are capable of running large cases and no longer require bootstrapping to determine the flux in extended areas of interest, both radial and axial.
Position 1.3.1 of RG 1.190 describes implementation of mesh densities in two or more "bootstrap" steps where computer-storage limitations prevent the implementation in a single-model representation. As such ceasing to use multiple models, bootstrapped together, does not represent a deviation in methodology; it simply represents an improvement in computational capabilities.
BAW-2241P-A, Reference (5), describes the A, B and C models that were used with appropriately small mesh increments (satisfying Regulatory Position 1.3.1 in RG 1.190).
However, [
Had computer capabilities been capable of handling a single-model representation with small mesh increments, there would be no bootstrapping of smaller models.
Consequently, the bootstraping is not an independent portion of the BAW-2241P-A methodology, but is a supplemental technique for improving the accuracy of large models. Since modern computer capabilities can now handle small mesh increments in large models, the supplemental technique is not needed.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 24 Results from the currently used large models have been confirmed by benchmark comparisons of calculated results to measured data for ANO-1 in beltline locations. The large models have unbiased results with a standard deviation that is consistent with the database established following the development of BAW-2241P-A, with smaller A, B and C models used at the time (late 1990s).
With the improved computing capabilities that are standard today, and consistent with RG 1.190 single-model representation, one large model encompassing all of the regions of interest is used for ANO-1. This eliminates the complexities associated with multiple smaller models bootstrapped together. Transport calculations are adequate to produce reliable fluence estimates in locations above the active height of the core (i.e., lower nozzle belt forging and outlet/inlet nozzles).
Spatial Modeling: The 2-dimensional RZ model has been expanded to include the upper and lower internal structures, and vessel and structural components such as the inlet and outlet nozzle connections. [
.] The method of determining the 3-dimensional flux results in the internal structures and other areas does not vary from the method used to determine the flux results for the beltline region.
[
]
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 25 I
Synthesis: The synthesis methodology does not change as a function of the extended RZ spatial modeling. The core support structure beyond the radial plane of the fuel is the "core barrel". This structure is a right-circular cylinder. Above the core barrel is the "core support shield". It is also a right-circular cylinder. In the radial plane beyond the core support structure are vessel shell components. These components are also right-circular cylinders or frustums. [
] The usage of the expanded synthesis model does not affect the calculated results in the axial direction beyond the active fuel/core.
The benchmark evaluations determined that the BAW-2241 P-A fluence methodology is unbiased. [
I The NRC's SE, in Section 2.3 and 3.3, summarizing the C/M database and uncertainty analysis methodology, and technical evaluations of the C/M data base uncertainties, respectively, provides validation.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 26 Details of the benchmark evaluations, which include standard deviations and calculated versus measured (C/M) ratios as discussed in Section 7.2 of BAW-2241P-A, Reference (5), are presented in Appendix A of BAW-2241 P-A. This demonstrates that a single standard deviation is appropriate for all dosimeters in the database, whose primary focus is the reactor vessel beltline. No variation in the accuracy of the calculation is expected when extending the model to include regions above the core.
Regarding the axial location of dosimetry above the active fuel for the Davis-Besse experiment, six dosimetry sets are of interest and warrant further description. [
- [
]
I
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Pa-qe 27
- [
- [
Summary: A portion of the LNBF is in the reactor vessel beltline and extends up with a change in thickness as part of the support of the INFs and ONFs. The approved methodology for predicting reactor vessel fluence, capsule specimens attached to the internal core support structures, and for dosimetry holders in the reactor cavity is described in BAW-2241 P-A, Reference (5). This methodology included a bootstrap technique, with synthesis and spatial modeling. The bootstrap technique is no longer required and thus is not used. The synthesis does not vary from the approved method; it is adequate regardless of the location that is synthesized to produce fluence estimates associated with the nozzle belt region. The spatial modeling does not vary from the methods used in the approved methodology, with spacing and intervals such that flux change between intervals is less than "2". Thus, the BAW-2241 P-A methodology is no different when modeling the beltline region or extended regions.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 28 2.5.2.2 Part (b)
An indication that a transport solution used an angular quadrature of too low an order is that the solution contains artificial "streaming rays", or regions of high flux that are not physical. One method of reducing the amplitude of these artificial streaming rays is increasing the order of the angular quadrature. Streaming rays generally emanate from regions of high flux to regions of low flux.
BAW-2241 P-A, Reference (5), reported the results of benchmarking calculated dosimetry results against measured dosimetry results for both in-core and ex-core locations. This benchmarking included the Davis-Besse cycle 6 cavity dosimetry experiment, which included extensive dosimeter locations in the reactor cavity. The locations for which results are reported for the Davis Besse cycle 6 experiment extend from the core mid-plane to above the top of active fuel. S8 symmetric quadrature was used for the transport calculation and proved to be adequate to achieve accurate results.
Multi-variable and parameter sensitivity evaluations indicated that the BAW-2241P-A fluence methodology is unbiased. Moreover, separating the dosimeter database into partial samples, where two sets were located around the midplane of the fuel (in-core and ex-core), and another set was located above the fuel, indicated that the standard deviations were related to the same database population. Thus, there is only one standard deviation for all dosimeters in the database.
The accuracy of the reactor cavity dosimetry results and the fact that there is no observable bias with respect to location provides strong evidence that the transport solution for the beltline model does not contain artificial streaming rays, which is one of the bases for concluding that S8 symmetric quadrature is adequate.
The mesh size in the transport solution may also contribute to the artificial streaming ray effect. The mesh size in the ANO-1 extended model uses the same methodology as the transport solution in the beltline model (See Section 2.5.2.1). Also, the ANO-1
Controlled Docuiment AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 29 extended model took advantage of improvements in computational capabilities and used a higher-order angular quadrature [ ] instead of S8 symmetric quadrature). The ANO-1 extended model is as good as or better than the beltline model with respect to the suppression of artificial streaming rays, so the accuracy of the transport solution has not been degraded by the choice of angular quadrature.
RG 1.190 Position 1.3.5 indicates that discrete ordinates calculations should use higher-order Sn calculations when off-midplane locations are analyzed and that a e-weighted difference model should be used. [
] , consistent with RG 1.190 and standard neutron transport theory. In addition, both the S8 and higher-order [ ] quadratures have been compared for ANO-1 in the reactor vessel beltline and nozzle areas of interest. As expected, the beltline results are equivalent to those found when developing the BAW-2241 P-A model. The [ ] results in the nozzle regions are accurate compared to the S8 results. Therefore the angular quadrature chosen for the transport solution is adequate and consistent with RG 1.190.
2.5.2.3 Part (c)
A qualified estimate of the accuracy and uncertainty of the fluence methods for nozzle locations was not completed for BAW-2241P-A, Reference (5). As indicated in Reference (3), pertinent regulatory guidance (i.e., RG 1.99R2 and RG 1.190) applies primarily to the region of the reactor vessel that directly surrounds the effective height of the active core, with only some guidance relative to excore regions. In addition, there is a lack of capsule information outside the reactor vessel beltline and limited data for dosimetry measurements in the reactor cavity at non-beltline elevations, such that a 95/95 confidence in the uncertainty is not feasible.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 30 Furthermore, it was demonstrated through an equivalent margins analysis that the nozzle-to-shell attachment welds are not limiting material with regard to neutron embrittlement damage and are not within the beltline region for a RV nozzle fluence of 1.5E18 n/cm 2 as described in BAW-2251A, Reference (6).
Projected fluence values at 54 EFPY are based on NRC approved Topical Report BAW-2241P-A, Revision 2, which complies with RG 1.190 as described in Section 3.0 of ANP-3300P, Revision 1, Reference (2), with a clarification. For reactor vessel beltline locations listed in Table 3-1 of ANP-3300P, Revision 1, Reference (2), the BAW-2241 P-A methodology is unbiased with an uncertainty that is less than the 20% (1a) of RG 1.190.
As defined in 10 CFR 50 Appendix G I1.F; reactor vessel beltline is the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.
For ANO-1 reactor vessel locations identified in ANP-3300P, Revision 1 Reference (2),
the beltline region includes the bottom portion of the LNBF, the weld connecting the upper shell to the LNBF (WF-182-1), the upper and lower shells, axial welds in these shells (WF-18) and the weld that connects the upper shell to the lower shell (WF-112).
These locations are shown on Figure 2-1 of ANP-3300P, Revision 1, Reference (2) as are nozzle forgings and welds.
Con~trolled Documnent AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 31 Flux and fluence at LNBF locations listed in Table 3-1 of ANP-3300P, Revision 1, Reference (2), as well as locations above them, were determined using the same semi-analytical calculation method described in BAW-2241P-A, Reference (5) with conservative approximations applied to calculate fluence values since there are insufficient measurements outside beltline elevations for reasonable assurance of a
'qualified' estimate. Therefore, calculated end of life (EOL) fluences at the LNBF and at locations above those locations are a conservative best-estimate.
Key factors leading to conservative fluence results in reactor vessel locations above the reactor vessel beltline are:
" The geometry and axial (RZ) model were updated to include locations of interest outside the reactor vessel beltline, including thickness changes in the LNBF just below the inlet and outlet nozzles. Components and structures inside the reactor cavity (e.g., outside the reactor vessel and above the nozzles) were considered for their impact on the fluence in those regions. Actual azimuthal/radial locations of the nozzles were represented. However, the maximum value from 0Qto 450 was used in the azimuthal/radial (RO) model for conservatism in the magnitude of the neutron fluence rate (time-averaged flux), and cumulative fluence calculated for locations above the weld connecting the upper shell to the LNBF (WF-182-1).
" The updated geometry was used with the same source and cross-sections used for reactor vessel beltline fluence determination per BAW-2241 P-A, as described in response to part a) (Section 2.2.2.1 above). Discrete Ordinate Transport (DORT) runs were performed and synthesized to calculate the flux for the cycles addressed in the most recent cavity dosimetry exchange fluence analysis for ANO-1. A reference cycle (Cycle 23) was selected that resulted in the highest flux at locations above the beltline.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 32 An approximation was used to conservatively estimate the fluence that had accumulated in locations above the beltline through that cycle, which had not been previously calculated (apart from license renewal and aging management evaluations). Using the outlet nozzle as an example, that is:
It is important to note that:
0 The I:sMt (Cl to 23) and 0ISM (C23) values are determined with BAW-2241 P-A methods and uncertainties adherent to RG 1.190.
o The ONo~zze (C23) value, calculated through the BAW-2241P-A semi-analytical methodology, is used to ratio the maximum reactor vessel inside surface fluence, within the beltline, to provide conservative estimates of Nozzle fluence.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 33 o As described on page 5 of Attachment 1 to Reference (1), "Fluence projections are checked each cycle and fluence analysis updated after every third cycle, when cavity dosimetry is exchanged."
As such, the magnitude of the maximum fluence in the reactor vessel beltline, calculated per BAW-2241P-A and adherent to RG 1.190, is used to conservatively estimate the cumulative fluence at locations outside the beltline with a ratio of the calculated neutron flux at the location to calculated neutron flux at the reactor vessel beltline location of the inside surface maximum fluence.
Projected 54 EFPY fluence values at the inside wetted surface of the ANO-1 reactor vessel are reported in Table 3-1, "Summary of ANO-1 RV Forging and Plate Data and Adjusted Reference Temperature Results at 54 EFPY," of ANP-3300P, Revision 1, Reference (2). Relative to USE/EMA, 48 EFPY and 54 EFPY fluence values are also listed on page 5 of Attachment 1 to Reference (1), along with the statement that "reactor 2
vessel locations not listed above have inside surface fluences below 1E+17 n/cm ."'
The values listed on page 5 of Attachment 1 to Reference (1) are "1/4T" values for direct comparison of projected 54 EFPY fluence values.
As noted in the RAI, there needs to be consistency between evaluations of RTNDT and RTPTs and the fluence evaluations. The following discusses specific fluence values associated with the nozzle regions and the relation to limiting materials that are the basis for the P-T limits.
Further evaluation has determined that the bottom of the outlet nozzle forging (ONF) weld to the LNBF is also projected to exceed 1E+17 n/cm 2 . The 48 EFPY and 54 EFPY wetted surface fluence values for the bottom of the ONF to LNBF forging weld and bottom of inlet nozzle forging (INF) to LNBF weld are:
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 34 As part of the BAW-2251A, Reference (6), evaluations, it was determined that nozzle welds are not limiting materials with respect to irradiation damage. Consequently it was concluded the nozzles, including welds, are not subject to surveillance. As noted in the safety evaluation for BAW-2251A, this conclusion has been accepted by the NRC and continues to be applicable to ANO-1 as described further on pages 5-7 of Attachment 1 to Reference (1) and Page 20 of the NRC SE for BAW-2251A. While the BAW-2251A determination is focused on USE/EMA, the same logic is applicable to RTPTS, and to RTNDT since the beltline definition is consistent between 10 CFR 50 Appendix G and 10 CFR 50.61.,
The conclusion that the nozzles, including nozzle welds, are not limiting materials relative to beltline locations is also consistent with the NRC Safety Evaluations for other utilities, such as References (8) and (9).
For ANO-1, the LNBF is addressed in Table 3-1, Summary of ANO-1 RV Forging and Plate Data and Adjusted Reference Temperature Results at 54 EFPY," of ANP-3300P, Revision 1, Reference (2), and was shown not to be limiting as clarified in the response to EVIB-RAI-4 part a) (Section 2.4.2.1) above. For the ONFs, the response to EVIB-RAI-1 part b) (Section 2.1.2.2 and Table 2-1) above confirms that the nozzles are not limiting even with a projected fluence above 1E+17n/cm 2 . The conservatively calculated ART does not affect the current 31 EFPY or requested 54 EFPY P-T curves, as the 60°F RTNDT remains valid. In conclusion, conservative best-estimate fluence values in nozzle locations are adequate for 54 EFPY irradiation shift considerations.
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 35 Summary: With the lack of measurements, for comparison of calculated to measured fluence, and guidance, for fluence determination as well as RTNDT and RTPTS outside the reactor vessel beltline, qualification of accuracy and uncertainty of fluence estimates in such locations is not feasible with reasonable confidence. Response to these RAIs provides demonstration that nozzle locations are not limiting with respect to irradiation damage in comparison to beltline materials. This demonstration included conservative best-estimates of accumulated fluence and calculated flux using models extended with the equivalent mesh/interval spacing as beltline fluence analyses per BAW-2241P-A and projected to end-of-life. Therefore, specific qualification and applicability of uncertainties or application of additional uncertainty is not warranted.
NOTE: The outlet nozzle is listed in Table 6-1, "Limiting Location Pressure Corrections Factors for ANO-1," of ANP-3300P, Revision 1, Reference (2), as a location addressed relative to temperature correction between uncorrected and corrected P-T limits. However, the reactor vessel outlet nozzles are not limiting materials for the 54 EFPY P-T Limit curves, even with conservative radiation shift considered, and are not within the beltline region at ANO-1.
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AREVA Inc. ANP-3300Q1 NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 36
3.0 REFERENCES
- 1. Entergy Letter 1CAN 111401, "License Amendment Request- Update the Reactor Coolant System Pressure and Temperature and Low Temperature Overpressure Protection System Limits Arkansas Nuclear One, Unit 1," November 21, 2014 (ADAMS Accession Number ML14330A249)
Unit 1 Pressure-Temperature Limits at 54 EFPY," November 2014, Attachment 4 to 1CAN 111401 (ADAMS Accession Number ML14330A250)
- 3. NRC "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," January 2015, TAC NO. MF5292
- 4. G. Troyer and M. DeVan, "An Assessment of Branch Technical Position 5-3 to Determine Unirradiated RTNDT for SA-508 Cl. 2 Forgings,"
Proceedings of the ASME 2014 Pressure Vessels & Piping Conference, July 20-24, 2014, Anaheim, California, USA.
- 5. AREVA Document BAW-2241P-A, Rev. 2, "Fluence and Uncertainty Methodologies," 2006 (ADAMS Accession Number ML031550365 for 3
submittal of proprietary version).
3 Revision 0 of BAW-2241P-A is for Babcock & Wilcox (B&W) reactor designs and includes the NRC Safety Evaluation Report (SER) applicable to ANO-1. Revisions 1 and 2 of BAW-2241P-A also contain the associated SERs and increase applicability to include Boiling Water Reactors (BWRs) and Westinghouse or Combustion Engineering (CE) reactors, respectively.
Controlled Document AREVA Inc. ANP-3300Q1 NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 37
- a. BAW-2241NP-A, Revision 2, "Fluence and Uncertainty Methodologies," April 30, 2006 (ADAMS Accession Number ML073310660).
- 6. M.A. Rinckel, J.R. Worsham III, et alia, "Demonstration of the Management of Aging Effects for the Reactor Vessel", BAW-2251-A, August, 1999.
- 7. BAW-1440, "Analysis of Capsule ANI-E from Arkansas Power & Light Company Arkansas Nuclear One, Unit 1 Reactor Vessel Materials Surveillance Program," April 1977.
- 8. NRC License Amendment, "Three Mile Island Nuclear Station, Unit 1 -
Issuance of Amendment RE: Revision to the Pressure and Temperature Limit Curves and the Low Temperature Overpressure Protection Limits (MF0424)," December 13, 2013 (ADAMS Accession Number ML13325A023)
- 9. NRC License Amendments, "Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Revised Pressure-Temperature Limits (TAC NOS. MF0763, MF0764, and MF0765)," December 13, 2013 (ADAMS Accession Number ML14041A093)
Controlled Document AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Paqe 38 APPENDIX A CLOSURE HEAD FORGING CMTR
Controlled Dc., nument AREVA Inc. ANP-3300Q1NP Revision 0 Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 Page 39 to 1CAN021502 Affidavit
AFFIDAVIT COMMONWEALTH OF VIRGINIA )
) ss.
CITY OF LYNCHBURG )
- 1. My name is Gayle Elliott. I am Manager, Product Licensing, for AREVA Inc.
(AREVA) and as such I am authorized to execute this Affidavit.
- 2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteria.
- 3. I am familiar with the AREVA information contained in ANP-3300Q1 P, Revision 0, entitled, "Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1," dated February 2015 and referred to herein as "Document."
Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.
- 4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is
requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."
- 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:
(a) The information reveals details of AREVA's research and development plans and programs or their results.
(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.
The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d) and 6(e) above.
- 7. In accordance with AREVA's policies governing the protection and control of infnrm*tian,nrnorietary nfnrmatinn cntained inrfhiD* 'n.urnnt has been made availahbl, nn a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8. AREVA policy requires that proprietary information be kept in a secured file or area-and-distributed-on-a-need--to--know-basis-.
- 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.
SUBSCRIBED before me this (A__
day of lft vv 2015.
V Sherry L. McFaden NOTARY PUBLIC, CQMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129 I-.