ML17163A412
| ML17163A412 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/08/2017 |
| From: | AREVA |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| L-17-190 ANP-3542Q1NP, Rev. 0 | |
| Download: ML17163A412 (31) | |
Text
Enclosure B Letter L-17-190 Davis-Besse Nuclear Power Station , Unit No. 1 (DBNPS)
AREVA Report AREVA Report ANP-354201 NP Revision 0, "Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals (Non-Proprietary) 30 pages follow
A AREVA Response to NRC Request for ANP-3542 1 Revision 0 o NP Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals June 2017 AREVA Inc.
(c) 2017 AREVA Inc.
ANP-354201 NP Revision 0 Copyright © 2017 AREVA Inc.
All Rights Reserved
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e i Nature of Changes Section(s)
Item or Page(s) Description and Justification 1 All Initial Issue
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e ii Contents Page
1.0 INTRODUCTION
............ ............... ............ ........ .... ......... .. ....... ... ... .... .... .. ....... ... 1-1 2.0 REQUESTS FOR ADDITIONAL INFORMATION .. ... .. .... ..... ..... ... ... ... ......... ...... . 2-1 2.1 RAl-1 ... ....... ...... ....... ....... ............. .. ... ....... .......... ..... .. ...... ... ...... ........ ........ 2-1 2.1.1 Statement of RAI ...... ......... .. ......... ...... ..... .... ....... ...... ............. .. ..... 2-1 2.1.2 Response to RAI ... ... ...... ........... ... .. .. .. ..... ........ .. .. ..... .. ........ .......... 2-1 2.2 RAl-2 ... ..... ..... ..... .. ......... .. ....... .. ............ ....... ................. ... .. ................ ... ... 2-2 2.2.1 Statement of RAI ... ..... ... .. ....... ........ ......... ..... ....... ... .. .... ........... .. ... 2-2 2.2.2 Response to RAI ...... .. .. .. ... ..... ... .. ....... .... .. .. .... .. .... ....................... . 2-2 2.3 RAl-3 ... .......... ..................... ...... .. .... ............ ....... .. ...... .......... ..... ............... 2-3 2.3 .1 Statement of RAI .... .... ...... .. ... ... .. ... ......... ... .. .. ............ ........... .. ... .. . 2-3 2.3.2 Response to RAI ..... ......... .... ..... ... ........... .. .. ... .... .... .. ........ ........... . 2-4 2.4 RAl-4 ....... ... .. ... .......... ..... ......... .. ........ ... .... .... .... .. ... .... .... .. .... .. .......... ...... 2-20 2.4 .1 Statement of RAI .. ... ... ........... .... ...... .. ... ............. ... ..... ............ .... . 2-20 2.4 .2 Response to RAI .. ...... ......... .... .. .. ............................................... 2-20 2.5 RAl-5 ... ..... .... .. ... ... .... ....... ........ .. ... .... .. ... ... ... .. ... .. ... ...... ...... ... ..... .. .. ... .... . 2-20 2.5 .1 Statement of RAI ... .. .. .... ...... .... ....... .. ................. .. ... .... ............... . 2-20 2.5.2 Response to RAI ..... .... .... ... ... ....... ....... .. .. ........ ... ... ............ ......... 2-21
3.0 REFERENCES
.......... ...... .... ........ ....... ........ ......... .. .. ........... .............. ........ ......... 3-1
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Add itional Information on Time-Limited Aging Ana lysis for the DB-1 Reactor Vessel Internals Pa e iii List of Tables Table 2.2-1 RVI Component Item Materials .... ... ..... .. .. .... .. .. ..... .. .................... ......... 2-2 Table 2.3-1 ] dpa Values ..... .................................................... ........... . 2-14 Table 2.3-2 Fluence Values ... ........ ....... ............ .. .. .. ... .. .... ... ..... .. .... ... ... ............. .... 2-17
AREVA Inc. ANP-354201 NP Revision 0 Response to NRG Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e iv List of Figures Figure 2.3-1 General Arrangement of Internals ... .. ........ ... .. ... ......... ..... .. .. .. ... .. .... .. .... 2-6 Figure 2.3-2 Sketch of r, z DORT Model .......... ....... .. .. .... ........... ..... ... ..... ........... .... 2-11 Figure 2.3-3 Sketch of r, 8 DORT Core Planar Model .. ... ....... ... .. .... ... .. ... .. ..... ..... ... 2-12 Figure 2.3-4 Representative [ ] Global Structural Model Dose Rate (dpa/sec) Contour Plot .. ..... ...... .. ... ..... ... ........... ..... ........ .. . 2-16
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Lim ited Aging Analysis for the DB-1 Reactor Vessel Interna ls Pa e v Nomenclature (If applicable)
Acronym Definition 10 CFR Title 10, Code of Federal Regulations AMP Aging Management Program ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox DB-1 <al Davis-Besse EFPY Effective Full Power Year FE NOC FirstEnergy Nuclear Operating Company LBB Leak Before Break LOCA Loss of Coolant Accident LRA License Renewal Appl ication MeV Mega (Million) electron Volt NRC U.S . Nuclear Regulatory Commission PWR Pressurized Water Reactor RAI Request for Additional Information Rv<al Reactor Vessel RVl (bl Reactor Vessel Internals SER Safety Evaluation Report Sy Yield Stress TLAA Time-Lim ited Aging Analysis UCN UFSAR Change Notice UFSAR Updated Final Safety Analysis Report a Used in the direct quotes by FENOC .
b Used in the direct quotes by NRC.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 1-1
1.0 INTRODUCTION
By letter dated January 23, 2017 (Agency-wide Documents Access and Management System (ADAMS) Package Accession No. ML17026A004 ), as supplemented by letter dated March 23 , 2017 (ADAMS Package Accession No. ML17086A019), FirstEnergy Nuclear Operating Company (FENOC , the licensee) submitted an evaluation by AREVA in response to Davis-Besse Nuclear Power Station , Unit No. 1, License Renewal Commitment No. 54 (Reference 6). The U.S . Nuclear Regulatory Commission (NRC) staff has determined that additional information is required to complete its review of the submittals.
Information considered proprietary to AREVA in the following discussions is enclosed in brackets [ ].
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analys is for the DB-1 Reactor Vessel Internals Pa e 2-1 2.0 REQUESTS FOR ADDITIONAL INFORMATION The NRC Requests for Additional Information (RAls) are reproduced from Reference (5) in Sections 2.1.1 , 2.2 .1, 2.3.1, 2.4 .1, and 2.5.1 . Responses are in Sections 2.1.2 , 2.2 .2, 2.3.2 , 2.4.2 , and 2.5 .2.
2.1 RA/-1 2.1.1 Statement of RAI For Alloy A-286 , Note 2 of Table 4-1 of the AREVA evaluation states: "Yield stress value based on 3 times (3x) design stress intensity at temperature (600 °F)." There is test data which supports the note for stainless steel piping ; however, Alloy A-286 is not used in piping . Justify use of Note 2 for Alloy A-286. For example , provide a comparison of the yield stresses based on 3 times design stress intensity at 600 °F and test-based yield stresses at similar temperatures.
2.1.2 Response to RAI The Note 2 statement regarding the Alloy A-286 material is in accordance with the ASME B&PV Code, 1968 Edition, Section Ill, Appendix II, "Basis for Establishing Design Stress Intensity Values. " It is stated in this Appendix that for bolting materials, the stress values in Table N-422 of the B&PV Code are based on 1/3 of the minimum specified yield stress at room temperature, or 1/3 of the yield stress at temperature up to a temperature of 800°F. The design stress intensity value, Sm , at 600°F reported in Table N-422 for SA-453, Grade 660 (Alloy A-286) is 27 ,000 psi . Therefore the yield stress (Sy) at temperature (i .e., 600°F) for this material is calculated as (27 ,000 psi
- 3) =
81,000 psi.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB- 1 Reactor Vessel Internals Pa e 2-2 2.2 RA/-2 2.2.1 Statement of RAI Table 4-2 of the AREVA evaluation lists the stress intensities for reactor vessel internal (RVI) components under different stress combinations . These stress intensities are compared to the unirradiated yield stress values in Table 4-1 of the AREVA evaluation to screen out RVls of no loss-of-ductility concern. Identify which material in Table 4-1 is applicable to each of the RVI components in Table 4-2 .
2.2.2 Response to RAI The material associated with each of the reactor vessel internal component items is provided in Table 2.2-1 below:
Table 2.2-1 RVI Component Item Materials Component Item Material Lower Grid Plate Type 304 Plenus (Plenum) Cover Type 304 Plenus (Plenum) Cylinder Reinforcing Plate CF8 Upper Guide Tubes Type 304L Upper Guide Tube Sectors Type 304L Core Support Shield Top Flange Type 304 Core Support Shield Lower Flange Type 304 Core Barrel Top Flange Type 304 Baffle Plates Type 304 Internal Bolts Core Barrel-Core Alloy X-750 I Alloy A-286 Support Shield Joint
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-3 2.3 RA/-3 2.3.1 Statement of RAI Section 5 of the AREVA evaluation describes how the neutron fluence estimates for several RVI component are used to ensure that relevant stress and strain margins are maintained throughout the period of extended operation . However, the fluence calculations do not provide (1) projected fluence values, (2) specific locations for the reported values , (3) an indication of uncertainty or margin of accuracy , or (4) if the chosen methodology is qualified for estimating fluence values at the various RVI component locations.
- a. Methods which are consistent with NRC Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence ," March 2001 (ADAMS Accession No ML010890301 ), may not be appropriate for estimating neutron fluence for some RVI components. For example, RG 1.190 permits representation of internal fuel assemblies in considerably less detail than peripheral assemblies, because the neutron flux on the reactor pressure vessel is primarily due to fuel at the core periphery. This is not the case for components such as the core barrel top flange. Explain how the methodology used in the AREVA evaluation is appropriate for estimating neutron fluence for RVI components in Category Item #2 and Category Item #3. The explanation should:
- i. Show that the current (r, z) spatial representation is sufficiently refined .
ii. Demonstrate that the detail represented in the (r, 8) model is adequate and produces a reliable neutron fluence estimate.
iii . Demonstrate that the chosen fluence methods are qualified for estimating neutron fluence at the various RVI component locations of interest.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-4 1v. Explain whether the neutron fluence values used have been augmented to account for any uncertainty associated with the calculation methods, nuclear data, or modeling accuracy.
- b. In Section 5.2 of the AREVA evaluation , for example, the estimated neutron fluence for a highly irradiated RVI component is used to determine the allowable irradiated yield stress for the component. If this fluence value is overestimated ,
then the large stress margin indicated may be nonconservative or, in the worst case, indicate that the allowable stress will be exceeded. Accounting for fluence uncertainty is needed to ensure that stress margins are maintained.
Provide neutron fluence (E > 1.0 MeV) values for the RVI components considered in the Category Item #2 assessment. Include irradiated yield stress margins after accounting for estimated RVI component nominal fluence values and their associated uncertainties.
- c. Provide neutron fluence (E > 1.0 MeV) values for the RVI components considered in the Category Item #3 assessment. Include uniform elongation margins after accounting for estimated RVI component nominal fluence values and their associated uncertainties.
2.3.2 Response to RAI As noted above (Section 2.3.1, "Statement of RAI"), RAl-3 has three parts: "a", "b" and "c". Moreover, part "a" not only requests an explanation of the fluence methodology but it also requests information associated with subparts "i" , "ii", "iii" and "iv".
In addition to the main part of "a" and the four subparts, requests "b" and "c" have two subparts . The first subpart for "b" and "c" involves the nominal fluence value and the upper and lower uncertainty range . The second subpart for "b" and "c" respectively involves the yield stress margins , and uniform elongation margins.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-5 The format for the response to RAl-3 is divided into the nine subparts outlined in the above discussion. The request associated with each of the nine subparts is stated above the response for each subpart.
a: Explain how the methodology used in the AREVA evaluation is appropriate for estimating neutron fluence for RVI components in Category Item #2 and Category Item #3.
The Category #2 and Category #3 RVI components range from [
] from the core. Figure 2.3-1 provides a schematic of the reactor vessel and the internal components. [
] As explained in AREVA's "Fluence and Uncertainty Methodologies" topical report ,[11 [
]
The AREVA evaluation associated with the neutron modeling is completely appropriate.
[
]
AREVA Inc. AN P-354201 NP Revision 0 Response to NRC Request for Additiona l Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-6 Figure 2.3-1 General Arrangement of Internals t'. J:l o E Ccre Support Shield 0.. Q) 0.. lll
- I lll (f) <l'.:
Q) u 0 Q)
UE (f) t J:l E
Q) lll i
J:l
. . . ... ... 0
- I a .. . .
'-+-+-+-...t--+-+-- Upp er Grid A.sse m bl y lll
<l'.: E Q)
+-' lll l
lll 0
l 0.. <l'.:
0.. Q)
- I "-
(f) "-
ru ......~-Therma l Shield Q) co u
0 Q) 0
. .. G " I * .. I " llJ I -41' n1-~- core Barrel u
+
Q)
~
0
~
ru _o c E Q)
Q)
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- <l'.:
( ote: som e com ponent i ems ar e ro at ed or cl ari ty}
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analys is for the DB-1 Reactor Vessel Internals Pa e 2-7 Addressing the appropriateness of AREVA's neutron modeling [
]
The benchmark database is the comparison of calculated results to measured data.
[
] Comparing the calculated results [
]
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for -Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-8
[
] ensures that the calculational methodology produces valid best-estimate results. Thus the methodology for evaluating each internal component's fluence with AREVA's synthesis, DORT, BUGLE , etc.
models is completely appropriate . [
]
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Ana lysis for the DB-1 Reactor Vessel Internals Pa e 2-9 Regulatory Guide 1 _99l 3l is based on a random fluence uncertainty of 20 percent,l21 but as the NRC noted , extrapolations of the fluence to 60 years would produce uncertainties greater than 20 percent.l4 l The increased uncertainties come from variations in cycle fuel-loadings and operation . The fluence uncertainties associated with subpart "b" for the yield stress margins and subpart "c" for the uniform elongation margins are presented below. [
]
a-i: Show that the current (r, z) spatial representation is sufficiently refined.
As suggested by the NRC ,l21 the (r, z) spatial representation may be shown to be sufficient by reducing the spatial increments by one-half and showing that the changes in the results are negligible. The spatial representation in AREVA's modeling has been demonstrated to be sufficient.l1l The fluence changes were neglig ible.
Figure 2.3-2 is a schematic of DORT (r, z) model. [
] This modeling is consistent with that approved by the NRC.l1l a-ii: Demonstrate that the detail represented in the (r, 8) model is adequate and produces a reliable neutron fluence estimate .
As suggested by the NRC ,l21 the (r, 8) spatial representation may be shown to be sufficient by reducing the spatial increments by one-half and showing that the changes in the resu lts are negligible. The spatial representation in AREVA's modeling has been demonstrated to be sufficient.l11 The fluence changes were negligible.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-10 Figure 2.3-3 is a schematic of DORT (r, 8) model. [
] This modeling is consistent with that approved by the NRC.[ 11 a-iii: Demonstrate that the chosen fluence methods are qualified for estimating neutron fluence at the various RVI component locations of interest.
As indicated when explaining how the AREVA evaluation is appropriate for estimating neutron fluence for RVI components in Category Item #2 and Category Item #3 , there are [
]
The explanation [
]
As demonstrated in AREVA's "Fluence and Uncertainty Methodologies" topical ,[1 1 the uncertainty in the fluence methods is [
]
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the 08-1 Reactor Vessel Internals Pa e2-11 Figure 2.3-2 Sketch of r, z DORT Model
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-12 Figure 2.3-3 Sketch of r, 9 DORT Core Planar Model
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e2-13
[
] The upper and lower range of fluence and dpa values are discussed in relation to the values noted in the tables for subparts "b" and "c".
a-iv: Explain whether the neutron fluence values used have been augmented to account for any uncertainty associated with the calculation methods, nuclear data, or modeling accuracy.
The fluence values per se have not been augmented to account for uncertainties in the calculational methods , nuclear data , or modeling . [
]
b-1: If fluence values are overestimated, then the large stress margins may be nonconservative. Provide neutron fluence (E > 1.0 MeV) values for the RVI components considered in the Category Item #2 assessment and include their associated uncertainties.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-14
[
] Table 2.3-1 gives the respective range of dpa values .
Table 2.3-1
[ ] dpa Values There are three uncertainty components related to the additional information in Table 2.3-1; they are discussed in the following paragraphs. The first uncertainty is
[
]
The values in Table 2.3-1 represent [
]
Likewise , [
]
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e2-15 As shown in Table 2.3-1 for the upper and lower dpa values , [
]
Table 2.3-1 represents [
]
b-2: Include irradiated yield stress margins after accounting for estimated RVI component nominal fluence values and their associated uncertainties.
Of the reactor internals components reviewed for Category Item #2 , only the [
] are expected to be highly irradiated [
] and are assessed using Category #2 using this projected fluence . For this criteria , higher irradiation results in higher irradiated yield strength margin relative to the faulted stress intensities. Therefore , the lower dose rates associated with the irradiation hardening of the [ ] is considered herein to assess the margin.
Calculated using the light water reactor (LWR) conversion factor of 1x10 n/cm (E > 1.0 MeV) = 15 dpa 22 2 c
(NUREG/CR-7027 (ANL-10/11 ), "Degradation of LWR Core Internal Materials due to Neutron Irradiation ,"
December 2010 (NRC Accession Number ML102790482).
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Req uest for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-16 Figure 2.3-4 Representative [ ] Global Structural Model Dose Rate (dpa/sec) Contour Plot The dose rates for the [ ] tend to vary in magnitude as shown in the representative contour plot (Figure 2.3-4 ). It is noted that this variability could result in a value of [ ] on the extreme top and bottom of the [ ] (per the response to RAI 3b-1 above). However, applying this low projected dose for the
[ ] to the curve in Figure 5-1 (Figure 13(a) of NUREG/CR-7027) of ANP-3542P[61 , results in an irradiated yield strength well above the faulted stress intensity and , therefore , the [ ] still experience irradiation hardening to where the material remains elastic with large margin to the irradiated yield stress , and the loss of ductility is acceptable.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-17 c-1: Provide neutron fluence (E > 1.0 MeV) values for the RVI components considered in the Category Item #3 assessment and include their associated uncertainties.
The RVI components considered in the Category Item #3 assessment are [
]
Table 2.3-2 Fluence Values The base value for the [
]
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Req uest for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-18 An upper fluence value of [
]
An upper fluence value of [
]
AREVA Inc. ANP-35420 1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e2-19 c-2: Include irradiated yield stress margins after accounting for estimated RVI component nominal fluence values and their associated uncertainties.
As part of the Category #3 assessment, the reactor internals components where the projected fluence is expected to be low enough such that neutron embrittlement is considered negligible include the [
] The projected fluences used in the Category #3 assessment for these components are [
] Per the response to RAI 3(c-1) above , [
] Applying these projected fluences to the trend lines in Figure 5-2 and Figure 5-3 of ANP-3542P[6 J indicates that these components will maintain greater than 20% uniform elongation at operating temperature (credited for 40 years in Appendix E of BAW-10008, *Part 1, Revision 1) and , therefore, the 8.6%
allowable strain specified in Appendix A of BAW-10008 , Part 1, Revision 1 continues to be met for these components.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Add itional Information on Time-Limited Aging Ana lysis for the DB-1 Reactor Vessel Internals Pa e 2-20 2.4 RAl-4 2.4.1 Statement of RAI Section 5.3 of the AREVA evaluation states, in part: "If the projected fluences of the remaining reactor vessel internals component items are applied to Figure 5-2 (Figure E-3 of BAW-10008 , Part 1, Revision 1),[11 the decrease in uniform elongation for the [specified RVI components] (all fabricated from Type 304 stainless steel) at both 572°F (300°C) and 752°F (400°C) (i .e., temperatures between which these component items would be expected to experience) is such that the 20 percent uniform elongation of irradiated material credited for 40 years in Appendix E of BAW-10008, Part1 ,
Revision 1 and the 8.6 percent allowable strain specified in Appendix A of BAW-10008, Part 1, Revision 1 is met for these component items."
Provide justification for this statement using the neutron fluence values provided in response to RAl-3 .
1 AREVA Document BAW-10008 Part 1, Revision 1, "Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake," June 1970.
2.4.2 Response to RAI See response to RAI 3c-2 (Section 2.3.2) above .
2.5 RAl-5 2.5.1 Statement of RAI Section 5.5 of the AREVA evaluation describes a further assessment based on recalculated faulted condition loads based on (1) asymmetric effects on pipe break loadings and (2) crediting leak-before-break by eliminating primary loop pipe breaks due from consideration .
Provide the stress intensities for the evaluated RVI components resulting from this further assessment. Identify the source document for the stress intensities (e .g. ,
provide references (title, date, etc.) to internal or industry documents).
AREVA Inc. ANP-35420 1NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 2-21 2.5.2 Response to RAI The stress intensity values for the components requiring further evaluation are given in the last paragraph of Section 5.5 . The source document for these stress intensity values are proprietary and can be reviewed in the AREVA offices in Lynchburg.
AREVA Inc. ANP-354201 NP Revision 0 Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the DB-1 Reactor Vessel Internals Pa e 3-1
3.0 REFERENCES
- 1. B & W Owners Group document# BAW-2241 P-A, "Fluence and Uncertainty Methodologies," J.R. Worsham Ill, et al ., Original ,
February, 1999; Revision 1, April, 2000 ; Revision 2, April, 2006 .
- 2. Regulatory Guide 1.190, "Calculational And Dosimetry Methods For Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission , March, 2001.
- 3. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement Of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission , May, 1988.
4 . B & W Owners Group document# BAW-2251-A, "Demonstration of the Management of Aging Effects for the Reactor Vessel ," J.R. Worsham Ill ,
et al., August, 1999.
- 5. Correspondence , Blake Purnell (U.S . NRC) to Phil Lashley (FENOC),
Subject:
Request for Additional Information Regarding Evaluation Submitted in Response to License Renewal Commitment No. 54 (CAC No. MF9126) (NRC ADAMS Accession #ML17129A411 ), May 9, 2017 .
- 6. ANP-3542P Revision 1, "Time-Limited Aging Analysis (TLAA) Regarding Reactor Vessel Internals Loss of Ductility for Davis-Besse Nuclear Power Station , Unit No. 1 at 60 Years ," March , 2017 .