Response to Request for Additional Information on Reactor NON … · 2015-02-19 · Controlled...
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Attachment 2 to
1 CAN021502
AREVA document ANP-3300QINP,
"Response to Request for Additional Information on ReactorCoolant System Pressure/Temperature and Low Temperature
Overpressure Protection System Limits to 54 EFPY forArkansas Nuclear One, Unit 1,"
NON-PROPRIETARY
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AAREVA
Response to Request for AdditionalInformation on Reactor Coolant SystemPressure/Temperature and LowTemperature Overpressure ProtectionSystem Limits to 54 EFPY for ArkansasNuclear One, Unit 1
ANP-3300Q1 NPRevision 0
February 2015
AREVA Inc.
(c) 2015 AREVA Inc.
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Copyright © 2015
AREVA Inc.All Rights Reserved
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Paqe i
Nature of Changes
Section(s)Item or Page(s) Description and Justification1 All Initial Issue
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Page iiContents
Page
1.0 IN T R O D U C T IO N .............................................................................................. 5
2.0 REQUESTS FOR ADDITIONAL INFORMATION .............................................. 5
2 .1 E V IB -R A I-1 .......................................................................................... . . 52.1.1 Statem ent of EVIB-RA I-1 ............................................................ 52.1.2 EVIB-RA I-1 Response: .............................................................. 7
2 .2 E V IB -R A I-2 ........................................................................................ . . 112.2.1 Statem ent of EVIB-RAI-2 .......................................................... 112.2.2 EVIB-RA I-2 Response: .............................................................. 12
2 .3 E V IB -RA I-3 .......................................................................................... 132.3.1 Statem ent of EVIB-RAI-3 .......................................................... 132.3.2 EVIB-RA I-3 Response: .............................................................. 14
2 .4 E V IB -R A I-4 .......................................................................................... 152.4.1 Statem ent of EVIB-RAI-4 .......................................................... 152.4.2 EVIB-RA I-4 Response: .............................................................. 15
2.5 S R X B -R A I-1 .......................................................................................... 202.5.1 Statement of SRXB-RAI-1 ........................................................ 202.5.2 SRXB-RA I-1 Response: ............................................................. 21
3.0 R E FE R E N C 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 (strongdirection) measured at 1 0°F and (for the same materials) the initialRTNDT determined per ASME Section III, NB-2331 ............................ 10
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Page iii
Nomenclature
(If applicable)
AcronymADAMS
ANO-1/ANO1ARTASMEB&WBWOGCFRCMTRC/MCuEMAEntergy
EOCEOLft-lbINFLNBFLARLR/LRANBFn/cm
2
n/cm 2/s
NiNRCONFP-T
PTSPWR
RAIRISRGRG 1.99R2
DefinitionAgencywide Documents Access and Management systemArkansas Nuclear One, Unit 1Adjusted Reference TemperatureAmerican Society of Mechanical EngineersBabcock & WilcoxBabcock & Wilcox Owner's GroupCode of Federal Regulations
Certified Material Test Report
Calculated versus MeasuredCopperEquivalent Margins AnalysisEntergy Operations, Inc.End of CycleEnd of Lifefoot-poundInlet Nozzle ForgingLower Nozzle Belt ForgingLicense Amendment RequestLicense Renewal/License Renewal ApplicationNozzle Belt ForgingNeutrons/square centimeter (time-averaged neutron flux)Neutrons/square centimeter/second (neutron fluence rate/flux)NickelNuclear Regulatory CommissionOutlet Nozzle ForgingPressure - Temperature
Pressurized Thermal ShockPressurized Water ReactorRequest for Additional InformationRegulatory Issue SummaryRegulatory GuideRegulatory Guidel.99, Revision 2
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Page iv
AcronymRTNDT
RTPTs
RPV/RVSn
SEUSEwt%'F0'i
DefinitionReference Temperature for Nil-Ductility temperatureReference Temperature for Pressurized Thermal ShockReactor Pressure Vessel/ Reactor VesselSymmetric quadrature with ordinate N (in discrete ordinate transport)
Safety EvaluationUpper Shelf EnergyWeight percentageDegree FahrenheitStandard deviation
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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 Federal Regulations (10 CFR) Part 50, Appendix G, requires that
P-T limits be developed to bound all ferritic materials in the reactor pressure vessel
(RPV). Sections I and IV.A of 10 CFR Part 50, Appendix G specify that all ferritic reactor
coolant pressure boundary (RCPB) components outside of the RPV must meet the
applicable requirements of American Society of Mechanical Engineers Boiler and
Pressure Vessel Code (ASME Code), Section III, "Rules for Construction of Nuclear
Facility Components."
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Paqe 6As 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.
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Paqe 72.1.2 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/cm2 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 600F,
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.
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Page 8The 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 +100F, 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).
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Page 9The 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 +100F 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.
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Page 10The calculated "1/4T" ART value of +59.4'F for the ANO-1 reactor vessel ONFs is
bounded by the RTNDT value of +600F 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 80Minimum 10'F Charpy V-Notch
Impact Energy, ft-lb
100 120
Note that only validated data from Reference (4) are shown in Figure 2-1,
as discussed above.
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Paqge 112.2 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.
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Page 12c) 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).
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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.
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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.
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Paqe 152.4 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.
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Page 16The 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 (+I0F) 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 ac value (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.
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Page 17The initial RTNDT value (+27.5°F) and the ai value (12.9 0F) 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).
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Paqe 18For 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.
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Paae 19For 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 (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 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
(+10F) and the a, value (26.90F) have been used to support the P-T curves for other
B&W unit reactor vessels with beltline plate materials.
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Page 202.5 SRXB-RAI-1
2.5.1 Statement of SRXB-RAI-1
Attachment 4 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 bethe case at a region above the core/ The solution-based guidance for ex-core regions recommends, moregenerally, that "a spatial mesh that ensures the flux in any group changes by less than a factor of -2between adjacent intervals should be applied..."2 The uncertainty analysis presented in BAW-2241A includes a significant contribution of data from theDavis-Besse Cycle 6 ex-vessel dosimetry campaign.
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Page 21b) 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.
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Page 22The 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.
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Page 23Boot-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.
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Page 24Results 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.
[
]
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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.
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Page 26Details 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
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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.
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Page 282.5.2.2 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
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Page 29extended 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.
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Page 30Furthermore, 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/cm2 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.
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Page 31Flux 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 0Q to 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.
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Page 32An 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.
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Page 33o 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
vessel locations not listed above have inside surface fluences below 1E+17 n/cm 2."'
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:
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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.
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Page 35Summary: 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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3.0 REFERENCES
1. Entergy Letter 1 CAN 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)
2. ANP-3300P, Revision 1 (77-3300P-001), "Arkansas Nuclear One (ANO)
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
submittal of proprietary version). 3
3 Revision 0 of BAW-2241P-A is for Babcock & Wilcox (B&W) reactor designs and includes the NRCSafety Evaluation Report (SER) applicable to ANO-1. Revisions 1 and 2 of BAW-2241P-A also containthe associated SERs and increase applicability to include Boiling Water Reactors (BWRs) andWestinghouse or Combustion Engineering (CE) reactors, respectively.
Controlled Document
AREVA Inc. ANP-3300Q1 NPRevision 0
Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature andLow 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-3300Q1NPRevision 0
Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature andLow 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-3300Q1NPRevision 0
Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature andLow Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1
Page 39
Attachment 3 to
1 CAN021502
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. McFadenNOTARY PUBLIC, CQMMONWEALTH OF VIRGINIAMY COMMISSION EXPIRES: 10/31/18Reg. # 7079129
I-.