ML20140D084
| ML20140D084 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna, 05000000 |
| Issue date: | 12/31/1985 |
| From: | Hirst C, Lau F, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20140D070 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR WCAP-11016, WCAP-11016-R01, WCAP-11016-R1, NUDOCS 8601290120 | |
| Download: ML20140D084 (78) | |
Text
WCAP-11016 Rev. 1 WESTINGHOUSE PROPRIETARY CLASS 3
~
CUSTOMER DESIGNATED DISTRIBUTION NORTH ANNA UNITS 1 AND 2 REACTOR VESSEL FLUENCE AND RT UATIONS PTS E. L. Furchi V. A. Perone M. Weaver G. N. Wrights Work Performed for Virginia Power Company December 1985 b
N' M
APPROVED:
APPROVED: T. A. Meyer, Mdnagsr F. L. Tau, Manager Radiation and Systems Structural Materials and Reliability Technology Analysis APPROVED:
[ ft/ Ms C. W. Hirst,' Manager Reactor Coolant System Components Licensing Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.
WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS P. O. 80X 355 PITTS8URGH, PENNSYLVANIA 15230 G601290120 860123 ADOCK 0500 0
3845e:1d/121285 fDR
TABLE OF CONTENTS PAGE i
TABLE OF CONTENTS ii LIST OF TABLES v
LIST OF FIGURES 1
I.
INTRODUCTION I.1 The Pressurized Thermal Shock Rule 1
3 I.2 The Calculation of RTPTS 5
II.
NEUTRON EXPOSURE EVALUATION II.1 Method of Analysis 5
11.2 Fast Neutron Fluence Results 8
35 III.
MATERIAL PROPERTIES III.1 Identification and Location of Beltline Region Materials 35 III.2 Definition and Source of Material Properties for All 35 Vessel locations 111.3 Summary of Plant-Specific Material Properties 36 VALUES FOR ALL BELTLINE 41 IV.
DETERMINATION OF RTPTS REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RTPTS 4I versus Fluence Results IV.2 Discussion of Results 42 V.
CONCLUSIONS AND RECOMMENDATIONS 47 49 VI.
REFERENCES VII.
APPENDICES A.
Power Distribution A-1 B-1 B.
Weld Chemistry Values of North Anna Units 1 and 2 Reactor Vessel C-1 C.
RTPTS Beltline Region Materials 3845e:ld/121285 i
LIST OF TABLES Page II. 2-1 North Anna Unit 1 Fast Neutron (E>l.0 MeV) Exposure at 12 the Pressure Vessel Inner Radius - 0* Azimuthal Angle II.2-2 North Anna Unit 1 Fast Neutron (E>l.0 Mev) Exposure at 13 the Pressure Vessel Inner Radius - 15' Azimuthal Angle 11.2-3 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 14 the Pressure Vessel Inner Radius - 30* Azimuthal Angle 11.2-4 North Ar.na Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 15 the Pressure Vessel Inner Radius - 45* Azimuthal Angle 11.2-5 North Anna Unit i Fast Neutron (E>1.0 MeV) Exposure at 16 the 15* Surveillance Capsule Center 11.2-6 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 17 the 25* Surveillance Capsule Center 11.2-7 North Anna Unit 1 Fast Neutron (E>l.0 MeV) Exposure at 18 the 35* Surveillance Capsule Center II.2-8 North Anna Unit 1 Fast Neutron (E>l.0 MeV) Exposure at 19 the 45* Surveillance Capsule Center 11.2-9 North Anna Unit 2 Fast Neutron (E>1.0 Mev) Exposure at 20 the Pressure Vessel Inner Radius 0* Azimuthal Angle I I. 2-10 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 21 the Pressure Vessel Inner Radius 15* Azimuthal Angle 11.2-11 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 22 the Pressure Vessel Inner Radius 30* Azimuthal Angle I I. 2-12 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 23 the Pressure Vessel Inner Radius 45* Azimuthal Angle 11.2-13 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 24 the 15' Surveillance Capsule Center II. 2-14 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 25 the 25* Surveillance Capsule Center 11.2-15 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 26 the 35* Surveillance Capsule Center 1
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LIST OF TABLES (Continued)
Page 11.2-16 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 27 the 45' Surveillance Capsule Center II I. 3 -1 North Anna Unit 1 Reactor Vessel Beltline Region 37 Material Properties 111.3-2 North Anna Unit 2 Reactor Vessel Beltline Region 38 Material Properties I V.1 -1 RTPTS Values for North Anna Unit 1 43 IV.1-2 RTPTS Values for North Anna Unit 2 44 A-1 North Anna Unit 1 Beginning-of-Cycle and End-of-Cycle A-3 Fuel Assembly Burnups A-2 North Anna Unit 2 Beginning-of-Cycle and End-of-Cycle A-4 Fuel Assembly Burnups A-3 North Anna Unit 1 Core Power Distributions Used in the A-5 Fluence Analysis A-4 North Anna Unit 2 Core Power Distributions Used in the A-6 Fluence Analysis B.1 -1 North Anna Unit 1 Intermediate to Lower Shell Circumferential B-2 Weld Chemistry From WOG Materials Data Base -
Wire Heat Number 25531 B.1 -2 North Anna Ur.it 2 Intermediate to Lower Shell Circumferential B-3 Weld Chemistry From WOG Materials Data Base - Wire Heat Number 716126 C.1 -1 RTpys Values for North Anna Unit 1 Reactor Vessel Beltline C-2 Region Materials @ Fluence = 1.0 x 10 8 n/cm2 1
C.1-2 RTPTS Values for North Anna Unit 1 Reactor Vessel Beltline C-3 Region Materials @ Fluence = 5.0 x 1018 n/cm2 C.1-3 RTPTS Values for North Anna Unit 1 Reactor Vessel Beltline C-4 Region Materials @ Fluence = 1.0 x 1019 n/cm2 C.1-4 RTPTS Values for North Anna Unit 1 Reactor Vessel Beltline C-5 Region Materials @ Current (4.7 EFPY) Fluence Values C.1-5 RTPTS Values for North Anna Unit 1 Reactor Vessel Beltline C-6 Region Materials @ License Expiration (25.0 EFPY) 3845e:ld/121285 111
LIST OF TABLES (Continued)
Page C. 2-1 RTPTS Values for North Anna Unit 2 Reactor Vessel Beltline C-7 Region Materials @ Fluence = 1.0 x 1018 n/cm2 C.2-2 RTPTS Values for North Anna Unit 2 Reactor Vessel Beltline C-8 Region Materials @ Fluence = 5.0 x 10 8 n/cm2 1
C.2-3 RTpys Values for North Anna Unit 2 Reactor Vessel Beltline C-9 Region Materials @ Fluence = 1.0 x 1019 n/cm2 C.2-4 RTPTS Values for North Anna Unit 2 Reactor Vessel Beltline C-10 Region Materials @ Current (3.5 EFPY) Fluence Values C.2-5 RTPTS Values for North Anna Unit 2 Reactor Vessel Beltline C-ll Region Materials @ License Expiration (23.8 EFPY) i 3845e:ld/121285 iv
LIST OF FIGURES PAE 11.1-1 North Anna Reactor Geometry 28 11.2-1 North Anna Unit 1 Maximum Fast Neutron (E>l.0 MeV) 29 Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time 4
11.2-2 North Anna Unit 2 Maximum Fast Neutron (E>l.0 MeV) 30 Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time II.2-3 North Anna Unit 1 Maximum Fast Neutron (E>1.0 MeV) 31 Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle II.2-4 North Anna Unit 2 Maximum Fast Neutron (E>l.0 MeV) 32 Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle r
11.2-5 North Anna Units 1 and 2 Relative Radial 33 Distribution of Fast Neutron (E>l.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 11.2-6 North Anna Units 1 and 2 Relative Axial 34 Distribution of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 111.1-1 Identification and Location of Beltline Region 39 Material for the North Anna Unit i Reactor Vessel 111.1-2 Identification and Location of Beltline Region 40 Material for the North Anna Unit 2 Reactor Vessel IV.1 -1 North Anna Unit 1 - RTPTS Curves per PTS Rule Methodology 45 IV.1-2 North Anna Unit 2 - RTPTS Curves per PTS Rule Methodology 46 A-1 North Anna Units 1 and 2 Core Description for Power A-7 Distribution Map 1
j 384Se:1d/121285 v
= -.
SECTIOE 1 INTRODUCTION The purpose of this report is to submit the reference temperature for pressurized thermal shock (RT aluu for 3 e M Anna Un us 1 and 2 PTS reactor vessels to address the Pressurized Thermal Shock (PTS) Rule.Section I discusses the Rule and provides the methodology for calculating RT PTS' Section 11 presents the results of the neutron exposure evaluation assessing the effects that past and present core management strategies have had on neutron fluence levels in the reactor vessel.
Section III provides the reactor vessels beltline region material properties for both units.Section IV provides the RT calculations from present through the projected PTS end-of-license fluence values.
1.1 THE PRESSURIZED THERMAL SHOCK RULE The Pressurized Thermal Shock (PTS) Rule [1] was approved by the U.S. Nuclear Regulatory Commissioners on June 20, 1985, and appeared in the Federal Register on July 23, 1985. The date that the Rule was published in the Federal Register is the date that the Rule became a regulatory requirement.
The Rule outlines regulations to address the potential for PTS events on pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC).
PTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.
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The Rule establishes the following requirements for all domestic, operating PWRs:
l l
Establishes the RTPTS (measure of fracture resistance) Screening Criterion for the reactor vessel beltline region 270*F for plates, forgings, axial welds 300*F for circumferential weld materials 6 Months From Date of Rule: All plants must submit their present RTPTS values (per the prescribed methodology) and projected RTPTS values at the expiration date of the operating license. The date that this submittal must be received by the NRC for plants with operating licenses is January 23, 1986.
9 Months From Date of Rule: Plants projected to exceed the PTS Screening Criterion shall submit an analysis and a schedule for implementation of such flux reduction programs as are reasonably practicable to avoid reaching the Screening Criterion. The data for this submittal must be received by the NRC for plants with operating i
i li:enses by April 23, 1986.
Requires plant-specific PTS Safety Analyses before a plant is within 3 years of reaching the Screening Criterion, including analyses of alternatives to minimize the PTS concern.
Requires NRC approval for operation beyond the Screening Criterion.
l For applicants of operating licenses, values of the projected RT are to PTS be provided in the Final Safety Analysis Report. This requirement is added as part of 10CFR Part 50.34.
In the Rule, the NRC provides guidance regarding the calculation of the j
toughness state of the reactor vessel materials - the " reference temperature for nil ductility transition" (RTNOT). For purposes of the Rule, RT NOT now defined as "the reference temperature for pressurized thermal shock" (RlPTS) and calculated as prescribed by 10 CFR 50.61(b) of the Rule. Each USNRC licensed PWR must submit a projection of RT values from the time of PTS the submittal to the license expiration date. This assessment must be submitted within 6 months af ter the effective date of the Rule, on January 23, 1986, with updates whenever changes occur af fecting projected values.
The l
l 3845e:1d/121285 2
l 1
---.---,-.,.-,-..-,-n.
- -. - - - - - ~ - - ~. - -
-. ~. - - -
.. - - -.. - - - - - = _ - -.
i calculation must be made for each weld and plate, or forging, in the reactor vessel beltline. The purpose of this report is to provide the RT values PTS for North Anna Units 1 and 2.
,I 1
1.2 THE CALCULATION OF RT PTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RT at a given time.
PTS The prescribed equations in the PTS rule for calculating RT are actually PTS one of several ways to calculate RT For the purpose of comparison with NDT.
the Screening Criterion, the value of RT or W reactor vessel must be PTS calculated for each weld and plate, or forging in the beltline region as given below.
For each material, RT is the lower of the results given by PTS Equations 1 and 2.
I Equation 1:
PTS = I + M + [-10 + 470(Cu) + 350(Cu)(Ni)] f.270 0
RT i
f Equation 2:
RTPTS = I + M + 283 f where j
l l
I = the initial reference transition temperature of the unirradiated material measured as defined in the ASME Code, NB-2331.
If a measured value is not available, the following generic mean values must be used: O'F for welds made with Linde 80 flux, and -56*F for welds made with Linde 0091, 1092 and 124 and i
ARCOS B-5 weld fluxes.
j M = the margin to be added to cover uncertainties in the values of initial RTNDT, c pper and nickel content, fluence, and calculation procedures.
In i
i 4
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5
Equation 1, M-48'F if a measured value of I was used, and M=59'F if the 1
generic mean value of I was used.
In Equation 2 M-0*F if a measured value of I was used, and M=34*F if the generic mean value of I was used.
Cu and. Ni = the best estimate weight percent of copper and nickel in the I
- material, i
f = the maximum neutron fluence, in units of 10I9n/cm2 (E greater than or equal to 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question.
Note, since the chemistry values given in equations 1 and 2 are best estimate mean values, and the margin, M, causes the RT values to be upper bound PTS predictions, the mean material chemistry values are to be used, when available, so as not to compound conservatism.
The basis for the Cu and Ni values used in the RT calculations for North Anna Units 1 and 2 are PTS discussed in Section III.2.
l 3845e:1d/121285 4
i
SECTION II NEUTRON EXPOSURE EVALUATION This section presents the results of the application of Westinghouse derived adjoint importance functions to the calculation of the North Anna Units 1 and 2 reactor vessel fluence for Virginia Power Company.
The use of adjoint importance functions provides a cost effective tool to assess the effects that past and present core management strategies have had on neutron fluence levels in the reactor vessel.
11.1 METHOD OF ANALYSIS A plan view of the North Anna Units 1 and 2 reactor geometry at the core midplane is shown in Figure 11.1-1.
Since the reactor exhibits 1/8th core symmetry only a 0*-45* sector is depicted.
Eight irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program.
The capsules are located at 45*, 55*,
65*,165*, 245*, 285*, 295* and 305* relative to the reactor geometry flat at 0*.
In performing the fast neutron exposure evaluations for the reactor geometry shown in Figure 11.1-1, two sets of transport calculations were carried out.
The first, a single computation in the conventional forward mode, was utilized to provide baseline data derived from a design basis core power distribution against which cycle by cycle plant specific calculations can be compared. The second set of calculations consisted of a series of adjoint analyses relating l
the response of interest (neutron flux > 1.0 MeV) at several selected locations within the reactor geometry to the power distributions in the l
reactor core.
These adjoint importance functions when combined with cycle specific core power distributions yield the plant specific exposure data for each operating fuel cycle.
I The forward transport calculation was carried out in R,e geometry using the DOT discrete ordinates code [2] and the SAILOR cross-section library [3]. The l
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SAILOR library is a 47 group, ENDF/B-IV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with aP expansion of W cmss-secdons. An S angular qua kature was usd.
3 6
The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop plants.
Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; i.e., f resh fuel on the core periphery.
Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal
+2a level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results. This is especially true in cases where low leakage fuel management has been employed.
The design basis core power distribution data used in the analysis is provided in Appendix A of this report. The data listed in Appendix A represents cycle averaged relative assembly powers.
The adjoint analyses were also carried out using the P c mss-secd on 3
approximation from the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions along the inner radius of the pressure vessel. Again, these calculations were run in R,e geometry to provide power distribution importance functions for the exposure parameter of interest (neutron flux > 1.0 MeV). Having the adjoint importance functions and appropriate core power distributions, the I
response of interest is calculated as:
I R,0 " I I
I I (R,0,E) F (R,0,E) dE R dR de R
R s E where:
R Response of interest (+ (E > 1.0 MeV)) at radius R and R,0 azimuthal angle e.
f I
l 3845e:ld/121285 6
l l
Adjoint importance function at radius R and azimuthal I (R,0,E) angle e for neutron energy group E.
Full power fission 'ensity at radius R and azimuthal angle F (R,0,E)
=
e for neutron energy group E.
The fission density distributions used reflect the burnup-dependent inventory of fissioning actinides, including U-235, U-238, Pu-239, and Pu-241.
Core power distributions for use in the plant specific fluence evaluations for North Anna Units 1 and 2 were derived from measured asser.ibly and cycle burnups for each operating cycle to date of the two reactors.
The specific power distribution data used in the analysis is provided in Appendix A of this report.
The data listed in Appendix A represents cycle averaged relative assembly powers. Therefore, the adjoint results were in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.
The projection of reactor vessel fast neutron fluence into the future to the expiration date of the operating license requires that a few key assumptions be made. Current neutron fluences, based on past core loadings, are defined as of September 30, 1985. The operating license for North Anna Unit 1 expires on February 18, 2011 and the operating license for North Anna Unit 2 expires on February 19, 2011 (each forty years after the construction permit was issued). This report includes fluence projections f rom September 30, 1985 to l
the respective license expiration dates using the cycle-averaged core power distribution of the current operating cycle (cycle 5 for Unit 1 and cycle 4 for Unit 2) and an assumed future capacity factor of 80%.
All fluence projections into the future reflect the low leakage fuel management strategies exemplified by the core loadings of the current cycles.
Finally, it has been assumed that the core thermal power will be uprated f rom 2775 MW to 2893 th MW during cycle 6 of Unit I and during cycle 5 of Unit 2.
th The transport methodology, both forward and adjoint, using the SAILOR l
l cross-section library has been benchmarked against the Oakridge National l
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l Laboratory (ORNL) Poolside Critical Assembly (PCA) facility as well as against i
i i
the Westinghouse power reactor surveillance capsule data. base [4).
The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be i
conservative by 7-22%. When plant specific power distributions are used with 1
the adjoint importance functions, the benchmarking studies show that fluence l
predictions are within i 15% of measured values at surveillance capsule l
locations.
11.2 FAST NEUTRON FLUENCE RESULTS Calculated fast neutron (E >1.0 MeV) exposure results for North Anna Units 1 l
and 2 are presented in Tables II.2-1 through 11.2-16 and in Figures 11.2-1 through 11.2-6.
Data is presented at several azimuthal locations on the inner l
radius of the pressure vessel as well as at the center of each surveillance capsule.
In Tables 11.2-1 through 11.2-4 cycle-specific maximum neutron flux and fluence levels at 0*,
15*, 30*, and 45' on the pressure vessel inner radius of North Anna Unit 1 are listed for the period of operation up to September 30, 1985, and projected to the expiration date of the operating license. Also presented are the design basis fluence levels predicted using the generic 3-loop core power distribution at the nominal + 2a level. Similar data for the center of surveillance capsules located at 15*, 25', 35*, and 45' are given in Tables 11.2-5 and 11.2-8, respectively.
l In addition to the calculated data given for the surveillance capsule locations, measured fluence data from previously withdrawn surveillance capsules are also presented for comparison with analytical results.
In the case of Unit 1, a capsule was removed from the 15' location at the end of l
cycle 1.
An error in the measured fluence has however precluded its use l
herein.
l l
l Cycle-specific and design basis fast neutron flux and fluence data at the inner radius of the pressure vessel of North Anna Unit 2 are given in Tables 1
l 3845e:ld/121285 8
l l
11.2-9 through 11.2-12 for the period of operation up to September 30, 1985, l
and projected to the expiration date of the operating license. As in the case of Unit 1, data are presented for the 0*,15*, 30*, and 45' azimuthal angles.
Evaluations of plant specific and design basis fluence levels at the four l
surveillance capsule locations are given in Tables 11.2-13 and 11.2-16.
For Unit 2, a surveillance capsule was removed from the 15' position following j
cycle 1.
A dosimetry evaluation from this capsule withdrawal is also listed in Table 11.2-13.
Several observations regarding the data presented in Tables 11.2-1 through II.2-16 are worthy of note. These observations may be summarized as follows:
1.
In North Anna Unit 2, calculated plant specific fast neutron (E > l.0 MeV) j fluence at the surveillance capsule center is in excellent agreement with measured data.
The difference between the plant specific calculation and the measurement is less than 7%.
Differences of this magnitude are well within the uncertainty of the experimental results.
i j
2.
For both North Anna units, the fast neutron (E > 1.0 MeV) flux incident on the pressure vessel during Cycle 1 was, on the average, 19% less than predictions based on the design basis core power distributions. This result is consistent with the statement that the design basis power distributions produce flux levels that tend to be conservative by 7-22%.
3.
The low leakage fuel management employed during cycle 5 of North Anna Unit 1, which is used for projection into the future, has reduced the peak fast neutron flux (0* azimuthal position) on the pressure vessel by a factor of 1.90 relative to the design basis flux.
(In subsequent discussions, factors of fast neutron flux reduction, defined as the ratio of the design basis flux to the cycle-specific flux, will be quoted.) The cycle 5 core loading produced flux reduction factors ranging f rom 1.58 to 1.79 at the other azimuthal locations.
i f
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1 4.
In North Anna Unit 2, the low leakage core loading used for projection into the future (cycle 4) yielded a flux reduction factor of 1.58 at the peak flux location and factors ranging from 1.25 to 1.45 at the remaining azimuths.
5.
Comparing the flux reduction factors resulting from the low leakage core loadings in cycle 5 of North Anna Unit 1 and cycle 4 of North Anna Unit 2, one observes differences that are attributable to the varying burnups of the fuel assemblies in peripheral locations (see burnup data in Appendix A).
Graphical presentations of the plant specific fast neutron fluence at the peak location on the pressure vessel (0* azimuthal position) are shown in Figures 11.2-1 and 11.2-2 as a function of full power operating time for North Anna Units 1 and 2, respectively.
In regard to Figure 11.2-1 and 11.2-2, the solid portions of the fluence curves are based directly on the cycle specific evaluations, as of September 30, 1985, presented in this report.
The dashed portions of these curves, however, involve a projection f rom September 30, 1985 to the respective license expiration dates.
m mentioned in Section 11.1, the fluence projections are based on the cycle-averaged core power distribution of the current operating cycle (cycle 5 for Unit 1 and cycle 4 for Unit 2) and an assumed future capacity factor of 80%.
It should be noted that implementation of a more severe low leakage pattern than that of the current operating cycle would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those j
projections.
In any event the RT assessment must be updated per PTS l
10CFR50.61(b)(1) whenever, among other things, changes in core loadings significantly impact the fluence and RTPTS projections.
In Figures 11.2-3 and 11.2-4, the azimuthal variation of maximum fast neutron I
(E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented i
as a function of azimuthal angle for Units 1 and 2, respectively.
Data are presented for both current and projected expiration-of-operating-license j
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conditions.
In Figure II.2-5, the relative radial variation of fast neutron I
flux and fluence within the pressure vessel wall is presented. Similar data showing the relative axial variation of fast neutron flux and fluence over the beltline region of the pressure vessel is shown in Figure 11.2-6.
A three-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figure 11.2-3 through 11.2-6 along with the relation
+(R, e,Z) = +(e) F(R) G(Z)
Fast neutron fluence at location R, e, Z within where: 4 (R,0,Z)
=
the pressure vessel wall Fast neutron fluence at azimuthal location e on
+ (e)
=
the pressure vessel inner radius from Figure II.2-3 or 11.2-4 Relative fast neutron flux at depth R into the i
F (R)
=
pressure vessel from Figure 11.2-5 Relative fast neutron flux at axial position Z from G (Z)
=
Figure 11.2-6 i
f Analysis has shown that the radial and axial variations within the vessel wall I
are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable evaluation of fluence gradients within the vessel wall.
3845e:ld/121285 11
l TA8LE II.2-1 l
l NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE (a) 8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg. Flux Plant Desiggb)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.1 5.30 x 10 1.89 x 10 2.42 x 10 10 18 18 CY-2 1.9 6.52 x 10 3.46 x 10 4.05 x 10 10 18 18 CV-3 2.8 4.56 x 10 4.82 x 10 6.07 x 10 10 18 18 CY-4 3.8 5.10 x 10 6.37 x 10 8.13 x 10 CY-5 (9/30/85)(C) 4.7 3.56 x 10 7.35 x 10 9.99 x 10 10 18 18 9/30/85 4 CY-6(d) 5.3 3.56 x 10 8.03 x 10 1.13 x 10 10 18 I9 I
10 I9 I9 CY-6 4 2/18/2011 ')
25.0 3.71 x 10 3.11 x 10 5.52 x 10 (a)
Applicable to the peak azimuthal locations (0', 90', 180*, 270') on the core beltline.
(b)
Design basis fast neutron flux = 6.78 x 1010 n/cm2-sec at 2775 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined.
(d) During CY-6, the core thermal power will be uprated to 2893 MWth.
Beyond 9/30/85 a 80% capacity factor is assumed.
l (e) Exposure period from the onset of the uprating to the license expiration date.
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TA8LE 11.2-2 1
NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15* AZIMUTHAL ANGLE Beltline Region l
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desig[a)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 lI 18 CY-1 1.1 2.54 x 10 9.06 x 10 1.10 x 10 10 18 18 CY-2 1.9 3.10 x 10 1.65 x 10 1.85 x 10 10 18 18 CY-3 2.8 2.18 x 10 2.30 x 10 2.77 x 10 10 18 18 CY-4 3.8 2.34 x 10 3.01 x 10 3.71 x 10 10 18 18 CY-5 (9/30/85)(b) 4.7 1.77 x 10 3.50 x 10 4.56 x 10 10 18 18 9/30/85 + CY-6(*
5.3 1,77 x 10 3.84 x 10 5.15 x 10 10 I9 I9 CY-6 + 2/18/2011(d) 25.0 1.84 x 10 1.53 x 10 2.52 x 10 i
(a) Design basis fast neutron flux = 3.09 x 1010 n/cm2-sec at 2775 MWth l
(b) 9/30/85 is the date at wb.ich the current neutron fluences are defined.
(c) During CY-6, the core thermal power will be uprated to 2893 MWth.
Beyond 9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration date.
l l
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l TABLE 11.2-3 i
NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE t
8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg}
Interval Time (EFPY)
(n/cm -sec)
Specific Basis l
10 lI I
CY-1 1.1 1.39 x 10 4.97 x 10 6.45 x 10 l
CY-2 1.9 1.66 x 10 8.98 x 10 1.08 x 10 10 I
18 10 18 18 CY-3 2.8 1.19 x 10 1.25 x 10 1.62 x 10 10 18 18 CY-4 3.8 1.10 x 10 1.59 x 10 2.17 x 10 l
CY-5 (9/30/85)(b) 4.7 1.01 x 10 1.87 x 10 2.67 x 10 10 18 18 9/30/85 + CY-6(c) 5.3 1.01 x 10 2.06 x 10 3.01 x 10 10 18 18 I
10 18 I
l CY-6 4 2/18/2011 25.0 1.05 x 10 8.60 x 10 1.47 x 10 '
j (a) Design basis fast neutron flux = 1.81 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-6, the core thermal power will be uprated to 2893 MWth.
Beyond j
9/30/85 a 80% capacity f acter is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration date.
l l
l 3845e:1d/121285 14 l
t
TABLE 11.2-4 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGM 8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 9
II ll CY-1 1.1 9.48 x 10 3.38 x 10 3.92 x 10 10 I
II CY-2 1.9 1.08 x 10 5.98 x 10 6.57 x 10 9
I7 II CY-3 2.8 7.75 x 10 8.30 x 10 9.86 x 10 9
18 18 CY-4 3.8 6.96 x 10 1.04 x 10 1.32 x 10 9
18 18 CY-5 (9/30/85)I )
4.7 6.98 x 10 1.23 x 10 1.62 x 10 9
18 I0 9/30/85 4 CY-6(c) 5.3 6.98 x 10 1.37 x 10 1.83 x 10 9
18 18 CY-6 4 2/18/2011(d) 25.0 7.28 x 10 5.89 x 10 8.92 x 10 I
(a) Design basis fast neutron flux = 1.10 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-6, the core thermal power will be uprated to 2893 MWth.
8eyond 9/30/85 a 80% capacity factor is assumed.
(d) Exposure period from the onset of the uprating to the license expiration date.
3845e:1d/121285 15
TABLE II.2-5 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15* SURVEILLANCE
~
CAPSULE CENTER Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg. Flux Plant Desigga)
Capsule Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data 10 18 18 CY-1 1.1 8.79 X 10 3.13 x 10 3.82 x 10 (g)
Il 18 18 CY-2 1.9 1.07 x 10 5.72 x 10 6.39 x 10 10 18 18 CY-3 2.8 7.47 x 10 7.95 x 10 9.59 x 10 10 I9 I9 CY-4 3.8 8.08 x 10 1.04 x 10 1.28 x 10 CY-5 (9/30/85)(b) 10 I9 I
4.7 6.04 x 10 1.21 x 10 1.58 x 10 I) 10 I9 I9 9/30/85 + CY-6 5.3 6.04 x 10 1.32 x 10 1.78 x 10 CY-6 + 2/18/2011(d) 25.0 6.30 x 10 5.24 x 10 8.75 x 10 10 I9 I9 (a) Design basis fast neutron flux = 1.07 x 10ll 2
n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-6, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(d) Exposure period from the onset of the uprating to the license expiration date.
(e) An error in the reported data of Reference 5 has precluded its use in the comparison of measured and calculated surveillance capsule fluence.
l l
38453: 1d/121285 16
TA8LE II.2-6 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25' SURVEILLANCE CAPSULE CENTER Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg. Flux Plant Desigga)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.1 5.60 x 10 2.00 x 10 2.43 x 10 10 18 18 CY-2 1.9 6.77 x 10 3.62 x 10 4.06 x 10 10 18 18 CY-3 2.8 4.90 x 10 5.09 x 10 6.10 x 10 10 18 18 CY-4 3.8 4.64 x 10 6.50 x 10 8.15 x 10 10 18 I9 4.7 4.04 x 10 7.61 x 10 1.00 x 10 CY-5 (9/30/85)IDI 10 18 I9 9/30/85 + CY-6(C}
5.3 4.04 x 10 8.38 x 10 1.13 x 10 10 I9 l9 Id) 25.0 4.21 x 10 3.45 x 10 5.54 x 10 CY-6 + 2/18/2011 10 n/cm -sec at 2775 MWth 2
(a) Design basis fast neutron flux = 6.80 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-6, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration date.
l l
3845e:1d/121285 17
TABLE 11.2-7 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35* SURVEILLANCE CAPSULE CENTER Beltline Region 2
Elapsed Cumulative Fluence (n/cn )
Irradiation Irradiation Avg.2 Flux Plant Desig[a)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.1 3.86 x 10 1.38 x 10 1.66 x 10 10 18 18 CY-2 1.9 4.54 x 10 2.47 x 10 2.78 x 10 10 18 18 CY-3 2.8 3.22 x 10 3.43 x 10 4.17 x 10 10 18 18 CY-4 3.8 2.90 x 10 4.31 x 10 5.58 x 10 CY-5 (9/30/85)I )
10 18 18 4.7 2.80 x 10 5.08 x 10 6.86 x 10 IC 10 18 18 9/30/85 + CY-6 5.3 2.80 x 10 5.61 x 10 7.75 x 10 CY-6 + 2/18/2011(d) 10 I9 I9 25.0 2.92 x 10 2.38 x 10 3.79 x 10 (a)
Design basis fast neutron flux = 4.65 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c)
During CY-6, the core thermal power will be uprated to 2893 MWth.
Beyond 9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration date.
3845e:1d/121285 18
TA8LE 11.2-8 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45' SURVEILLANCE CAPSULE CENTER Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigga)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 IO CY-1 1.1 3.08 x 10 1.10 x 10 1.31 x 10 10 18 18 CY-2 1.9 3.51 x 10 1.94 x 10 2.19 x 10 10 18 18 CY-3 2.8 2.49 x 10 2.69 x 10 3.29 x 10 10 18 18 CY-4 3.8 2.24 x 10 3.37 x 10 4.40 x 10 10 18 18 CY-5 (9/30/85)(b) 4.7 2.24 x 10 3.98 x 10 5.41 x 10 10 18 18 9/30/85 4 CY-6(c) 5.3 2.24 x 10 4.41 x 10 6.11 x 10 10 I9 I9 CY-6 + 2/18/2011(d) 25.0 2.34 x 10 1.90 x 10 2.99 x 10 (a) Design basis fast neutron flux = 3.67 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-6, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(d) Exposure period from the onset of the uprating to the license expiration date.
3845e:1d/121285 19
TABLE II.2-9 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE (a)
Beltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.0 5.30 X 10 1.73 x 10 2.22 x 10 10 18 18 CY-2 1.6 6.54 x 10 2.97 x 10 3.50 x 10 10 18 18 CY-3 2.7 3.87 x 10 4.26 x 10 5.75 x 10 CY-4 (9/30/85)IC 10 18 18 3.5 4.30 x 10 5.31 x 10 7.41 x 10 9/30/85 4 CY-5(d) 10 18 18 4.4 4.30 x 10 6.56 x 10 9.38 x 10 I
CY-5 4 2/19/2011 ')
10 I
I 23.8 4.48 x 10 3.40 x 10 '
5.26 x 10 '
(a) Applicable to the peak azimuthal locations (0', 90', 180*, 270') on the core beltline.
(b)
Design basis fast neutron flux = 6.78 x 1010 n/cm2-sec at 2775 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined.
(d) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(e)
Exposure period from the onset of the uprating to the license expiration date.
3845e:1d/121285 20
TABLE II.2-10 l
i NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15' AZIMUTHAL ANGLE Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigga)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 I
18 CY-1 1.0 2.54 X 10 8.29 x 10 1.01 x 10 10 18 18 CY-2 1.6 3.11 x 10 1.42 x 10 1.60 x 10 10 18 I8 CY-3 2.7 2.12 x 10 2.12 x 10 2.62 x 10 10 18 18 CY-4 (9/30/85)I )
3.5 2.15 x 10 2.65 x 10 3.38 x 10 I
10 18 I0 9/30/85 + CY-S '}
4.4 2.15 x 10 3.27 x 10 4.28 x 10 10 I9 I9 CY-5 -+ 2/19/2011(d) 23.8 2.24 x 10 1.70 x 10 2.40 x 10 4
(a) Design basis fast neutron flux = 3.09 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
1 (d) Exposure period from the onset of the uprating to the license expiration date.
I 1
3845e:1d/121285 21
TABLE 11.2-11 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE 8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg. Flux Plant Desigga)
Interval Time (E7PY)
(n/cm -sec)
Specific Basis 10 I7 II CY-1 1.0 1.39 X 10 4.54 x 10 5.92 x 10 10 I
I CY-2 1.6 1.68 x 10 7.72 x 10 9.35 x 10 10 18 18 CY-3 2.7 1.41 x 10 1.24 x 10 1.54 x 10 CY-4 (9/30/85)I )
10 18 18 3.5 1.25 x 10 1.54 x 10 1.98 x 10 9/30/85 4 CY-5(c) 10 18 18 4.4 1.25 x 10 1.91 x 10 2.50 x 10 CY-5 4 2/19/2011(d) 10 18 I9 23.8 1.30 x 10 9.87 x 10 1.40 x 10 (a) Design basis fast neutron flux = 1.81 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c)
During CY-5, the core t' ermal power will be uprated to 2893 MWth.
Beyond 9/30/85 a 80% capacity tactor is assumed.
(d)
Exposure period f rom the onset of the uprating to the license expiration date.
3845e:1d/121285 22
TABLE 11.2-12 l
NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45* AZIMUTHAL ANGLE Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigga)
Basis i
Interval Time (EFPY)
(n/cm -sec)
Specific 9
lI lI CY-1 1.0 9.37 X 10 3.07 x 10 3.60 x 10 10 I
I CY-2 1.6 1.11 x 10 5.17 x 10 5.68 x 10 9
I I7 CY-3 2.7 9.73 x 10 8.39 x 10 9.33 x 10 18 18 CY-4 (9/30/85)( }
3.5 8.80 x 10 1.06 x 10 1.20 x 10 18 18 9/30/85 -* CY-5(c) 4.4 8.80 x 10 1.31 x 10 1.52 x 10 18 I)
CY-5 4 2/19/2011 23.8 9.17 x 10 6.93 x 10 8.50 x 10 10 n/cm2-sec at 2775 MWth (a)
Design basis fast neutron flux = 1.10 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c)
During CY-5, the core thermal power will be uprated to 2893 MWth.
Beyond 9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration date.
I 3845e:1d/121285 23
TABLE 11.2-13 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15' SURVEILLAhCE CAPSULE CENTER j
8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg,)
Capsule i
Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data 10 18 18 18 CY -1 1.0 8.78 X 10 2.87 x 10 3.50 x 10 2.70 x 10 CY-2 1.6 1.08 x 10" 4.92 x 10 5.53 x 10 18 18 10 18 18 CV-3 2.7 7.22 x 10 7.31 x 10 9.07 x 10 CY-4 (9/30/85)ID) 3.5 7.38 x 10 9.13 x 10 1.17 x 10 10 18 I9 9/30/85 + CY-SIC) 10 I9 I9 4.4 7.38 x 10 1.13 x 10 1.48 x 10 CY-5 4 2/19/2011(d) 10 I9 I9 23.8 7.69 x 10 5.84 x 10 8.34 x 10 (a) Design basis fast neutron flux = 1.07 x 10ll n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
l (c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(d) Exposure period f rom the onset of the uprating to the license expiration date.
(e) Reflects adjustments made to the spectrum-averaged reaction cross-sections and dosimeter location reported in Reference 6.
l 3845e:1d/121285 24 1
a
, ~.... -. -
,.---.m----.
TABLE II.2-14 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25' SURVEILLANCE CAPSULE CENTER l
Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.0 5.58 X 10 1.83 x 10 2.22 x 10 10 18 18 i
CY-2 1.6 6.78 x 10 3.11 x 10 3.51 x 10 10 18 18 CY-3 2.7 5.54 x 10 4.95 x 10 5.77 x 10 10 18 18 CY-4 (9/30/85)(b) 3.5 4.93 x 10 6.16 x 10 7.44 x 10 10 18 18 9/30/85 -> CY-5(c) 4.4 4.93 x 10 7.59 x 10 9.41 x 10 I}
10 I9 I9 CY-5 -+ 2/19/2011 23.8 5.14 x 10 3.90 x 10 5.28 x 10 (a) Design basis f ast neutron flux = 6.80 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(d) Exposure period from the onset of the uprating to the license expiration date.
l l
l l
l l
l 3845e:1d/121285 25 l
1 TA8LE II.2-15 l
l NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35' SURVEILLANCE CAPSULE CENTER Beltline Region 2
l Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.0 3.83 X 10 1.25 x 10 1.52 x 10 10 18 18 CY-2 1.6 4.59 x 10 2.12 x 10 2.40 x 10 i
CY-3 2.7 3.92 x 10 3.42 x 10 3.94 x 10 10 18 18 l
CY-4 (9/30/85)I I 3.5 3.50 x 10 4.28 x 10 5.09 x 10 10 18 18 IC) 10 18 18 l
9/30/85 + CY-S 4.4 3.50 x 10 5.30 x 10 6.43 x 10 CY-5 4 2/19/2011(d) 10 I9 I9 23.8 3.65 x 10 2.77 x 10 3.61 x 10 (a)
Design basis fast neutron flux = 4.65 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c)
During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond l
9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration date.
l l
\\
l 3845e:1d/121285 26
TA8LE II.2-16 i
NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45* SURVEll. LANCE CAPSULE CENTER 8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 lI 18 CY-1 1.0 3.04 X 10 9.96 x 10 1.20 x 10 10 18 18 CY-2 1.6 3.60 x 10 1.68 x 10 1.90 x 10 10 18 18 CY-3 2.7 3.15 x 10 2.73 x 10 3.11 x 10 10 18 18 CY-4 (9/30/85)( )
3.5 2.84 x 10 3.43 x 10 4.02 x 10 10 18 18 9/30/85 4 CY-5(c) 4.4 2.84 x 10 4.25 x 10 5.08 x 10 10 I9 I9 CY-5 4 2/19/2011(d) 23.8 2.97 x 10 2.24 x 10 2.85 x 10 (a) Design basis f ast neutron flux = 3.67 x 1010 n/cm2-sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
i (c)
During CY-5, the core thermal power will be uprated to 2893 MWth.
Beyond 9/30/85 a 80% capacity factor is assumed.
(d) Exposure period from the onset of the uprating to the license expiration date.
l l
i l
i 3845e:1d/121285 27
16003-1 O'
(MAJOR AXIS) 15*
(CAPSULES V,X)
I r
25' (CAPSULES Y,W,U) j
/
/
35' (CAPSULES Z,T)
/
l
/
g\\
45' (CAPSULE S)
I
/' /
/
ANNNNN j f
Ny PRESSURE VESSEL j
vmuxx
/
\\
/
/
/
\\
/
/
/
N'Nxx
[
f
/
N THERMAL SHIELD I
/
l
/
/
/
/
/
/
/
/
CORE BARREL
~fI t
- I,,/,/
f BAFFLE
/
I f /
/ f REACTOR CORE
/
4
I lis I//
Figure 11.1-1. North Anna Reactor Geometry 28
1 16003-3 I
l i
1020 7
5 N
b 3
O
/
\\
\\
s' C
/
W
/
o
/
z
/
W
/
3
/
J
/
IO I9
/
2
[
ACTUAL 7
PROJECTED a
5 F
U1<
L 3
LICENSE 9/30/85 EXPIRATION l
1 I
10 18 O
10 20 30 40 50 60 70 l
OPERATING TIME (EFPY) i Figure l1.21, North Anna Unit 1 Maximum Fast Neutron (E > 1.0 MeV)
Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time 1
I r
29 i
i
a
.,s a
-.-..a..u-
--a-..a.
a-u
- =.-
l l
l 16003-3 1020 7
5 l
N
- Oo E
/
s" 3
/p' C
p
/
~
W
/
O
/
z
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W
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D
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l 10 19
/
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ACTUAL o
/
tr 7
l r
/
PROJECTED I
\\
~
l z
5
<k 3
i LICENSE 9/30/85 EXPIRATION l
I l
l l
l 10 18 0
10 20 30 40 50 60 70 OPERATING TIME (EFPY)
Figure 11.2 2. North Anna Unit 2 Maximum Fast Neutron (E > 1.0 MeV)
Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time 30 1
16003 4 k
i i
1020 i
I 7
ACTUAL l
5
PROJECTED N
Eo 3 -s i
s.
s C
i
\\
\\
i W
\\
4 O
N 4
Z
\\
l w
\\
D N
J
- s i
L lal9 N
Z s N o
s 1
m 7
s s 1
e s~~-.
4 3
Wz 5
LICENSE EXPIRATION b
l
~
<L 3
1 9/30/85 10 18 I
I I
I l
O 10 20 30 40 50 60 70 l
AZIMUTHAL ANGLE (DEG.)
i l
i Figure ll.2-3. North Anna Unit 1 Maximum Fast Neutron (E > 1.0 MeV) j j
Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle i
l
'l
,,,_.,-._.-m,~-
n.-,------r
..,-m,m,,y_,-,wm,c--r,-enm-----
_cv---
---_r--e
-,n-,mn,
16003-5 l020 7
ACTUAL 5
PROJECTED N
E u
3
% s N
s C
\\ N m
W
~
g O
N z
s W
s 3
s%s J
N L
jol9 s*s 2
s's O
N 7
s' LICENSE EXPIRATION l
~
z 5
l F-tn<
h.
3 l
t l
9/30/85 10 18 I
I I
l I
O 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.)
i l
Figure 11.2 4. North Anna Unit 2 Maximum Fast Neutron (E > 1.0 MeV)
Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 32
. -. ~
16003 6 10 7
l
]
5 3
i i
i 199.39 W
l.O 204.66 d
7
\\
CLAD 5
IR
~
3 1/4T 214.41 i
W j
Q 219.29 I
O.1 i
3/4T 1
7 45' t
1 i
5 O'
l i
REACTOR 3
VESSEL OR 1
i I
i I
I I
I I
I O.01
{
195 199 203 207 211 215 219 223
.i RADIUS (em) i i
Figure 11.2 5. North Anna Units 1 and 2 Relative Radial Distribution of Fast j
Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure i
Vessel Wall i
33 i
6
.,r,...
1CD03 7 l
l l.O 1
7 5
3 w
Oz 3
o.I u.
s x
7 o
d 5
z 3
sa w>
p o.01 d
7 a-5 3
CORE MIDPLANE I
I I
I o.ool
-300
-200
-100 o
100 200 300 DISTANCE FROM CORE MIDPLANE (cm)
Figure 11.2 6. North Anna Units 1 and 2 Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall i
34
\\
~ -
. = _
SECTION 111 MATERIAL PROPERTIES For the RT calculation, the best estimate copper and nickel chemical py3 composition of the reactor vessel beltline material is necessary. The l
material properties for the North Anna Units 1 and 2 beltline region will be presented in this section.
I 111.1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS The beltline region is defined by the Rule [1] to be "the region of the i
reactor vessel (shell material including welds, heat affected zones, and l
plates or forgings) that directly surrounds the ef fective height of the active f
core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the i
selection of the most limiting material with regard to radiation damage."
f Figures 111.1-1 and 111.1-2 identify and indicate the location of all beltline region materials for the North Anna Units 1 and 2 reactor vessels.
4 111.2 DEFINITION AND SOURCE OF MATERIAL PROPERTIES FOR ALL VESSEL LOCATIONS f
h
.t l
Material property values for the forgings, which have been docketed with the i
NRC in Reference 7, were derived ' rom vessel fabrication test certificate results. The property data for the welds have also been docketed with the NRC in Reference 7, however, the weld properties cannot be used in the same direct
]
manner as the properties for the forgings.
l Fast neutron irradiation-induced changes in the tension, fracture, and impact properties of reactor vessel materials are largely dependent on chemical composition, particularly in the copper concentration. The variability in irradiation-induced property changes, which exists in general, is compounded l
by the variability of copper concentration within the weldments.
l l
1 i
t i
I l
3845e:Id/121285 35 i
~
-,.. _ _ _, _ _ _. _. ~, _., _, _ _. _ _. -
l l
For each weld in the North Anna Units 1 and 2 beltline region, a material data search was performed using the WOG Reactor Vessel Beltline Region Weld Metal l
Data Base.
The WOG data base, which was developed in 1984 and is continually being updated, contains information from weld qualification records, surveillance capsule reports, the B&W report BAW-1799[8), and the Materials Properties Council (MPC) and the NRC Mender MATSURV data bases.
Searches on the WOG Data Base were performed for materials having the identical weld wire heat number as those in the North Anna vessels, but any combination of wire and flux was allowed.
For all of the data found for a particular wire, the copper, nickel, phosphorous and silicon values were averaged and the standard deviations were calculated. Although phosphorous and silicon are not needed for the PTS Rule, they are provided for the sake of l
completeness.
The information obtained from the data base searches is found in Appendix B.
111.3
SUMMARY
OF PLANT-SPECIFIC MATERIAL PROPERTIES A summary of the pertinent chemical and mechanical properties of the beltline region shell and weld materials of the North Anna Units 1 and 2 reactor i
vessels are respectively given in Tables 111.3-1 and 111.3-2 along with the references for this information. Alth;ogh phosphorus is no longer used in the calculation of RT with respe:t to the PTS rule [1], it is given for NOT reference since it is currently used in the Regulatory Guide 1.99, Revision 1 trend curve (9).
i The initial RT shown for all North Anna Units 1 and 2 shells and NDT weldments are the actual values.
The data in Tables !!!.3-1 and !!!.3-2 are used to evaluate the RT values i
py3 for the North Anna Unit I and 2 reactor vessels.
1 l
l 3845e:1d/121285 36 i
e
[
TABLE 111.3-1 I
i NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P
I i
(Wt.M LWt,%)
(Wt %)
(*F1 Source j
l, Intermediate Shell 04 Heat 0.12 0.82 0.010 17 Ref. [7]
t 990311/298244 Circumf erential Weld - Intermed. to Lower Shell WO4, Heat 25531, Smit 89, 4
Flux 1211:
i i
i 0.086 0.11 0.02 19 WOG Material i
Data nase j
f l
Lower Shell 03 Heat 0.15 0.80 0.009 38 Ref. [7]
l 990400/292332 i
e i
I l
I l
)
+
i j
i i
)
t i
3845e:1d/121285 37 I
1 TABLE 111.3-2 I
NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION l
MATERIAL PROPERTIES Cu Ni P
I (Wt.5) {Wt.5) (Wt.5) M Source Intermediate Shell 04 Heat 0.09 0.83 0.011 75 Ref. [7]
990496/292424 i
Lower Shell 03 Heat 0.13 0.83 0.013 56 Ref. [7]
990533/297355 Circumferential Weld - Intermed, to Lower Shell WO4, Heat 716126, Grau Lo 26 i
~
0.069 0.051 0.016
-48 WOG Material Data Base l
I i
r i
I l
3845e:1d/121285 38 l
- - ~
~
FIGURE III.1-1 Identification and Location of North Anna Unit No.1 Reactor Vessel Beltline Material H
aggj Weld Seam WOS g,o.
~
- E Forginq 04
- 1o o-m 144.0" g
B
.'i 1
c ORE (17.1
,o.
Weld Seam WO4 ste' 1
cet Foraina 03 e.
- 1se' h
't 3"
49.3"
,o.
C 10
_ _ _ _ _ _ _ _ _. _ _ _ _ _ _ = _
l i
l FIGURE III.1-2 Identification and Location of North Anna Unit No. 2 heactor Wessel Beltline Materf al
/
rio' as!
Weld Seam WOS
.i t
l
}
Forainq 04
_ 3,.
o.
x 144.0" E
T 1
C o8tE [17.2 M'
E!!
Weld Seam WO4 r7o' j
Foraina 03 cet C'
iso
- 5 u
u 49.3" n'
l 40
,._---,e--_-,w.-%_ _ _ _ _,,
-,,-.._%w-
--.,w,_
.,m%_
.,,7
--q-,
.ww,.,
4->
yw9
.y.
y.v, eg9-
SECTION IV DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS l
Using the methodology prescribed in Section 1.2, the results of the f ast f
l neutron exposure provided in Section 11, and the material properties discussed in Section Ill, the RT values for North Anna Units 1 and 2 can now be PTS determined.
i j
IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT PTS l
RESULTS Using the prescribed PTS Rule methodology, RT values were generated for i
PTS i
all beltline region materials of the North Anna Units I and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes.
The i
l tabulated results from the total evaluation are presented in Appendix C for all beltline region materials for both units.
i 1
Figures IV.1-1 and IV.1-2 present the RT values for the limiting shell PTS and circumferential weld of the North Anna Units 1 and 2 vessels in terms of i,
1 j
RT versus fluence
- curves.
The curves in these figures can be used*
l PTS I
[
l 0
to provide guidelines to evaluate fuel reload options in relation to the NRC RT Screening Criterion for PTS (i.e., RT values can be I
PTS p73 i
readily projected for any options under consideration, provided fluence is r
i known),and j
f o
to show the current (4.7 EFPY for North Anna 1 and 3.5 EFPY for North Anna
{
- 2) and end-of-license (25.0 EFPY for North Anna 1 and 23.8 EFPY for North Anna 2) RT values using actual and projecM Muence.
py3 t
f i
i j
- The EFPY can be determined using Figure 11.2-1 for Unit 1 and Figure 11.2-2 for Unit 2.
i I
3845e:Id/121285 41
4 Table IV.1-1 and IV.1-2 provide a summary of the RT values for all PTS beltline region materials for the lifetime of interest.
IV.2 DISCUSSION OF RESULTS As shown in Figures IV.1-1 and IV.1-2, the lower shells are the governing locations for both reactor vessels relative to PTS.
remain below the NRC screening values for PTS using the projected fluence values through license expiration.
The most limiting RT value at license PTS expiration is 225'F for the lower shell of Unit 1 and 227'F for the lower shell of Unit 2.
l I
l t
l I
l l
l l
l I
i l
i 3845e:ld/121285 42
TABLE IV.1-1 Ripys VALUES FOR NOR1H ANNA UNIT 1 Ripts Values (*F)
Present End-of-License Screening Location vessel Material.
14.7 EFPY)
(25.0 EFPY)_
Criteria i
Intermediate shell 04 139 175 270 2
Intermediate to lower shell 100 116 300 circumferential weld WO4 3
Lower shell 03 180 225 270 3045e:1d/121285 43
d J
TABLE IV.1-2 Ripts VALUES FOR NORTH ANNA UNii 7 1
I a e RIPTS Values (*F)
Present End'-of-t.lcense
$F.reening Location Vessel Material 1 L El f R (231 EFPYL Criteria I
l 1
Intermediate shell 04 1Y2 204 270 i
I 2
Lower shell 03 179 227 270 l
l i
3 Intermediate to lower shcll 2b 34 300 circumferential weld WO4 8
/
\\
\\
p i
W i
Y s'
VI 3845e:1d/12120$
44
. ~.
CURVES PER PTS RULE METHOD [I] DOCKETED NORTH ANNA UNYT 1 - RTp FIGURE IV.1-1 BASEMATERIALANDWOGDIhABASEMEANWELDMATERIALPROPERTIES NRC SCREENING VALUE FOR CIRCUMFERENTIAL WELDS 300 280 - NRC SCREENING VALUE FOR FORGINGS 260 -
i 240 -
1 220 -
200 - LIMITING FORGING (LOWER SHELL) m180 5v160 we
- -140
,a-CIRCUMFERENTIAL WELD i
t m
E 120 l
100
)
M 60 40 20
.j
iI l
0 1018 1018 102t FLLENCE. IEUTROItS / Cu2 s Present (4.7 EFPY)
W Licens* Expiration (25.0 EFPY)
FIGURE IV.1-2 NORTH ANNA UNIT 2 - RT PTS CURVES PER PTS RULE METHOD [1]
DOCKETED BASE MATERIAL AND WOG DATA BASE MEAN WELD MATERIAL PROPERTIES NRC SCREENING VALUE FOR CIRCUMFERENTIAL WELDS 300
~---
280 NRC SCREENING VALUE FOR FORGINGS 264 240 220 200 n180 M
LIMITING FORGING (LOWER SHELL) v160 S
U 140 E 120 100 80 80 40 CIRCUM ERENTIAL WELD
\\
20 0
' ' ' ' I 0
1018 1018 102C FLUDICE. MTRONS / Cg2 e Present (3.5 EFPY) m License Expiration (23.8 EFPY)
A hM
l SECTION V CONCLUSIONS AND RECOMMENDATIONS Calculations have been completed in order to submit RT values for the PTS North Anna Units 1 and 2 reactor vessels in meeting the requirements of the NRC Rule for Pressurized Thermal Shock [1]. This work entailed a neutron exposure evaluation and a reactor vessel material study in order to determine the RT values.
PTS Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methodology have been completed for the North Anna Units 1 and 2 pressure vessels.
Explicit talculations were performed for the operating cycles of both units as of September 30, 1985.
For both units, projection of the f ast neutron exposure beyond the current operating cycle was based on continued implementation of low leakage fuel management similar to that employed during cycle 5 for Unit 1 and cycle 4 for Unit 2.
In regard to the low leakage fuel management already in place at the North Anna Units, the cycle-specific evaluations have demonstrated that for the low leakage case (cycle 5 in Unit 1 and cycle 4 in Unit 2) the peak fast neutron flux at the 0* azimuthal position has been reduced by a factor of 1.90 in Unit 1 and a factor of 1.58 in Unit 2 relative to the flux based on the design basis core power distribution.
This location represents the maximum fast neutron flux incident on the reactor pressure vessel. At other locations on the vessel, as well as at the surveillance capsules, the impact of low leakage will differ from the data presented above.
It should be noted that *itnificant deviations from the low leakage scheme already in place wit 1 af fect the exposure projections beyond the current operating cycle. A move toward a more severe form of low leakage (lower relative power on the periphery) would tend to reduce the projection.
On the other hand, a relaxation of the loading pattern toward higher relative power 3845e:1d/121285 41
on the core periphery would increase the projections beyond those reported.
As each future fuel cycle evolves, the loading patterns su uld be analyzed to determine their potential impact on vessel and caps G 9p m are.
The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence levels derived from neutron dosimetry contained in the surveillance capsules withdrawn from each of the North Anna Units.
For Unit 1, an error in the reported measured fluence precluded its For Unit 2, the ratio of calculated to measured fluence is 0.94 for the use.
15* surveillance capsule withdrawn following cycle 1.
This excellent agreement between calculation and measurement supports the use of this analytical approach to perform a plant specific evaluations for the North Anna reactors.
Material property values for the North Anna Units 1 and 2 reactor vessel beltline region components were determined.
The pertinent chemical and mechanical properties for the shells remain the same as those that have been docketed with the NRC in Reference 7.
The weld material properties are determined using the WOG Materials Data Base.
Using the prescribed PTS Rule methodology, RT values were generated for PTS all beltline region materials of the North Anna Units 1 and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes.
For both reactor vessels, all the RT values remain below the NRC screening PTS vf *s for PTS using the projected fluence exposure through license expiration.
The most limiting values at end-of-license (25.0 EFPY for North Anna 1 and 23.8 EFPY for North Anna 2) are 225'F and 227'F for the lower shells of Unit 1 and Unit 2, respectively.
The results in this report are provided to enable Virginia Power Company to comply with the initial 6 months submittal requirements of the USNRC PTS Rule.
l l
3845e:ld/121285 48
SECTION VI REFERENCES 1.
Nuclear Regulatory Commission,10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," Federal Register, Vol. 50, No. 141, July 23, 1985.
2.
Soltesz, R.
G., Disney, R.
K.,
Jedruch, J. and Ziegler, S.
L.,
" Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.
3.
" SAILOR RSIC Data Library Collection DLC-76." Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P,
oss4eWon Way for UgM WaW 3
Reactors.
4.
Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -
to be published.
5.
BAW-1638, " Analysis of Capsule V f rom the Virginia Electric and Power Company North Anna Unit No. 1 Reactor Vessel Materials Surveillance Program," A. L. Lowe, Jr., et al., May 1981.
6.
BAW-1794, " Analysis of Capsule V f rom the Virginia Electric and Power Company North Anna Unit No. 2 Reactor Vessel Materials Surveillance Program," A. L. Lowe, Jr., et al., October 1983.
7.
Letter from C. M. Stallings of Virginia Power to H. R. Denton of the NRC, Serial No. 501B, December 11, 1978.
8.
B&W Owners Group Report, BAW-1799, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study", July 1983.
3845e:ld/121285 49
9.
" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Regulatory Guide 1.99 - Revision 1, U.S. Nuclear Regulatory Commission, Washington, April 1977.
- 10. Letter f rom K. L. Basehore of Virginia Power Company to D. R. Beynon, Jr.
of Westinghouse Electric Corporation transmitting measured fuel assembly an:1 cycle burnups for the Surry and North Anna Units, dated October 7, 1985.
3845e:1d/121285 50
APPENDIX A i
)
POWER DISTRIBUTIONS Core power distributions used in the plant specific f ast neutron exposure analysis of the North Anna pressure vessels were derived from the measured fuel assembly and cycle burnup data supplied by Virginia Power [10]. The beginning-of-cycle (B0C) and end-of-cycle (EOC) fuel assembly burnups, based on incore flux maps, were provided for selected peripheral fuel assembly locations for each of the previous cycles of operation.
In addition, estimated data was provided for the current cycle of operation (cycle 5 of Unit 1 and cycle 4 of Unit 2).
Table A-1 shows the North Anna Unit 1 fuel assembly and cycle burnups for Cycles 1 through 5.
Similar data for North Anna Unit 2 are shown in Table A-2.
(The fuel assembly locations in the North Anna cores are numbered according to Figure A-1.)
Cycle-averaged relative assembly powers for each cycle were computed using the following relation Relative Assembly Power = E0C Assembly Burnup - B0C Assembly Burnup Cycle Burnup and are shown in Tables A-3 and A-4 for North Anna Units 1 and 2, respectively.
The cycle-averaged relative assembly powers representing the design basis core power distribution are also shown in Tables A-3 and A-4.
Due to the extreme self-shielding of the reactor core, neutrons born in fuel assemblies inboard of those for which burnup data were requested do not contribute significantly to the fast neutron exposure either at the surveillance capsules or at the pressure vessel.
Therefore, power distributiop data for these interior assemblies are not given in Tables A-3 and A-4.
3845e:ld/121285 A-1
In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels. For peripheral assembly locations 1, 2, 3, 4, 5 and their symmetric partners, these spatial gradients also include adjustments to account for analytical deficiencies that tend to occur near the boundaries of the core region.
i 1
3845e:ld/121285 A-2
TABLE A-1 NORTH ANNA UNIT 1 BEGINNING-0F-CYCLE AND END-0F-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assembly 8urnup (MWD /MTU)
Fuel Cycle (a) 1(15892) 2(10711) 3(13335) 4(13478) 5(13000)
Assembly B0C E0C B0C E0C B0C E0C B0C E0C 80C E0C 1
0 11500 0
9570 0
9040 0
9920 9922 16202 2
0 9095 0
7845 26466 31468 16299 22625 30715 34608 3
0 13550 0
10895 0
12105 0
12120 16867 25462 4
0 9435 0
7953 22432 28195 27414 32475 31976 36424 5
0 10458 0
8165 24721 30658 28455 33675 26933 32090 6
0 15070 17095 27140 7992 21585 11819 25375 17518 29523 7
0 15428 0
12530 0
14045 0
14180 0
12705 8
0 16810 19120 28850 12484 27413 14047 29058 13050 28565 9
0 16465 18795 28688 0
15898 0
15810 0
14790 10 0
15560 0
12483 0
14448 0
13005 0
13435 11 0
17045 15428 26468 0
16295 0
15853 0
15637 12 0
14200 18558 27125 7979 20795 9004 20950 12139 23738 (a) The number in parentheses beside the cycle number is the fuel cycle length in MWD /MTU.
3845e:ld/121285 A-3
TABLE A-2 NORTH ANNA UNIT 2 BEGINNING-OF-CYCLE AND END-OF-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assemb1v Burnup (MWD /MTU)
Fuel Cycle (a) 1(14494) 2(8436) 3(14717) 4(16000)
Assembiv 80C E0C 80C EOC 80C EOC 80C E0C 1
0 10500 0
7510 6060 13685 13424 21998 2
0 8288 0
6145 23520 28638 17561 24292 3
0 12325 0
8715 0
13700 18422 29624 4
0 8530 0
6253 9377 17560 18900 26484 5
0 9448 0
6645 6644 16075 18257 17092 6
0 13680 17310 25095 18485 30175 13684 2'.054 7
0 14025 U
9760 0
14875 0
16715 8
0 15383 17472 25438 21863 35763 16037 33669 9
0 14895 17130 25330 0
17445 9
18759 10 0
14035 0
9378 8717 23665 0
17269 11 0
15370 16775 24860 0
18420 0
19628 12 0
12820 8204 16878 6204 21983 13878 28673 (a) The number in parentheses beside the cycle number is the fuel cycle length in MWD /MTU.
3845e:1d/121285 A-4 I
TA8LE A-3 NORTH ANNA UNIT 1 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaged Relative Assembly Power Design Fuel Cycle Basis Assembiv Relative Power 1
2 3
4 5
1 1.00 0.72 0.89 0.68 0.74 0.48 2
0.83 0.57 0.73 0.38 0.47 0.30 3
1.21 0.85 1.02 0.91 0.90 0.66 4
0.86 0.59 0.74 0.43 0.38 0.34 5
0.92 0.66 0.76 0.45 0.39 0.40 6
0.98 0.95 0.94 1.02 1.01 0.92 7
1.10 0.97 1.17 1.05 1.05 0.98 8
1.00 1.06 0.91 1.12 1.11 1.19 9
1.05 1.04 0.92 1.19 1.17 1.14 10 1.08 0.98 1.17 1.08 0.97 1.03 11 1.06 1.07 1.03 1.22 1.18 1.20 12 0.95 0.89 0.80 0.96 0.89 0.89
TABLE A-4 NORTH ANNA UNIT 2 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaged Relative Assembly Power Design Fuel Cycle Basis Assembly Relative Power 1
2 3
4 w.
I 1.00 0.72 0.89 0.52 0.54 2
0.83 0.57 0.73 0.35 0.42 3
1.21 0.85 1.03 0.93 0.70 4
0.86 0.59 0.74 0.56 0.47 5
0.92 0.65 0.79 0.64 0.55 6
0.98 0.94 0.92 0.79 0.96 7
1.10 0.97 1.16 1.01 1.05 8
1.00 1.06 0.94 0.94 1.10 9
1.05 1.03 0.97 1.19 1,17 10 1.08 0.97 1.11 1.02 1.08 11 1.06 1.06 0.96 1.25 1.23 12 0.95 0.89 1.03 1.07 0.93 3845e:ld/121285 A-6
16003-8 0
(MAJOR AXIS)
BAFFLE
//
CORE BARREL
\\NN\\
l 2
\\
45 xNNNN p
6 7
3 4
N, 8
9 10 5
/
II I2 Figure A-1. North Anna Units 1 & 2 Core Description for Power Distribution Map A-7
APPENDIX B t
WELD CHEMISTRY Tables B.1-1 and B.1-2 provide the weld data output f rom the WOG Material Data Base. Given are the searches of all available data for the wire heat in the North Anna Units 1 and 2 reactor vessels beltline regioii.
r e pertinent material chemical compositions are given, along with the wire / flux identification. The mean chemistry values and the population standard deviation are then calculated. The mean values of copper and nickel are used in the RT analysis.
PTS Weld Chemistry Data Source and Plant:
Cu Weight % of Copper MATSURV NRC Mender MATSURV Data Base MPC Materials Properties Council Data Base Ni Weight % of Nickel RDM Rotterdam P
Weight % of Phosphorous SC Surveillance Capsule Si Weight % of Silicon SSP Ringhals 2 VGB North Anna 2 VRA North Anna 1 WQ Weld Qualification 3845e:1d/121285 B-1
TABLE B.1-1 North Anna Unit 1 Intermediate to Lower Shell Circumferential Weld Chemistry From WOG Materials Data Base - Wire Heat Number 25531 ID WIRE WIAE FLUM FLUX WELDCHEM Cu Na P
I HLAT T vF E TYPE LOT DATA Got FCV l
=-
0646 25531 SMIT 40 SMIT A9 1211 RDM,WO O.260 SSP NOZZLE TO INTER SHELL VRA SURVEILLANCE WELD 0721 25531 SMIT 40 SMIT 09 1211 VFA.SC O.086 0.110 0.020 0.350 SSP NOZZLE TO INTER SHELL f73 VR4 SURVF ILL ANCE WEL D l3 v
O.096000 O.I10000 O.020000 O.305000 mean O. OOOO* H) O. O*XKp'H t O.<N'MMOO O.063640
~
std.dev.
e
\\
\\
TABLE B.1-2 North Anna Unit 2 Intermediate to Lower Shell Circumferential Weld Chemistry From WOG Materials Data Base - Wire Heat Number 716126 SELECT REFORT ID WIRE WIRE FLUM Fl uX WELDCHE M Cu Na P
St PLANT DESCRIPTION HE AT TYFE TYPE LOT DATA SOURCE 0630 716126 S3MO LW320 26 RDN,WO O.061 0.030 0.012 0.237 0631 716126 S!NO LW320 26 RDM.WO O.062 0.030 0.013 0.227 0632 716326 S3NO LWT20 26 RDM.WQ o.079 0.040 0.010 0.227 0633 716126 S?MO LW320 26 RDM.WD O.064 0.n40 n.nts o.225 0634 716126 S3MO LW320 26 RDM,WQ O.060 0.000 0.016 0.188 o/48 716126 S?MO LW320 26 VGl*, SC O.088 0.004 0.017 0.250 f
og 0.069000 0.O*20667 0.015667 0.225667 mean o.O!!662 0.024712 0.002582 0.020704 std.dev.
...................................---..--...............--........-==............................................................
APPENDIX C 1
PT VALUES OF NORTH ANNA UNITS 1 AND 2 PTS R_EACTOR VESSEL BELTLINE REGION MATERIALS C.1 NORTH ANNA UNIT 1 Tables C.1-1 through C.1-5 provide the RT values, as a function of both PTS constant fluence and constant EFPY (assuming the projected fluences values),
for all beltline region materials of the North Anna Unit 1 reactor vessel.
The RT values are calculated in accordance with the FTS rule, which is PTS Reference [1] in the main body of this report.
The vessel location numbers in the following tables correspond to the vessel materials identified below and in Table 111.3-1 of the main report.
Location Vessel Material 1
Intermediate shell 04 2
Intermediate to lower shell circumferential weld WO4 3
Lower shell 03 C.2 NORTH ANNA UNIT 2 values, as a function of both
~
Tables C.2-1 through C.2-5 provide the RTPTS constant fluence and constant EFPY (assuming the projected fluence values),
for all beltline region materials of the North Anna Unit 2 reactor vessel.
The RT values are calculated in accordance with the proposed PTS rule, PTS which is Reference [1] in the main body of this report. The vessel location numbers in the following tables correspond to the vessel materials identified below and in Table 111.3-2 of the main report.
Location Vessel Material 1
Intermediate shell 04 2
Lower shell 03 3
Intermediate to lower shell circumferential weld WO4 3845e:ld/121285 C -1
TABLE C.1-1 18 2
RT VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS @ FLUENCE = 1.0 x 10 n/cm PTS CU i NI I P 1 I
I vALUE 1 TYPE IFLUENCE: RT
---.__--_---_________--_---__-_--___-----._-__--_---pyg LOC _IPLANTI 1
i vnA i.ial.s2i.oiol
- 17. I ACTUAL I B.M.
f.10E+191 108.1 2
i vnA i.o91.111.0201 is. i Actual C.w.
i.10E+191 86.1 ro 3
i vnA i.is
.sor.OO91
- 38. I ACTUAL 1 8.M.
- 1. TOE +191 141.1 1
1 Notes:
B.M. = Base Material (Shell)
C.W. = Circumferential Weld Reference Temperature are in F
TABLE C.1-2 18 2
RT VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS @ FLUENCE = 5.0 x 10 n/cm PTS 1 P f I I VALUE I TYPE iFLUENCri RT
..................................................-.......py3 LOC netAitif CU f NI 1
I VRA B.121.82.010:
- 57. I ACTUAL 1 8.M.
i.50E+19i 132.1 2
vna i.osi. tis.o2oi se. i AcTuat i c.w.
i.soE+tsi e7.
3
- vna i.isi.aoi.cosi
- 38. i AcTuat i s.u.
i.soE+is 17s.:
...................--.....................-~.............--
aW i
1
TABLE C.1-3 RT VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS 0 FLUENCE =1.0 x 10I9 2
PTS n/cm I TYPE IFLUENCE, RTPTS LOC iPLurri CU l NI I P I I
i VALUE 1
1 VRA 1.t21.821.0101
- 17. I ACTUAL I B.M.
1.10E+201 146.1 2
i vRA I.os
.itt.o2OI
- 19. I ACTUAL I C.W.
1.10E+2OI 103.1 3
8 va' I $st.80s.OO91
- 38. I ACTUAL i S.M.
f.10E+2OI 189.1 O
sb l
l
TABLE C.1-4 RT VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE MATERIALS @ CURRENT (4.7 EFPY) FLUENCE VALUES PTS 1
l I P I I
I VALUE I TYPE IFLUENCE I RT
___'C.B PL ANT I CU l NI LO
..........__...__.......__..............................P.T.S.
\\
l 1
I VRA 1.121.821.0101
- 17. I ACTUAL I 8.M.
l.74E+191 139.1 2
I VRA I.091.111.0201
- 19. 1 ACTUAL I C.W.
l.74E+191 100.I 3
i VRA I.151.801.0091
- 38. I ACTUAL 1 8.M.
l.74E+191 180.1 e
U1
TABLE C.1-5 RTPTS VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE MATERIAL @ LICENSE EXPIRATION (25.0 E l
l l
........_______....,,,,,,,,P,T,S, l
VALUE I TYPE IFLUENCE,RT LOC.lPLANTI CU l NI I P I
..-.....................I....I 1
I VRA 1.121.821.0101
- 17. I ACTUAL I B.M.
1.31E+2OI 175.1 l
l 2
I VRA i.Q91.111.0201
- 19. I ACTUAL 1 C.W.
l.31E+2OI 116.1 3
1 vRA I.151.801.0091
- 38. I ACTUAL I B.M.
I.31E+2OI 225.1 l
l e
G
TABLE C.2-1 IO 2
RT VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 0 FLUENCE = 1.0 x 10 n/cm PTS IPLANTI CU 1 NI I Pl I
I VALUE - I TYPE IFLUENCEI RT
.........................................................2IS LOC 1
I VGS I.091'.831.0111
- 75. I ACTUAL I 5.M.
I.f0E+191 154.1 2
I VG8 8.138.831.0131
- 56. I ACTUAL I a.M.
1.10E+191 152.1
............ -......--------------------~ ~-~~-~--" " "-"
3 i VS8 ' 878 05' ' **** ' ***""' ' '"'
1
+i9t t3't
... -............. - -... -.. - - - - - - - - - - - - - - - - ~ ~ - - - - - " " " " - " "
s N
l
)
TABLE C.2-2 18 2
RT VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 0 FLUENCE = 5.0 x 10 n/cm PTS I
I VALUE I TYPE IFLUENCEI RTf.TS LOC IPLawil Cu I n: I eI 1
I VG8 1.091.831.0111
- 75. I ACTUAL I B.M.
1.50E+191 171.1 2
I vG8 I.tst.8al.Otal 5s. I ACTUAL i S.M.
l.50E+191 178.1
. - -.... - -...... -. - - - - - - -.... - -.... - -.... ~... -.. - - - - -. - - -
3 I va8 I.071.051.0161
-48.
I ACTUAL I C.W.
I.50E+191 20.1 l
m e
l l
l 1
TABLE C.2-3 l9 2
VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 0 FLUENCE = 1.0 x 10 n/cm RTPTS LOC f FLANTI CU i NI I P 1 I
I VALUE I TYPE IFLUENCEI RTo
.........--..............................................I.T.S 1
1 VGB I.091.831.0111
- 75. I ACTUAL I B.M.
1.10E+2OI 181.1 I VG8 I.131.831.0131
- 56. I ACTUAL I B.M.
1.10E+2OI 193.1 l
2 3 I vca I.o71.ost.oist
- 48. I ACTUAL I C.W.
1.10E+2OI 24.1 e
(O 1
TABLE C.2-4 RT VALUES FOR NORTH UNIT UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 0 CURRENT (3.5 EFPY) FLUENCE VALUES PTS RT PLANT: cu we i ei I
- vAtut i Type irtuENCE:
LOC.........................................................2IS I
I VGB I.oel.s3I.ofil
- 75. I ACTUAL I B.M.
f.53E+191 172.1 l
2 i VG8 I.131.531.o131
- 56. I ACTUAL I B.M.
1.53E+198 179.8 l
3 1 vos
.o75.ost.ots: -4s. i Actual : C.W.
1.53E+191 2o.f t
e-a C
N i
M
l TABLE C.2-5 VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATER RTPTS TYPE FLUENCE' N PTS CU 1 NI P 1 I
i VALUE LOC.iPLANT:...................................----....................
1 i VG8 :.091.s3i.Otti 75.
ACTUAL i B.M.
i.34E+20: 204.1 2
vca
. tai.sai.ota ss. i actual i s.u.
i.34E+2oi 227.
3 i vca
.o7:.osi.otsi -4s. i 4CTual i c.w.i.34E+2oi 34.i c,
.I.a Da l
1 l
I l
l l
l l
l l
i i
l l
i l
l l
l l
l