ML20237C058
| ML20237C058 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna, 05000000 |
| Issue date: | 12/31/1985 |
| From: | Furchi E, Meyer T, Perone V, Weaver M, Wrights G WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20140D070 | List: |
| References | |
| TAC-59988, TAC-59989, WCAP-11015, NUDOCS 8712190010 | |
| Download: ML20237C058 (89) | |
Text
- - ___ - _ _
WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-11015 CUSTDMER DESIGNATED DISTRIBUTION i
m et i
SURRY UNITS 1 AND 2 REACTOR VESSEL FLUENCE AND RT EVALUATIONS PTS s
E. L. Furchi V. A. Perone M.' Weaver G. N. Wrights i
Work Performed for Virginia Power Company December 1985 o
APPROVED:
bM APPROVED:
T. A. Meyer, Nanager F. L. Lau, Manager Structural Materials Radiation and Systems and Reliability Technology Analysis APPROVED:
[M 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
$2]go O
P. O. BOX 355 P
PITTSBURGH, PENNSYLVANIA 15230 3844e:1d/121385
TABLE OF CONTENTS
)
5 PAGE i
TABLE OF CONTENTS 11 LIST OF TABLES v
LIST OF FIGURES 1
I.
INTRODUCTION 1
I.1 The Pressurized Thermal Shock Rule 3
I.2 The Calculation of RTPTS 5
II.
NEUTRON EXPOSURE EVALUATION 5
11.1 Method of Analysis 8
11.2 Fast Neutron Fluence Results 35 III.
MATERIAL PROPERTIES 111.1 Identification and Location of Beltline Region Materials 35
.III.2 Definition and Source of Material Properties for All 35 Vessel Locations 36 III.3 Summary of Plant-Specific Material Properties IV.
DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE 42 REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RTPTS 42 versus Fluence Results 43 IV.2 Discussion of Results 48 V.
CONCLUSIONS AND RECOMMENDATIONS 50 VI.
REFERENCES VII.
APPENDICES A-1 A.
Power Distribution B-1 B.
Weld Chemistry C-1 C.
RTpis Values of Surry Units 1 and 2 Reactor Vessel Beltline Region Materials 3844e:ld/121385 i
LIST OF TABLES Page 12 11.2-1 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel Inner Radius - 0* Azimuthal Angle 13 11.2-2 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel Inner Radius - 15' Azimuthal Angle 11.2-3 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 14 Pressure Vessel Inner Radius - 30* Azimuthal Angle 15 11.2-4 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel Inner Radius - 45' Azimuthal Angle 11.2-5 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 16 15* Surveillance Capsule Center 11.2-6 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 17 25' Surveillance Capsule Center 18 11.2-7 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 35'Sgg 11ance Capsule Center 11.2-8 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 19 45' Surveillance Capsule Center 11.2-9 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 20 Pressure Vessel Inner Radius 0* Azimuthal Angle I I. 2-10 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 21 Pressure Vessel Inner Radius 15* Azimuthal Angle 11.2-11 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 22 Pressure Vessel Inner Radius 30* Azimuthal Angle 11.2-12 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 23 Pressure Vessel Inner Radius 45' Azimuthal Angle 11.2-13 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 24 15' Surveillance capsule Center 11.2-14 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 25 25* Surveillance Capsule Center 11.2-15 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 26 35' Surveillance Capsule Center 11.2-16 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 27 45' Surveillance Capsule Center 3844e:1d/121385 ii
F LIST OF TABLES (Continued)
Page III. 3-1 Surry Unit 1 Reactor Vessel Beltline Region Material 38 Properties 111.3-2 Surry Unit 2 Reactor Vessel Beltline Region Material 39 Properties I V.1 -1 RTPTS Values for Surry Unit 1 44 IV.1 -2 RTPTS Values for Surry Unit 2 45 A-1 Surry Unit 1 Beginning-of-Cycle and End-of-Cycle Fuel A-3 Assembly Burnups A-2 Surry Unit 2 Beginning-of-Cycle and End-of-Cycle Fuel A-5 Assembly Burnups A-3 Surry Unit 1 Core Power Distributions Used in the Fluence A-7 Analysis A-4 Surry Unit 2 Core Power Distributions Used in the Fluence A-9 Analysis B.1-1 Surry Unit 1 Intermediate and Lower Shell Longitudinal B-2 Welds Chemistry From WOG Materials Data Base -
Wire Heat Number 8T1554 B.1-2 Surry Unit 1 Beltline Circumferential Weld and Surry Unit 2 B-3 Intermediate Shell Longitudinal Weld Chemistry From WOG Materials Data Base - Wire Heat Number 72445 B-5 l
B.1-3 Surry Unit 2 Lower Shell Longitudinal Weld Chemistry From WOG Material Data Base - Wire Heat Number 8T1762 B.1 -4 Surry Unit 2 Beltline Circumferential Weld Chemistry From B-6 i
WOG Materials Data Base - Wire Heat Number 0227
(
C.1-1 RT TS Values for Surry Unit 1 Reactor Vessel Beltline C-3 i
Re ion Materials @ Fluence = 1.0 x 1018 n/cm2 Values for Surry Unit 1 Reactor Yessel Beltline C-4 RTPTS C.1-2 Region Materials @ Fluence = 5.0 x 10 8 n/cm2 1
i C.1-3 RT TS Values for Surry Unit 1 Reactor Vessel Beltline C-5 i
Re ion Materials @ Fluence = 1.0 x 1019 n/cm2 RTpy3 Values for Surry Unit 1 Reactor Vessel Beltline C-6 C.1-4 Region Materials @ Current (7.4 EFPY) Fluence Values 3844e:ld/121385 iii f
LIST OF TABLES (Continued)
Page A
Values for Surry Unit 1 Reactor Vessel Beltline C-7 C.1-5 RTPTS Region Materials @ License Expiration (25.6 EFPY)
Values for Surry Unit 2 Reactor Vessel Beltline C-8 C.2-1 RTPTS Region Materials @ Fluence = 1.0 x 1018 n/cm2 Values for Surry Unit 2 Reactor Vessel Beltline C-9 C.2-2 RTPTS Region Materials @ Fluence = 5.0 x 1018 n/cm2 Values for Surry Unit 2 Reactor Vessel Beltline C-10 C.2-3 RTPTS Region Materials @ Fluence = 1.0 x 1019 n/cm2 Values for Surry Unit 2 Reactor Vessel Beltline C-11 C.2-4 RTPTS Region Materials @ Current (7.6 EFPY) Fluence Values Values for Surry Unit 2 Reactor Vessel Beltline C-12 C.2-5 RTPTS Region Materials @ License Expiration (25.8 EFPY)
_L P
3844e:ld/121385 iv
i LIST OF FIGURES PAGE
.'4 11.1-1 Surry Reactor Geometry 28 11.2-1 Surry Unit 1 Maximum Fast Neutron (E>l.0 MeV) Fluence 29 at the Beltline Weld Locations as a Function of Full Power Operating Time II.2-2 Surry Unit 2 Maximum Fast Neutron (E>l.0 MeV) Fluence 30 at the Beltline Weld Locations as a Function of Full Power Operating Time 11.2-3 Surry Unit 1 Maximum Fast Neutron (E>1.0 MeV) Fluence 31 at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 11.2-4 Surry Unit 2 Maximum Fast Neutron (E>1.0 MeV) Fluence 32 at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle II.2-5 Surry Units 1 and 2 Relative Radial Distribution 33 of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 11.2-6 Surry Units 1 and 2 Relative Axial Distribution of 34 Fast Neutron (E>l.0 MeV) Flux and Fluence Within the l
Pressure Vessel Wall l
III.1-1 Identification and Location of Beltline Region Material 40 i
for the Surry Unit Reactor Vessel 111.1-2 Identification and Location of Beltline Region Material 41 l
for the Surry Unit Reactor Vessel (g)
IV.1 -1 Surry Unit 1 - RTpis Curves per PTS Rule Method [1]
46 Docketed Base Material and WOG Data Base Mean Weld Material Properties I V.1 -2 Surry Unit 2 - RTPTS Curves per PTS Rule Method [1]
47 Docketed Base Material and WOG Data Base Mean Weld Material Properties A-1 Surry Units 1 and 2 Core Description for Power A-ll Distribution Map i
1-l l
l l
l 3844e:ld/121385 y
l
SECTION I INTRODUCTION The purpose of this report is to submit the reference temperature for pressurized thermal shock (RTPTS) values for the Surry Units 1 and 2 reactor vessels to address the Pressurized Thermal Shock (PTS) Rule.
Section I discusses the Rule and provides the methodology for calculating RTPTS*
Section II 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.
I.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 bec'ame 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 surf ace, thereby potentially affecting the integrity of the vessel.
3844e:1d/121385 1
The Rule establishes the following requirements for all domestic, operating PWRs:
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 un 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 licenses 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, i
areko For applicants of operating licenses, values of the projected RTPTS 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 toughness state of the reactor vessel materials - the " reference temperature for nil ductility transition" (RTNDT).
For purposes of the Rule, RT 5
NDT now defined as "the reference temperature for pressurized thermal shock" Each (RTPTS) and calculated as prescribed by 10 CFR 50.61(b) of the Rule.
USNRC licensed PWR must submit a projection of RT values from the time of pg the submittal to the license expiration date. This assessment must be submitted within 6 months after the effective date of the Rule, on January 23,
~
1986, with updates whenever changes occur affecting projected values.
The 3844e:ld/121385 2
calculation must be made for each weld and plate, or forging, in the reactor values vessel beltline. The purpose of th% report is to provide the RTPTS a
for Surry Units 1 and 2.
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 are actuaHy The prescribed equations in the PTS rule for calculating RTPTS For the purpose of comparison with one of several ways to calculate RTNDT.
f r the reactor vessel must be the Screening Criterion, the value of RTPTS 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.
Equation 1:
PTS = 1 + M + [-10 + 470(Cu) + 350(Cu)(Ni)] f.270 0
RT Equation 2:
= 1 + M + 283 f.194 0
PTS where
! = 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: 0*F for welds made with Linde 80 flux, and -56*F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.
3844e:ld/121385 3
M = the margin to be added to cover uncertainties in the values of initial In NDT, c pper and nickel content, fluenca, and calculation procedures.
RT Equation 1, M=48'F if a measured value of I was used, and M=59'F if the 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 material.
f = the maximum neutron fluence, in units of 10 n/cm2 (E greater than or I9 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 valvas 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 Surry Units 1 and 2 are discussed PTS in Section 111.2.
i 3844e:1d/121385 4
SECTION 11 NEUTRON EXPOSURE EVALUATION g
This section presents the results of the application of Westinghouse derived
=
adjoint importance functions to the calculation of the Surry 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 Surry Units 1 and 2 reactor geometry at the core midplane is shown in Figure II.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 O*.
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 f rom 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 the response of interest (neutron flux > 1.0 Mev) at several selected locations within the reactor geometry to the power distributions in the 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.
The forward transport calculation was carried out in R,0 geometry using the DOT discrete ordinates code [2] and the SAILOR cross-section library [3).
The 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 the cross-sections. An S angular quadrature was used.
3 6
3844e:1d/121385 5
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
~j 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.
cmss-secdon The adjoint analyses were also carried out using the P3 approximation f rom the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions
_i along the inner radius of the pressure vessel. Again, these calculations were l
run in R,0 geom' try to provide power distribution importan~ce f unctions for e
the exposure parameter of interest (neutron flux > 1.0 MeV). Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as-R
- I I
I 0,0 F @,0,Q M R @ de R,0 R s E where:
esponse of intenst M U > 1.0 %) at radus R and R
=
R,0 azimuthal angle e.
1 Adjoint importance function at racius R and azimuthal I (R,0,E)
=
angle e for neutron energy group E.
Full power fission density at radius R and azimuthal angle F (R,0,E)
=
e for neutron energy group E.
3844e:1d/121385 6
L
The fission density distributions used reflect the burnup-dependent inventory of fissioning actinides, including U-235, U-238, Pu-239, and Pu-241.
a Core power distributions for use in the plant specific fluence evaluations for Surry Units 1 and 2 were derived f rom measured assembly 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 the Surry Units expires on June 25, 2008 (forty years after the construction permit was issued).
This report includes fluence projections from September 30, 1985 to June 25, 2008 using the cycle-averaged core power distribution of the current operating
'k cycle (Cycle 8 for each Surry Unit) and an assumed future capacity f actor of 80%. All fluence projections into the future reflect the low leakage fuel management strategy exemplified by the Cycle 8 core loadings.
Finally, it has been assumed that the Surry cores will be uprated from 2441 MW to 2546 th MW at the beginning of Cycle 11.
th The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oakridge National Laboratory (ORNL) Poolside Critical Assembly (PCA) facility as well as against 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 conservative by 7-22%. When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that fluence predictions are within 15% of measured values at surveillance capsule locations.
3844e:1d/121385 7
II.2 FAST NEUTRON FLUENCE RESULTS Calculated fast neutron (E >1.0 MeV) exposure results for Surry Units 1 and 2 are presented in Tables 11.2-1 through 11.2-16 and in Figures 11.2-1 through 11.2-6.
Data is presented at several azimuthal locations on the inner 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 Surry 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 co~re power distribution at the nominal + 2a level.
Similar data for the center of surveillance capsu'les located at 15*, 25*, 35' and 45* are given in Tables 11.2-5 and 11.2-8, respectively.
In addition to the calculated data given for the surveillance capsule locations, measured fluence data f rom previously withdrawn surveillance capsules are also presented for comparison with analytical results.
In the case of Unit 1, capsules were removed f rom the 15* location at the end of cycle 1 and the 35' location at the end of cycle 4.
Cycle-specific and design basis fast neutron flux and fluence data at the inner radius of the pressure vessel of Surry Unit 2 are given in Tables II.2-9 through 11.2-12 for the period of operation up to September 30, 1985, 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 surveillance capsule locations are given in Tables II.2-13 and 11.2-16.
For Unit 2, a surveillance capsule was removed from the 15* position following cycle 1.
A dosimetry evaluation from this capsule withdrawal is listed in Table 11.2-13.
l 1
3844e:ld/121385 8
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.
For both Surry units, calculated plant specific fast neutron (E > 1.0 MeV) fluence levels at the surveillance capsule center are in excellent agreement with measured data. The maximum difference between the plant specific calculations and the measurements is less than 8%.
Differences of this magnitude are well within the uncertainty of the experimental results.
2.
For both Surry units, the fast neutron (E > 1.0 MeV) flux incident on the pressure vessel during Cycle 1 was, on the average,14% 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 8 of Surry Unit 1, which is used for projection into the future, has reduced the peak f ast neutron flux (0* azimuthal position) on the pressure vessel by a factor of i
1.47 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 8 core loading produced flux reduction f actors ranging f rom 1.45 to 1.61 at the other azimuthal locations.
4.
In Surry Unit 2, the low leakage core loading used for projection into the future (cycle 8) yielded a flux reduction factor of 1.85 at the peak flux location and f actors ranging f rom 1.35 to 1.59 at the remaining azimuths.
5.
Comparing the flux reduction factors resulting from the cycle 8 low leakage core loadings in the Surry units, one observes differences that are attributable to the varying burnups of the fuel asseublies in peripheral locations (see burnup data in Appendix A).
3844e:1d/121385 9
6.
While the peak fast neutron flux location occurs at the 0* location in both Surry units, the materials having the most limiting RT values PTS (see Section IV) are not necessarily located in the peak flux.
In Surry Unit 1, the limiting material is the circumferential weld which sees the peak flux at the 0* azimuthal position. On the other hand, the limiting material in Surry Unit 2 is a longitudinal weld which is situated at the 45' azimuth.
Graphical presentations of the plant specific fast neutron fluence at key locations on the pressure vessel are shown in Figures 11.2-1 and 11.2-2 as a function of full power operating time for Surry Units 1 and 2, respectively.
For both Units 1 and 2, pressure vessel data is presented for the 0* location on the circumferential weld as well as for the 45* longitudinal welds (see Section III.1).
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 core loadings as of September 30, 1985. The dashed portions of these curves, however, involve a projection
~
into the future. As mentioned in Section II.1, the neutron flux average over cycle 8 of each Surry unit was used to project futtire fluence levels.
It should be noted that implementation of a more severe low leakage pattern than that used in cycle 8 of each unit 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 projections. The RT assessment must be updated per 10CFR50.61(b)(1)
PTS whenever, among other things, changes in core loadings significantly impact the fluence and RT projections.
PTS In Figures II.2-3 and 11.2-4, the azimuthal variation of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented as a function of azimuthal angle for Units 1 and 2, respectively.
Data are presented for both current and projected expiration-of-operating-license conditions.
In Figure II.2-5, the relative radial variation of fast neutron flux and fluence within the pressure vessel wall is presented. Similar data 3844e:1d/121385 10 L
l
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 4
three-dimensional description of the f ast 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 4(R, e,Z) = 4(e) F(R) G(Z)
Fast neutron fluence at location R, e, Z within where: 4 (R,e,Z)
=
the pressure vessel wall Fast neutron fluence at azimuthal location e on 4 (e)
=
the pressure vessel inner radius from Figure 11.2-3 or 11.2-4 Relative fast neutron flux at depth R into the F (R)
=
pressure vessel from Figure 11.2-5 Relative fast neutron flux at axial position Z from G (2)
=
Figure 11.2-6 Analysis has shown that the radial and axial variations within the vessel wall 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.
1 3844e:1d/121385 11
TABLE 11.2-1 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE (a)
Deltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg Interval Time (EFPY)
(n/cm -sec_1 SDecific Basis 10 18 18 CY-1 1.1 5.03 X 10 1.70 x 10 2.02 x 10 10 18 18 CY-2 1.6 5.73 x 10 2.70 x 10 3.06 x 10 10 18 18 CY-3 2.3 5.22 x 10 3.87 x 10 4.39 x 10 10 18 18 CY-4 3.4 4.86 x 10 5.49 x 10 6.37 x 10 10 18 18 CY-5 4.6
- 4.40 x 10 7.10 x 10 8.54 x 10 10 18 I9 CY-6 5.9 3.96 x 10 8.75 x 10 1.10 x 10 10 I9 I
CY-7 6.8 5.91 x 10 1.05 x 10 1.28 x 10 '
7.4 4.05 x 10
- 1.13 ' x 10 1.39 x 10 '
10 I9 I
CY-8 (9/30/85)(c) 9/30/85 4 CY-10(d) 10.3 4.05 x 10 1.49 x 10 '
1.93 x 10 '
10 I
I 10 I9 I9
~
- 6/25/2008(')
25.6 4.22 x 10 3.54 x 10 4.94 x 10 (a) Applicable to the peak locations (0*, 90', 180', 270') on the intermediate j
and lower shell plates and the intermediate to lower shell circumferential l
weld.
(b) Design basis fast neutron flux = 5.96 x 1010 n/cm2-sec at 2441 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined.
(d) At the beginning of CY-11, the core thermal power will be uprated to 2546 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.
1 3844e:1d/121385 12 l
TABLE 11.2-2 s
SURRY UNIT 1 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. Flux Plant Desig A)
Interval Time (EFPY)
(n/cln ~sec)
Specific Basis 10 CY-1 1.1 2.40 X 10 8.12 x 10 9.21 x 10 10 18 8
CY-2 1.6 2.72 x 10 1.29 x 10 1.40 x 10 18 CY-3 2.3 2.49 x 10 1.84 x 10 2.00 x 10 10 18 18 CY-4 3.4 2.34 x 10 2.62 x 10 2.91 x 10 10 8
I CY-5 4.6 2.06 x 10 3.38 x 10 3.90 x 10 10 18 18 CY-6 5.9 1.88 x 10 4.16 x 10 5.04 x 10 8
18 CY-7 6.8 2.50 x 10 4.92 x 10 5.87 x 10 10 18 18 CY-8 (9/30/85)I )
7.4 1.88 x 19 5.28 x 10 6.38 x 10 10 18 18 9/30/85 4 CY-10(c) 10.3 1.88 x 10 6.96 x 10 8.82 x 10 I9 I9 Id) 25.6 1.97 x 10 1.65 x 10 2'.26 x 10
~CY-11 4 6/25/2008 10 2
n/cm -sec at 2441 MWth (a) Design basis fast neutron flux = 2.72 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 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.
3844e:1d/121385 13
TABLE 11.2-3 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg}
Interval Time (EFPV)
(n/cm -sec)
Specific Basis 10 II I7 CY-1 1.1 1.30 X 10 4.40 x 10 5.38 x 10 CY-2 1.6 1.54 x 10 7.09 x 10 8.16 x 10 10 18 18 CY-3 2.3 1.34 x 10 1.01 x 10 1.17 x 10 CY-4 3.4 1.30 x 10 1.44 x 10 1.70 x 10 10 18 18 CY-5 4.6 1.09 x 10 1.84 x 10 2.28 x 10 CY-6 5.9 1.02 x 10 2.27 x 10 2.95 x 10 9
18 18 CY-7 6.8 9.80 x 10 2.56 x 10 3.43 x 10 9
18 18 CY-8 (9/30/85)(b) 7.4 9.86 x 10 2.75 x 10 3.73 x 10 9/30/85 4 CY-10 10.3 9.86 x 10 3.63 x 10 5.15 x 10 '
IC) 8 I
C Y-11 4 6/25/2008 25.6 1.03 x 10 8.62 x 10 1.32 x 10 '
18 I
I l
10 2
n/cm -sec at 2441 MWth l
(a)
Design basis f ast neutron flux = 1.59 x 10 l
(b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period from the onset of the uprating to the license expiration l
date.
l i
14 3844e:1d/121385
TABLE 11.2-4 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE "}
I Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigo )
Interval Time (EFPY)
(n/cm -sec)
Specific BasisID I
8.59 X 10' 2.91 x 10 3.27 x 10 CY-1 1.1 10 I
CY-2
- 1. 6' 1.05 x 10 4.75 x 10 4.96 x 10 I
I CY-3 2.3 9.08 x 10' 6.78 x 10 7.12 x 10 9
CY-4 3.4 8.71 x 10 9.68 x 10 1.03 x 10 9
18 18 CY-5 4.6 7.11 x 10 1.23 x 10 1.39 x 10 I
18 6.86 x 10' 1,51 x 10 1.79 x 10 CY-6 5.9 18 6.14 x 10' 1.70 x'10 2.08 x 10 CY-7 6.8 CY-8 (9/30/85)(c) 7.4 6.54 x 10 1.82 x 10 2.26 x 10 '
9 18 l
9 18 18 9/30/85 4 CY-10(d) 10.3 6.54 x 10 2.41 x 10 3.13 x 10 I
8 5
6.82 x 10' 5.71 x 10 8.01 x 10 CY-11 4 6/25/2008 '}
25.6 (a) Applicable to the longitudinal welds at 45*, 135', 225*, 315* in the peak axial flux.
(b) Design basis fast neutron flux = 9.66 x 109 n/cm2-sec at 2441 MWth (c) 9/30/85 is tre date at which the current neutron fluences are defined.
(d) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity f actor is assumed.
(e) Exposure period from the onset of the uprating to the license expiration date.
3844e:1d/121385 15
TABLE 11.2-5 e1 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15' SURVEILLANCE t
CAPSULE CENTER
)
I
{
8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Capsule Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data CY-1 1.1 8.31 X 10 2.81 x 10 3.19 x 10 2.89 x 1018(ej 10 18 18 I
CY-2 1.6 9.42 x 10 4.46 x 10 4.84 x 10 10 18 18 CY-3 2.3 8.61 x 10 6.39 x 10 6.95 x 10 CY-4 3.4 8.11 x 10 9.08 x 10 1.01 x 10 10 I9 I9 CY-5 4.6 7.08 X 10 1.17 x 10 1.35 x 10 CY-6 5.9 6.47 x 10 1.44 x 10 1.75 x 10 10 I9 I9 CY-7 6.8 8.76 x 10 1.70 x 10 2.03 x 10
-/
10 I9 I9 CY-8 (9/30/85)(b) 7.4 6.46 x 10 1.83 x 10 2.21 x 10 10 I9 I9 9/30/85 + CY-10(c) 10.3 6.46 x 10 2.40 x 10 3.05 x 10
}
CY-11 + 6/25/2008 25.6 6.74 x 10 5.67 x 10 7.82 x 10 2
10 n/cm -sec at 2441 MWth (a)
Design basis fast neutron flux = 9.43 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the b? ginning of CY-11, the core thermal power will be uprated to 2546 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) Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 5.
De f
16 f
3844e:1d/121385
1
.\\
)
N:
s s.
1 i
.idE11.2-6
\\'
DRRY UNIT 1 FAST NEUTRON (E > 1.0 H n _fxP050RE AT THE 25" SURVElttANCE
,(f.h3ICENTER
~
I-Beltlhe Region 2
Elapsed Cumulatise fluen:e (n/cm )
s Irradiation sl Irradiation Avg.2 Flux Plant '
Desigg Interval
_a Time ir T ),
(n/cm -sec_1 Specific Basis t
I CY-1 1.i
'.i.26 x 10 1.78 x 10 2.02 x 10 CY-2 i 1.6 6.14 x 10 2.85 x 10 3.07 x 10 10 18 CY-3 2.3 5.40 x 10 4.06 x 10 4.41 x 10 '
18 I9 '
CY-4 3.4
' t.24 x 1010,,gg) x 3g 6.40,x 10 10 18 18 4.6 4.48 x 10 7.44 x 10 8.58 x 10 CY-5 0
18 I9 CY-6 5.9 4.15 x 10 9.17 x 10 1.11 x 10 CY-7 i f.B i,
4.17 x 10 1.04 x 10" 1.29 x 10 10 l
10 I9 I9 CY-8 (9/30/85)(b)
C'. 4 4.02 x 10 1.12 x 10 1.40 x 10 10 I9 9/30/85 + CY-10(c) l'O. 3 4.02 x 10 1.48 x 10 1.94 x 10 CY-11 4 6/25/2008(
25.6 4.20 x 10 3.52 x 10 4.96 x 10 (a) Design basis fast neutron flux = 5.98 x 1010 n/cm2-sec at 2441 MWO (b) 9/30/85 is the date at which the current neutron fluences are defined.
T (c) At the beginning of CY '1, the core thermal power will be uprated to 2546 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.
4 i
~
s f
?t 3844e:1d/121385 17
i
,7 -
n 9,
TABLE II.2-7 l
SURRY UNIT 1 f,AST NEUTRON (E > 1 Q,heV) EXPOSORE AT THE 35' JMRVE1LLANCE.
2 CAPSULE CENTER
/
BeltiknaRegion
'H!
Elapted Cumulativg Fluence (n/cm )
'l Ddsigga)
Capsule Irradiation Irradiation Avg.gFlux Plant
~
Data huis Interval Time (EFP Q
[n/cin -sec)
Specific _
l0 CY-1 1,.1
. 3.% % 10 1.21 x 10 1.33 x 10
/
18 18 CY-2
- 1. 6 4.23 x-10
?.95 x 10 2 10 x 10 18 18 CY-3 2,. 3
'3.71 x 10 2.78 x.10 3.02 x'10 18 M
18 (e CY-4 J.4 3.58 x 10 3.97 x 10 f,,3g x )g 4.31 x 10 18 CY-5 4.6 2.93 X 10
-5.05 x 10 5.87 x'1D 10 18 CY-6 5.9 2.78 x 10 6.21 x 10 7.58 x 10 10 0
CY-7 6.8 2.58 x 10 6.99 x 10 B.82 x 10 3
}
CY-8 (9/30/85) 7.4 2.68 x 10 7.50 x 1D 9.59't iO D
18 I9 9/30/85 + CY-10(C) 10.3 2.68 x 10 9.90 x 10 1.33 x 10 I9 I9 CY-11 + 6/25/2008(d) 25.6 2.80 x 10 2.35 x 10 3.39 x 10 (a)
Design basis fast neutron flux = 4.09 x 1010 n/c',r2-sec~at 2441 MWth (b) 9/30/85 is the date at which the current neutrcn fluences are detined.
(c)
At the beginning of CY-11, the core thermal power will be uprated to 2'446 MWth.
Beyond 9/30/B3 a 80% capacit/ factor is assumea.
(d)
Exposure period frc,t the onset of the uprating to the license expiricion date.
l I
(e)
Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 6.
l l
i 38440:1d/121385 18 l
I
,t 2
J
t
- q TA8LE II.2-8 o
SURRY UNIT 1 i'
FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45' SURVEILLANCE CAPSULE CENTER lT Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation-Avg.2 Flux Plant Desigg}
Interval Time (EFPY)
(n/cm -sec)
Specifir Basis 10 18 CY-1 1.1 2.79 x 10 9.45 x 10" 1.09 x 10 10 18 18
'c CY-2 1.6 3.43 x 10 1.55 x 10 1.66 x 10 8
CY-3 2.3 2.95 x 10 2.21 x 10 2.38 x 10 10 18 18 CY-4 3.4 2.83 x 10 3.15 x 10 3.46 x 10 10 18 18 CY-5 4.6 2.29 x 10 3.98 x 10 4.64 x 10 10 18 18-CY-6 5.9 2.21 x 10 4.91 x 10 5.98 x 10 10 18 18 CY-7 6.8 1.97 x 10 5.50 x 10 6.97 x 10 10 18 18 CY-8 (9/30/85)(b) 7.4 2.10 x 10 5.90 x 10 7.58 x 10 10 I9 I9 9/30785 4 CY-10(C) 10.3 2.10 x 10 7.78 x 10 1.05.x 10 10 I9 I9 CY-11 4 6/25/2008(d) 25.6 2.19 x 10 1.84 x 10 2.68 x 10
,k (a) Design basis fast neutron flux = 3.23 x 1010 2
n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 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 3844e:1d/121385 19
TABLE II.2-9 SURRY UNIT 2 FAST NEUTRON (E-> 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - O' AZIMUTHAL ANGLE (a)
Beltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desig[b)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.2 4.96 X 10 1.84 x 10 2.21 x 10 CY-2 1.9 5.16 x 10 3.02 x 10 3.57 x 10
~
10 18 18 CY-3 2.7 4.97 x 10 4.20 x 10 4.99 x 10 CY-4 3.8 5.21 x 10 6.01 x 10 7.06 x 10 10 18 18 CY-5 4.9 4.35 x 10 7.56 x 10 9.18 x 10 1.16 x 10 '
CY-6 6.2 4.16 x 10 9.25 x 10 10 I9 I9 CY-7 7.4 4.67 x 10 1.10 x 10 1.38 x 10 10 I9 I9 CY-8 (9/30/85)(C) 7.6 3.22 x 10 1.12 x 10 1.43 x 10 10 I9 I9 9/30/85 + CY 10(d) 10.7 3.22 x 10 1.44 x 10 2.01 x 10 CY-11 -* 6/25/2008 '}
25.8 3.36 x 10 3.04 x 10 4.97 x 10 '
I I
(a) Applicable to the peak locations (0*, 90',180*, 270') on the intermediate and lower shell plates and the intermediate to lower shell circumferential weld.
(b)
Design basis fast neutron flux = 5.96 x 1010 n/cm2-sec at 2441 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined.
(d) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. 8eyond 9/30/85 a 80% capacity factor is assumed.
l l
(e)
Exposure period from the onset of the uprating to the license expiration i
date.
.l l
i 3844e:1d/121385 20 f
TABLE 11.2-10 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15* A7IMUTHAL ANGl_E 8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigga)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 0
8 CY-1 1.2 2.37 X 10 8.81 x 10 1.01 x 10 8
8 CY-2 1.9 2.52 x 10 1.45 x 10 1.63 x 10 10 18 18 CY-3 2.7 2.49 x 10 2.05 x 10 2.28 x 10 CY-4 3.8 2.50 x 10 2.92 x 10 3.22 x 10 10 18 I
CY-5 4.9 2.06 x 10 3.65 x 10 4.19 x 10 CY-6 6.2 2.00 x 10 4.46 x 10 5.30 x 10 10 18 18 CY-7 7.4 2.10 x 10 5.25 x 10 6.32 x 10 10 18 18 CY-8 (9/30/85)(b) 7.6 1.72 x 10 5.37 x 10 6.51 x 10 10 18 18 9/30/85 + CY-10(c) 10.7 1.72 x 10 7.06 x 10 9.17 x 10 CY-11 4 6/25/2008(
25.8 1.80 x 10 1.56 x 10 2.27 x 10 '
10 I9 I
10 2
a/cm -sec at 2441 MWth (a) Design basis f ast neutron flux = 2.72 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth.
8eyond 9/30/85 a 80% capacity factor is assumed.
(d)
Exposure period f rom the onset of the uprating to the license expiration date.
3844e:1d/121385 21
TABLE 11.2-11 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE Beltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desig Interval Time (EFPY)
(n/cm -sec)
Specific _
Basis a)
CY-1 1.2 1.30 X 10 4.81 x 10" 5.91 x 10" 10 8.01 x 10" 9.53 x 10 CY-2 1.9 1.41 x 10 10 18 18 CY-3 2.7 1.45 x 10 1.15 x 10 1.33 x 10 I
18 18 CY-4 3.8 1.38 x 10 1.63 x 10 1.88 x 10 10 18 18 CY-5 4.9 1.13 x 10 2.03 x 10 2.45 x 10 10 CY-6 6.2 1.14 x 10 2.49 x 10 3.10 x 10 10 I0 18 CY-7 7.4 1.07 x 10 2.89 x 10 3.69 x 10 10 18 18 CY-8 (9/30/85)(b) 7.6 1.00 x 10 2.96 x 10 3.80 x 10 IC) 10 18 18 9/30/85 + CY-10 10.7 1.00 x 10 3.95 x 10 5.36 x 10 10 18 I9 6/25/2008( }
25.8 1.05 x 10 8.94 x 10 1.33 x 10 CY-11 4 2
10 n/cm.sec at 2441 MWth (a) Design basis fast neutron flux = 1.59 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 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
3844e:1d/121385 22
TABLE 11.2-12 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE "
8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigo )
Interval Time (EFPY)
(n/cm -sec)
Specific Basistb CY-1 1.2 8.61 X 10 3.20 x 10" 3.59 x 10" CY-2 1.9 9.74 x 10 5.41 x 10 5.79 x 10 I
CY-3 2.7 9.86 x 10 7.76 x 10" 8.09 x 10 8
CY-4 3.8 9.20 x 10 1.10 x 10 1.14 x 10 18 18 CY-5 4.9 7.95 x 10' 1.38 x 10 1.49 x 10 CY-6 6.2 7.70 x 10 1.69 x 10 1.88 x 10 8
8 CY-7 7.4 7.66 x 10 1.98 x 10 2.24 x 10 9
18 18 CY-8 (9/30/85)(c) 7.6 7.14 x 10 2.03 x 10 2.31 x 10 9/30/85 4 CY-10(d) 10.7 7.14 x 10 2.73 x 10 3.26 x 10 9
IO 18
~
I 9
18 18 CY-11 4 6/25/2008 'I 25.8 7.44 x 10 6.27 x 10 8.05 x 10 i
(a) Applicable to the longitudinal welds at 45', 135', 225', 315' in the peak axial flux.
9 n/cm -sec at 2441 MWth 2
(b) Design basis fast neutron flux = 9.66 x 10 (c) 9/30/85 is the date at which the current neutron fluences are defined.
(d) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
(e) Exposure period f rom the onset of the uprating to the license expiration date.
i 3844e:1d/121385 23
TABLE 11.2-13 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15' SURVEILLANCE CAPSULE CENTER 8eltline Region Elapsed Cumulative Fluence (n/,cm )
Irradiation Irradiation Avg. Flux Plant Desigpa)
Capsule Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data 10 18 18 18 (e.
CY-1 1.2 8.21 X 10 3.05 x 10 3.50 x 10 3.01 x 10 8
CY-2 1.9 8.73 x 10 5.04 x 10 5.65 x 10 10 18 18 CY-3 2.7 8.63 x 10 7.09 x 10 7.89 x 10 CY-4 3.8 8.67 x 10 1.01 x 10 '
1.12 x 10 '
10 I9 I9 CY-5 4.9 7.06 x 10 1.26 x 10 1.45 x 10 0
I I
CY-6 6.2 6.87 x 10 1,54 x 10 1.84 x 10 10 I9 CY-7 7.4 7.25 x 10 1.81 x 10 2.19 x 10 10 I9 I9 CY-8 (9/30/85)(b) 7.6 5.92 x 10 1.86 x 10 2.26 x 10 9/30/85 '+ CY-10(c) 10.7 5.92 x 10 2.43 x 10 3.i8 x 10" 10 I9 10 I9 CY-11'4 6/25/2008(d) 25.8 6.17 x 10 5.37 x 10 7.86 x 10 (a) Design basis fast neutron flux = 9.43 x 1010 n/cm2-sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 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) Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 7.
1 24 3844e:1d/121385
I' TABLE 11.2-14 SURRY UNIT 2
' FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25' SURVEILLANCE CAPSULE CENTER' Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Desigg Irradiation Irradiation Avg.2 Flux Plant Basis,)
Interval Time (EFPY)
(n/cm -sec)
Specific 10 18 18 CY.1.2 5.24 X 10 1.95 x 10 2.22 x 10 j
I 18-CY-2 1.9 5.62 x 10 3.23 x 10 3.58 x 10 10 18 18.
.CY-3 2.7 5.77 x 10 4.60 x 10 5.01 x 10 I
18 CY-4 3.8 5.57 x 10 6.54 x 10 7.09 x 10 10 18 18 CY-5 4.9 4.53 x 10 8.14 x 10 9.21 x 10 0
I I9 CY-6 6.2 4.59 x 10 1.00 x 10
.1.16 x 10
.CY-7 7.4 4.29 x 10 1.16 x 10 '
1.39 x 10 '
10 I
I 10 I9 I9 CY-8 (9/30/85)I I 7.6 3.98 x 10 1.19 x 10 1.43 x 10 10 I
I 9/30/85 + CY-10(c) 10.7 3.98 x 10 1.58 x 10 '
2.02 x 10 '
0 I
CY-11 + 6/25/2008( }
25.8 4.15 x 10 3.56 x 10 '
4.99 x 10 10 n/cm2-sec at 2441 MWth (a) Design basis fast neutron flux = 5.98 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
j (c) At the beginning of CY-11, the core thermal power will be uprated to 2546
{
MWth. Beyond 9/30/85 a 80% capacity factor is assumed.
1 (d) ' Exposure period f rom the onset of the uprating to the license expiration date.
J 3844e:1d/121385 25 q
TABLE 11.2-15
-l SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35' SURVEILLANCE CAPSULE CENTER 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 1.2 3.55 x 10 1.32 x 10 1.52 x 10 CY-2 1.9 3.92 x 10 2.21 x 10 2.45 x 10 10 18 18 CY-3 2.7 4.00 x 10 3.17 x 10 3.42 x 10 CY-4 3.8 3.80 x 10 4.49 x 10 4.85 x 10 10 18 18 CY-5 4.9 3.14 X 10 5.61 x 10 6.30 x 10 CY-6 6.2 3.12 x 10 6.88 x 10 7.96 x 10 10 18 18 CY-7 7.4 3.01 x 10 8.01 x 10 9.50 x 10 10 18 18 CY-8 (9/30/85)(b) 7.6 2.81 x 10 8.20 x 10 9.79 x 10 10 I9 I9 9/30/85 4 CY-10(c) 10.7 2.81 x 10 1.10 x 10 1.38 x 10 I}
10 I
I CY-11 + 6/25/2008 25.8 2.93 x 10 2.49 x 10 3.41 x 10 (a)
Design basis fast neutron fiux = 4.09 x 1010 n/cm2-sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.
)
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 J
1 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
.i 3844e:1d/121385 26
TABLE 11.2-16 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45' SURVEILLANCE CAPSULE CENTER Beltline Region Elapsed Cumulative Fluence (n/cm )
Desig Irradiation Irradiation Avg.2 Flux Plant Basis [a)
Interval Time (EFPY)
(n/cm -sec)
SDecific 10 18 18 CY-1 1.2 2.00 X 10 3 04 x 10 1.20 x 10 I
18 CY-2 1.9 3.17 x 10 1.76 x 10 1.94 x 10 10 18 I0 CY-3 2.7 3.21 x 10 2.52 x 10 2.70 x 10 10 18 18 CY-4 3.8 2.99 x 10 3.56 x 10 3.83 x 10 10 I0 18 CY-5 4.9 2.58 x 10 4.48 x 10 4.97 x 10 CY-6 6.2 2.49 x 10 5.49 x 10 6.29 x 10 10 8
18 CY-7 7.4 2.48 x 10 6.42 x 10 7.50 x 10 10 18 18 CY-8 (9/30/85)(b) 7.6 2.30 x 10 6.58 x 10 7.73 x 10 10 18 I
9/30/85 4' CY-10(c) 10.7 2.30 x 10 8.84 x 10 1.09 x 10
~
CY-11 4 6/25/2008 25.8 2.40 x 10 2.03 x 10 2.69 x 10 '
I 10 n/cm2-sec at 2441 MWth (a)
Design basis fast neutron flux = 3.23 x 10 (b) 9/30/85 is the date at which the current neutron fluences are defined.
(c) At the beginning of CY-11, the core thermal power will be uprated to 2546 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.
3844e:1d/121385 27
F 16003 19 1
1 O'
(MAJOR AXIS) i.
15' (CAPSULES V,T 4 X,V)
- ff 25' (CAPSULES S,Z,X & Y,W,U)
/
/
/
/
35' (CAPSULES Y,W & Z,T) i.
/
Q\\
45' (CAPSULE U & S)
/
V//
/
mxxx,
r PRESSURE VESSEL g
vmms i
/
/
l l
W f
/
/
THERMAL SHIELD I
\\
l
/
l I
/
/
/
/
/
/ l/ l
/
CORE BARREL
~l//
/'
BAFFLE
-l
/
//
//
-CEACTOR CORE
/ /
-4 ////
///
+ UNIT I SURV. CAPSULE I.D.
6 UNIT 2 SURV. CAPSULE I.D.
Figure 11.1-1. Surry Reactor Geometry 28 j
16003 20 iO20 7
5 3
/
/
/
CIRCUMFERENTIAL WELD, O'
f
/
/
/
/
n N
IO 19 s
C
~
7 W
,s',
N 5
W
/
b LONGITUDINAL WELDS, 45'
/
3 s'
5
/
tr
/
y
/
y th I
10 38 Z
ACTUAL 7
PROJECTED 5
3 LICENSE EXPIRATION 9/30/85 I
I I
I l
I U
10 17 O
10 20 30 40 50 60 70 OPERATING TIME (EFPY)
Figure 11.2-1. Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Weld Locations as a Function of Full Power Operating Time 29
16003 2'-
1O20 7
5 3
e'
,/
CIRCUMFERENTIAL WELD, O'
/
/
/
/
i
/
n 6
10 I9 s
C
~
7 W
N 5
,/
W
/
]
/
LONGITUDINAL WELDS, 45' f
5
,/
5
/
tr
/
/
a 1-Z H
tn 10 18 Z
ACTUAL 7
.---- PROJECTED 5
3 LICENSE EXPIRATION 9/30/85
i i
f f
l l
10 37 0
10 20 30 40 50 60 70 OPERATING TIME (EFPY)
Figure 11.2-2. Surry Unit 2 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Weld Locatioris as a Function of Full Power Operating Time 30
16003 22 h
I I
1020 I
~
ACTUAL-5
PROJECTED N
-%,%'s Eu-3
.\\
\\
C
\\N m
g OZ
\\
Lu
\\s 3
N s lO l9
s h-Z
%'s o-7
's 0
s ~ ~ _. _
tuz 5
LICENSE EXPIRATION g
h_
3 9/30/85 10 I8 I
I I
I I
I O
10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.)
I' Figure 11.2-3, Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 31
16003-23 1020 7
ACTUAL 5
PROJECTED N
Eo 3 --. s ' s N
C
% N w
W
\\s o
s z
s
's J
's 10 19 s s Z
's o
s s 7
's s a
LUz 5
LICENSE EXPIRATION g
L 3
h 9/30/85 10 18 I
I l
l I
I O
10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.)
Figure 11.2-4. Surry Unit 2 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 1
32
l : --
j'
.i 16003-34
+
.10 7
5
.3 l
.199.39 t
ii W
i b.
N l.0
-204.90 2
d' 7
)
D' CLAD
]
5 IR e
bg 3
1/4T l
wU 215.13
~
W>
s d
O. I 220.-24 l
m 1
7 3/4T l
45'
)
5 O'
3 i
REACTOR VESSEL OR t
I I
I I
I I
j O.01 195 199 203 207 211 215 219 223 i
RADIUS (cm)
Figure 11.2-5. Surry Units 1 and 2 Relative Radial Distribution of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 33 J
16CO3 25 1.0
\\
y 5
3 W
O bDy O.I N
X 7
3 u.
5 Z
0 3
Bw Z
N
[
0.01
~
b
~
7 W
5 3
CORE MIDPLANE I
I O.001
-300
-200
-100 O
100 200 300 DISTANCE FROM CORE MIDPLANE (cm)
Figure 11.2-6. Surry Units 1 and 2 Relative Axial Variation of Fast Neutron j
(E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 34
I SECTION III MATERIAL PROPERTIES For the RT calculation, the best estimate copper and nickel chemical PTS composition of the reactor vessel beltline material is necessary. The material properties for the Surry Units 1 and 2 beltline region will be presented in this section.
III.1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS The beltline region is defined by the Rule [1] to be "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 irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage."
Figures 111.1-1 and III.1-2 identify and indicate the location of all beltline
~
l region materials for the Surry Units 1 and 2 reactor vessels.
- t..
III.2 DEFINITION AND SOURCE OF MATERIAL PROPERTIES'FOR ALL VESSEL LOCATIONS Material property values for the shell plates, which have been docketed with the NRC in Reference 8, were derived from vessel fabrication test certificate results. The property data for the welds have also been docketed with the NRC in Reference 8, however, the weld properties cannot be used in the same direct j
manner as the properties for the plates.
l Fast neutron irradiation-induced changes in the tension, f racture, 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 by the variability of copper concentration within the weldments.
To address the variation in chemistry, Babcock & Wilcox (B&W) performed a reactor vessel beltline weld chemistry study of eight B&W vessels, including l
3844e:1d/121385 35
Surry Units 1 and 2, and reported the results in BAW-1799 [9] for the Westinghouse Owners Group (WOG). The scope of the work included collecting existing sources of chemistry data, performing extensive chemical analysis on the available archive reactor vessel weldments, and developing predictive methods with the aid of statistical analyses to establish the chemistry of the reactor vessel beltline weldments in question.
In addition to the B&W report BAW-1799, the WOG Reactor Vessel Beltline Region Weld Metal Data Base was used. 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, and the Materials Properties Council (MPC) and the NRC Mender MATSURV data bases.
i For each of the welds in the Surry Units 1 and 2 beltline region, a material data search was performed using the WOG data base.
Searches were performed for materials having the identical weld wire heat number as those in the Surry vessels, but any combination of wire and flux was allowvd.
For all of the
_j data found for a particular wire, the copper, nickel, phosphorous and silicon values were averaged and the standard deviations were calculated. Although
,1 phosphorous and silicon are not needed for the PTS Rule, they are provided for the sake of completeness. The information obtained from the data base searches is found in Appendix B.
III.3
SUMMARY
OF PLANT-SPECIFIC MATERIAL PROPERTIES A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Surry Units 1 and 2 reactor vessels are
{
respectively given in Tables 111.3-1 and 111.3-2 along with the ref erences for this information.
Although phosphorus is no longer used in the calculation of RT with respect to the PTS rule [1], it is given for reference since it 4
NDT is currently used in the Regulatory Guide 1.99, Revision 1 trend curve [10].
The initial RT value of 0*F, which is shown for all of the Surry Units 1 NDT
~
and 2 reactor vessel beltline weldments made with Linde 80 flux, is the 3844e:1d/121385 36
generic mean value defined in the PTS rule [1] for welds made with Linde 80 flux. Grau Lo flux was used in manufacturing the circumferential weld in Surry Unit 2.
The initial RT value of 0*F was estimated per the NRC NDT Standard Review Plan [11).
The data in Tables III.3-1 and.111.3-2 are used to evaluate the RT values PTS for the Surry Unit 1 and 2 reactor vessels.
i l
I 3844e:1d/121385 37 3
TABLE 111.3-1 SURRY UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P
I (Wt.%)
(Wt.%) -(Wt.%)
(*F)
Source Intermediate Shell Plate C4326-1:
0.11 0.55 0.008 10(a)
Ref. [8]
Intermediate Shell Plate C4326-2: 0.11 0.55 0.008 0(a)
Ref. [8]
Lower Shell Plate C4415-1:
0.11 0.50 0.014 20(a)
Ref. [8]
Lower Shell Plate C4415-2:
0.11 0.50 0.014 0(a)
Ref. [8]
Longitudinal Welds - Intermed. & Lower Shells L1/L4 Heat No. 8T1554, Linde 80 Flux 8579:
0.18 0.63 0.014 0(b)
WOG Material Data Base, BAW 1799 [9]
1 Circumferential Weld - Intermed, to Lower Shell WOS, Heat No. 72445, Linde 80 Flux 8579/8632:
0.21 0.58 0.016 0(b)
WOG Material Data Base Notes:
(a) The initial RTHDT value for this plate is estimated according to Branch Position MTEB 5-2 [11]
(b)
The initial RTNDT values for the welds are the generic mean value defined by the PTS rule [1] for welds with Linde 80 flux.
3844e:1d/121385 38
TABLE 111.3-2 SURRY UNIT 2 REACTOR VESSEL BELTLINE REGION MATCRIAL PROPERTIES A
Cu Ni P
I
( Wt '. %) (Wt.%)
(Wt.%).(*F)
Source j
Intermediate Shell Plate C4339-1:
0.11 0.54 0.012 30(a)
Ref.-[8]
l-Intermediate Shell Plate C4208-2: 0.15 0.55 0.008
-30(a)
Ref. [8]
Lower Shell Plate C4331-2:
0.12 0.60 0.009 10(a)
Ref
[8]
1 Lower Shell Plate C4339-2:
0.11 0.54 0.012 10(a)
Ref. [8]
Longitudinal Welds - Intermed. Shell L4 and L3, Heat No. 72445, Flux 8597:
0.21 0.58 0.016 0(b)
WOG Material Data Base Longitudinal Welds - Lower shell L1 and L2, Heat No. 8T1762, Flux 8597/8632:
0.29 0.55 0.013 0(b) wog Material Data Base, BAW 1799 [9]
Circumferential Weld - Intermed to Lower Shell WOS, Heat No. 0227, Grau Lo LW320:
0.19 0.56 0.017 0(a) wog Material Data Base Notes:
(a) The initial RTNOT value for these plates and weld are estimated according to Branch Position MTEB 5-2 [11]
(b) The initial RTNDT values for the welds are the generic mean value
)
defined by the PTS rule [1] for welds with Linde 80 flux.
4 i
3844e:1d/121385 39
~
FIGURE III.1 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE SURRY UNIT 1 REACTOR VESSEL CIRCUMrERENTI AL SE AMS VERTICAL SEAMS.
270*
L4 9
C4326-1 2 - WO6 t
.9. 0 " '
I 45 CORE
\\;
0*
180' CORE C
5 s
=
C4326-2 L3 90*
144" 33 CL
_JL__
19.7" 14
-WO5 o
270' L2 C4415-1 15' 7,
O' 180*-
48.3" f
C4415-2 L1 90*
40 -
FIGURE III.1-2 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL l
FOR THE SURRY UNIT 2 REACTOR VESSEL w
CIRCUMFERENTI AL SEAMS VERTICAL SEAMS, 270' L4
,ej C4339-1 1
- WO6 9.0" I
45 CORE j
180' O'
'~'
CORE I
C4208-2 L3 144" 3
90*
E CL n-19.7" I4
-WOS o
270' C-4331-2 L2 l
15' NE e
180*-
0' o -
.f 48.3" C4339-2
~
L1 90'
~
i 41
SECTION IV DETERMINATION OF RT VALUES FOR ALL BELTLINE REGION MATERIALS PTS Using the methodology prescribed in Section I.2, the results of the fast neutron exposure provided in Section II, and the material properties discussed in Section III, the RT values for Surry Units 1 and 2 can now be PTS determined.
IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT PTS RESULTS values were generated for Using the prescribed PTS Rule methodology, RTPTS all beltline region materials of the Surry Units 1 and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes. The tabulated results from the total evaluation are presented in Appendix C for all beltline region materials for both units.
values for n e H m W ng Figures IV.1-1 and IV.1-2 present the RTPTS longitudinal weld, circumferential weld and shell plate of the Surry Units 1 and 2 vessels in terms of RT versus fluence
- curves. The curves in these PTS figures can be used:
to provide guidelines to evaluate fuel reload options in relation to the o
values can be NRC RT Screening CrheHon for MS (i.e., RTPTS PTS readily projected for any options under consideration, provided fluence is known), and to show the current (7.4 EFPY for Surry 1 and 7.6 EFPY for Surry 2), and o
end-of-license (25.6 EFPY for Surry 1 and 25.8 EFPY for Surry 2) RTPTS values using actual and projected fluence.
- The EFPY can be determined using Figure 11.2-1 for Unit 1 and Figure 11.2-2 for Unit 2.
3844e:1d/121385 42
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.
l IV.2 DISCUSSION OF RESULTS As shown in Figures IV.1-1 ar.d IV.1-2, the welds are the governing locations for both reactor vessels relative to PTS. All the RT values remain below PTS the NRC screening values for PTS using the projected fluence values.through the license expiration. The;most limiting RT va ue at-license expiration PTS is 244*F for the circumferential weld of Unit 1 and 220*F for the longitudinal welds in the lower shell of Unit 2.
)
i 1
3844e:1d/121385 43
TABLE IV.1-1 RT VALUES FOR SURRY UNIT 1 PTS 6
- "'8'I
)
PTS Present End-of-License Screening location-Vessel Material (7.4 EFPY)
(25.6 EFP1)
Criteria 2
Intermediate shell plate C4326-1 123 146 270 I
l; 1
Intermediate shell plate C4326-2 113 136 270 3
Lower shell plate C4415-1 131 154 270 4
Lower shell plate C4415-2 111 134 270
~
l' 5
Intermediate shell longitudinal 131 157 270 welds L1, L2, L3, L4 6
Intermediate to lower shell 195 244 300 circumferential weld WOS l
l 1
3844e:1d/121385 44 t=
Dl m -
tr,
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~ q' k,,
si
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f s
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-. y.
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.,f:6 a
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(
i v
f R,
T,. VALUES FOR SURRY UNIT 2 v
,Vy\\
11.
>\\
)
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f t
.'t RT Vg. fos.
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PTS 7
's Present End-of-License. Screening Location Vassel Material (7.4 EFPY)
(25.6EFPb Rf,teria s.
)
v
,?
1 Intermediate shell plate C4339-1 142 162 270 4
4 h
2 Intermediate shell plate C4208-2 110 139 270 l\\
t-3 Lower shell plate C4331-2 132 155 270' s
\\
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i t' 4 Lower shell plate C4339-2 122 142 270 I
A. '-
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4 welds L4,L3
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f; W
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L1 L2 tJ-N f'
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SECTION.V CONCLUSIONS AND RECOMMENDATIONS S
Calculations have been completed in order to submit RT values for the PTS Surry 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 Surry Units 1 and 2 pressure vessels.
Explicit calculations were performed for the first eight operating cycles of both units as of September 30, 1985.
For both units, projection of the fast neutron exposure beyond September 30, 1985 was based on continued implementation of low leakage fuel management similar to that employed during cycle 8 of each a
unit.
e l
In regard to the low leakage fuel management already in place at the Surry Units, the plant specific evaluations have demonstrated that for the cycle 8 case the peak fast neutron flux at the O' azimuthal position has been reduced by a factor of 1.47 in Unit 1 and a f actor of 1.85 in Unit 2 relative to the i
flux based on the design basis core power distribution.
l l
l This location represents the maximum fast neutron flux incident on the reactor i
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 l
presented above.
It should be noted that significant deviations f rom the low leakage scheme l-already in place will affect the exposure projections beyond the current
{
i 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 3844e:1d/121385 48 i
l i
on the core periphery would increase the projections beyond those reported.
As each future fuel cycle evolves, the loading patterns should be analyzed to y
determine their potential impact on vessel and capsule exposure.
a1 The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence levels derived from neutron dosimetry l
contained in surveillance capsules withdrawn from each of the Surry Units.
For Unit 1, the ratio of calculated to measured fluence values ranges f rom 0.92 to 0.97 for the two capsule data points. The corresponding ratio for Unit 2 is 1.01 for the capsule removed from that reactor. This excellent agreement between calculation and measurement supports the use of this analytical approach to perform a plant specific evaluations for the Surry reactors.
I Material property vC.ues for the Surry Units 1 and 2 reactor vessel beltline region components were determined.
The pertinent chemical and mechanical properties for the shell plates remain the same as those that have been docketed with the NRC in Reference 8.
The weld material properties are obtained f rom the WOG Material Data Base.
Using the prescribed PTS Rule methodology, RT va es w e gen nated for PTS all beltline region materials of the Surry 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 values PTS for PTS using the projected fluence exposure through the expiration date of the operating license. The most limiting values at end-of-license (25.6 EFPY y
l for Surry Unit 1 and 25.8 EFPY for Surry Unit 2) are 244*F and 220*F for the circumferential weld for Unit 1 and the longitudinal welds in the lower shell of Unit 2, respectively.
l i
The results in this report are provided to enable Virginia Power Company to l
comply with the initial 6 months submittal requirements of the USNRC PTS Rule.
-i e
3844e:1d/121385 49
SECTION VI REFERENCES o
1.
Nuclear Regulatory Commission,10CFR Part 50, " Analysis of Potential l
l 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, Cross-Section Library for Light Water 3
Reactors.
4.
Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -
~
to be published.
5.
"Surry Unit No.1 Pressure Vessel Irradiation Capsule Program:
Examination and Analysis of Capsule T,"
J. S. Perrin, et al., June 24, 1975.
6.
"Surry Unit No. 1 Pressure Vessel Irradiation Capsule Program:
Examination and Analysis of Capsule W," J. S. Perrin, et al., March 30, 1979.
7.
"Surry Unit No. 2 Pressure Vessel Irradiation Capsule Program:
Examination and Analysis of Capsule X,"
J. S. Perrin, et al., September 2, 1975.
8.
Letter f rom C. M. Stallings of Virginia Power to E. G. Case of the NRC, Serial No. 081 February 15, 1978.
3844e:1d/121385 50
1 9.
B&W Owners Group Report, BAW-1799, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study", July 1983.
- 10. " Effects of Residual Elements on Predicted Radiation Damage to Reactor
,i Vessel Materials," Regulatory Guide 1.99 - Revision 1 U.S. Nuclear j
Regulatory Commission, Washington, April 1977.
- 11. NUREG-0300 - U.S. NRC Standard Review Plan, Branch Technical Position 5-2, Revision ', July 1981.
- 12. Letter f rom K. L. Basehore of Virginia Power to D. R. Beynon, Jr. of Westinghouse Electric Corporation transmitting measured fuel assembly and cycle burnops for the Surry and North Anna Units, dated October 7, 1985.
1
. 1 l
e
' l l
1 3844e:1d/121385 51
APPENDIX A i.
POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the Surry pressure vessels were derived from the measured fuel i
assembly and cycle burnup data supplied by Virginia Power [12]. 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 first 7 cycles of operation.
In addition, estimated data was provided for the current cycle of operation (Cycle 8).
Table A-1 shows the Surry Unit 1 fuel assembly and cycle burnups for Cycles 1 through 8.
Similar data for Surry Unit 2 are shown in Table A-2.
(The fuel assembly locations in the Surry cores are numbered according to Figure A-1).
Cycle-averaged relative assembly powers for each cycle were computed using the following relation OC Assembly Burnup - BOC Assembly Burnup i
Relative Assembly Power =
Cycle Burnup and are shown in Tables A-3 and A-4 for Surry 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 distribution data for these interior assemblies are not given in Tables A-3 and A-4.
In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels.
For peripheral 3844e:ld/121385 A-1
1 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.
4 1
l l
3844e:1d/121385 A-2
TABLE A-1 o
l SURRY UNIT 1 BEGINNING-0F-CYCLE AND END-OF-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assembly BurnuD (MWD /MTU)
Fuel Cvele(a) 1(13547) 2(6915) 3(8944) 4(13107)
Assembiv Eqq EOC 82q EOC 80C E0C BOC EOC
'l 0
10615 0
6365 6400 13380 0
9825 2-0 8425 0
5075 5075 10763 0
7905 3
0 12275 0
7305 6450 14473 0
11825 4
0 8623 0
5438 6050 11730 0
8328 5
0 9285 0
6070 5450 11668 0
8928 6
0 12390 14550 20175 8650 18655 8300 20720
'7.
0 13900 0
6453 7300 16923 0
13320 8
0 13753 14475 21513 13725 21370 14475 28715 9
0
'14180 14250 20000 8425
- 18570, 0
14088 10 0
13328 0
7415 7400 17093 0
13373 11 0
14468 16275 23005 7625 16175 11725 25593 12 0
11555 0
6883 8625 17920 13425 25895 (a) The number in parenthesis beside the cycle number is the fuel cycle length in MWD /MTU.
3844e:1d/121385 A-3
L TA8LE A-1 (Continued)-
SURRY UNIT 1 l~
SEGINNING-OF-CYCLE AND END-OF-CYCLE FUEL ASSEM8LY BURNUPS Fuel Assemb1v Burnuo (ptdD/MTU)
I Fuel Cycle "
5(14390) 6(16491) 7(11984) ~
8(14500)
Assembiv SQE
.M RQE-M' M
M.
gQg M
'l 0
10730 0
10965 0
11125 8794 18233 2
14088 20363' 16552 23188 0
8883 26041 31676 3
0 13685 0
14238 20034 28135 13824 24780 4
L15570 22105 25593 32388 25995 30730 28496 34121 5-15128 22115 17374 25063 25701 30363
.22530 28747 6
7965!
22515 10958 26340' 25981 37125 11123' 27319
-7' O
15335 0
16045
.0 13693 0
16178 8
8330 24743 15336 33000 19146 32393 30220 44443 0
16553 0
18945 0
14253 0
17569 9
'10 0'
15450-0 16473
-0 12925 0
15824 11 0
'17623 0'
19590 0
15205 0
18023-12 7853 21533
.10725 25888 28995 37585 15271-28592 i
)
4
/
~
(a) The number in parenthesis beside the cycle number is the fuel cycle length in MWO/MTV.
1 3844e:1d/121385 A-4 l
TABLE A-2 SURRY UNIT 2 j{, GINNING-0F-CYCLE AND END-OF-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assemb1v Burnuo (MWD /MTU)
Fuel Cvele ")
I 1(14870)-
2(9054) 3(9422) 4(13678)
Assembiv
0 11390 0
7095 7095 13775 0
10955 2
0 9145 0
5938 0
6140 0
8943 3
0 13423 0
8678 6215 14755 0
12663 4
0 9445 0
6215 0
6853 0
9385 5
0 10178 0
7095 0
7425 0
9995 6
0 13515 16950 24860 13515 20985 17450 29975 7
0 15183 0
10365 5938 15980 0
14290 8
0 15058 14775 24468 10878 20975 14750 28998 9
0 15593 17350 25908 9145 19815 0
15420 10 0
14650 0
10040 7095 17413 0
14118 11 0
15845 15850 25125 10178 20620 0
15698 12 0
12608 0
10090 13505 23145 17425 29873 (a) The number in parenthesis beside the cycle number is the fuel cycle length in MWD /MTU.
3844e:1d/121385 A-5
TABLE A-2 (Continued)
SURRY UNIT 2 BEGINNING-OF-CYCLE AND END-0F-CYCLE FUEL ASSEMBLY 80RNUPS Fuel Assembiv Burnup (MWD /MTU)
Fuel Cvele(a) 5(13971) 6(16006) 7(14802) 8(13000)
Assembiv BOC
0 10250 0
11210 0
11405 32167 37372 2
16268 22393 17672 24465 16811 24225 26980 31662 3
0 13125 0
14890 14893 25975 18204 27394 4
22130 28423 16405 24128 25586 31918 25974 31173 5
9388 17700 16812 25588 18846 26980 24223 30606 6
8880 23265 15202 30325 15639 30560 11406 25115 7
0 14575 0
15648 0
15320 0
13554 8
14118 29685 16876 34278 19868 36033 15323 31390 9
0 16813 0
18845 0
17788 0
15761 10 0
15045 0
16810 0
lbs;3 0
13606 11 0
16408 0
19508 0
18208 0
16497 12 15403 28278 10264 25745 19149 32600 15975 28199 (a) The number in parenthesis beside the cycle number is the fuel cycle length in NWD/MTU.
3844e:1d/121385 A-6
TABLE A-3 o
'SURRY UNIT 1 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS-Plant SDecific Cycle Averaged Relative Assembiv Power Design Fuel Cycle Basis Assembiv-Relative Power 1
2 3
4 1
1.00 0.78 0.92 0.78 0.75 2
0.83 0.62
.0.73 0.64
-0.60 3:
1.21-0.91 1.06 0.90 0.90 4-0.86 0.64 0.79 0.64 0.64 5
0.92 0.69 0.88 0.70 0.68 6
0.98 0.91 0.81 1.12 0.95, 7
1.10 1.03 0.93 1.08 1.02 8-1.00 1.02 1.02 0.85 1.09 9
1.05.
1.05 0.83 1.13 1.07 10 1.08-0.98 1.07 1.08 1.02 11 1.06 1.07 0.97 0.96 1.06 12 0.95 0.85 1.00 1.04 0.95 I
3 i
l l
l 1
1 J
t 3844e:1d/121385 A-7
TA8LE A-3 (Continued).
SURRY UNIT 1 J
CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS
-)
Plant SDecific Cycle Averaged Relative Assembiv Power Design Fuel Cycle Basis Assembiv Relative Power 5
6 7
8 1
1.00 0.75 0.66 0.93 0.65 j
2 0.83 0.44 0.40 0.74 0.39, 3
1.21 0.95 0.86 0.68 0.76 4
0.86 0.45 0.41 0.40 0.39 5
0.92 0.49 0.47 0.39 0.43 6
0.98 1.01 0.93 0.93 1.12 7
1.10 1.07 0.97 1.14 1.12
^
8 1.00 1.14 1.07 1.10 0.98 9
1.05 1.15 1.15 1.19 1.21 10 1.08 1.07 1.00 1.08 1.09 11 1.06 1.22 1.19 1.27 1.24 12 0.95 0.95 0.92 0.72 0.92
)
I 3844e:1d/121385 A-8
TABLE A-4 SURRY 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 1
1.00 0.77 0.78 0.71 0.80 2
0.83 0.61 0.66 0.65 0.65 3
1.21 0.90 0.96 0.91 0.93 4
0.86 0.64 0.69 0.73 0.69 5
0.92 0.68 0.78 0.79 0.73 6
0.98 0.91 0.87 0.79 0.92 7
1.10 1.02 1.14 1.07 1.05 8
1.00 1.01 1.07 1.07 1.04 9
1.05 1.05 0.95 1.13 1.13 10 1.08 0.99 1.11 1.10 1.03 11 1.06 1.07 1.02 1.11 1.15 12 0.95 0.85 1.11 1.02 0.91 3844e:1d/121385 A-9
pn a=
TABLE A-4 (Continued) l SURRY UNIT 2 CORE POWER DISTRIBUTIONS USED IN THE-FLUENCE ANALYSIS Plant Specific Cycle Averaaed Relative Assemb1v Power Design Fuel Cycle-Basis Assembiv Relative Power 5
6 7
8 1
1.00 0.73 0.70 0.77, 0.40 2
0.83 0.44 0.42 0.50 0.36 3
1.21 0.94 0.93 0.75 0.71
'4 0.86 0.45 0.48 0.43 0.40 5
0.92 0.60 0.55 0.55 0.49 6
0.98 1.03 0.95 1.01 1.05 7
1.10 1.04 0.98 1.03 1.04 8
1.00 1.11 1.09 1.09 1.24 5
9 1.05 1.20 1.18 1.20 1.21 10 1.08 1.08 1.05 1.08 1.05 11 1.06 1.17 1.22 1.23 1.27 12 0.95 0.92 0.97 0.91 0.94 I
i 3844e:1d/121385 A-10 1
16C03-26 O'
(MAJOR AXIS)
BAFFLE 7
///
~
CORE BARREL x N N 'N I
2
\\
45 x\\NNy p
6 7
3 4
\\
sN\\
p/
8 9
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1I 12 I
Figure A-1. Surry Units 1 & 2 Core Description for Power Distribution Map A-11
APPENDIX B WELD CHEMISTRY Tables B.1-1 through B.1-4 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 Surry Units 1 and 2 reactor vessels beltline region. The 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:
AN1 Arkansas Nuclear 1 BAW-1799 Babcock & Wilcox Report Number Babcock & Wilcox B&W CE Combustion Engineering 8'
CR3 Crystal River 3 Weight % of Copper Cu-CWE Zion 1 Emission Spectrographic Analysis ESA MATSURV NRC Mender MATSURV Data Base MPC Materials Properties Council Data Base Ni Weight % of Nickel 001 Oconee 1 l
P Weight % of Phosphorous SC Surveillance Capsule Si Weight % of Silicon TMI Three Mile Island 1 VPA Surry 1 VIR Surry 2 P
WEP Point Beach 1 WQ Weld Qualification i
l l
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B-S
APPENDIX C 1
l e
RT VALUES OF SURRY UNITS 1 AND 2 PTS REACTOR VESSEL BELTLINE REGION MATERIALS C.1 SURR( UNIT 1 values, as a function of both Tables C.1-1 through C.1-5 provide the RTPTS constant fluence and constant EFPY (assuming the projected fluences values),
for all beltline region materials of the Surry Unit 1 reactor vessel. The RT values are calculated in accordance with the PTS 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 plate C4326-1 2
Intermediate shell plate C4326-2 3
Lower shell plate C4415-1 4
Lower shell plate C4415-2 5
Intermediate and lower shells longitudinal welds L1, L2, L3, L4 6
Intermediate to lower shell circumferential weld WO5 C.2 SURRY 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 Surry Unit 2 reactor vessel.
The
-(
RT values are calculated in accordance with the PTS rule, which is PTS Reference [1] in the main body of this report. The vessel location numbers i
3844e:1d/121385 C-1
)
in.the following tables correspond to the vessel materials identified below.
1 and in Table 111.3-2 of the main report.
location Vessel Material 4
1 Intermediate shell plate C4339-1 2
Intermediate shell plate C4208-2 3
Lower shell plate C4331-2 4
Lower shell plate C4339-2 5
Intermediate shell longitudinal welds L4,L3 l
6 Lower shell longitudinal welds L2,L1 7
Intermediate to lower shell circumferential weld WOS I
i s
3844e:1d/121385 C-2
ECNEUL' F
9 SL-A I
RETAM NO I
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s REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR: 8601290115 DOC. DATE: 86/01/23 NOTARIZED: NO DOCKET O FACIL: 50-OOO Generic Dock et 05000000 50-280 Surry Power Station, Unit 1.
Virginia Electric & Powe 05000280
(
50-281 Surry Power Station, Unit 2, Virginia Electric & Powe 05000281 50-338 North Anna Power Station, Unit 1, Virginia Electric &
05000338 50-339 North Anna Power Station, Unit 2, Virginia Electric &
05000339 AUTH.NAME AUTHOR AFFILIATION STEWART,W.L.
Virginia Power (Virginia Electric & Power Co.)
RECIP.NAME RECIPIENT AFFILIATION DENTON,H.R.
Office of Nuclear Reactor Regulation, Director (post 851125 RUBENSTEIN,L.S.
PWR Project Directorate 2
SUBJECT:
Forwards WCAP-11015 & Rev 1 to WCAP-11016 reflecting values of current & projected ref temps for pressurized thermal shock (PTS). Projected values do not exceed PTS screen n
criteria.W/o WCAP-11015.
DISTRIBUTION CODE: A049D COPIES RECEIVED: LTR __ ENCL __,
ZE:
TITLE: OR Submittal: Thermal Shock to Reactor Vessel W NOTES; icy NMSS/FCAF/PM.
05000280 OL: 05/25/72 1ey NMSS/FCAF/PM.
05000281 OL: 01/29/73 LPDR 2cys.
05000338 OL: 11/26/77 LPDR 2cys.
05000339 OL: 11/04/80 RECIPIENT COPIES RECIPIENT COPIES ID CODE /NAME LTTR ENCL ID CODE /NAME LTTR ENCL PWR-A ADTS 1
1 PWR-A PD2 PD 01 5
5 CRAN,T 1
1 ENGLE,L 1
1 INTERNAL: ACRS 10 6
6 ADM/LFMB 1
O ELD /HDS4 12 1
O ELD /HDS2 12 1
O NRR LOIS,L 1
1 NRR MILLER,C 1
1 NRR VISSING,GO4 1
1 REC FILE 05 1
1 RES RANDALL,P 1
1 RES/DET 1
1 RGN2 1
1 RGN1 ADMSTR 1
1 EXTERNAL: 24X 1
1 LPDR 03 4
4 NRC PDR O2 1
1 NSIC 06 1
1 l
l NOTES:
3 3
Nd,'
Alb-J. k, night (l+r. onlyh EB (Scdic ed)
GlCSB (ROSA)
Pss (onmmm)
(. Berlin 9er) l hB MB g(Bene-cy%) q l
TOTAL NUMBER OF COPIES REGUIRED: LTTR ENCL