ML20072F266
| ML20072F266 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/30/1994 |
| From: | Shaun Anderson, Lau F, Lippincott E CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20072F244 | List: |
| References | |
| SE-REA-074-94, SE-REA-074-94-S01, SE-REA-74-94, SE-REA-74-94-S1, NUDOCS 9408230267 | |
| Download: ML20072F266 (26) | |
Text
.
(h 1
P g
9s tJ NEUTRON DOSIMETRY RESULTS FOR THE SURVEILLANCE CAPSULE REMOVED AT THE CONCLUSION OF FUEL CYCLE 8 (25 PAGES) e l
j 9408230267 940817 PDR ADOCK 05000324 P
SE REA-074/94 NEUTRON DOSIMETRY RESULTS FOR THE SURVEILLANCE CAPSULE REMOVED AT THE CONCLUSION OF FUEL CYCLE 8 June 1994 Prepared By:
[. d.
S.
L.
Anderson, Engineer Radiation Engineering and Analysis 0r
/
^
Verified By:
s E.
P.
Lippincott, Engineer Radiation Engineering and Analysis Approved By:
b
/2/ L_
F.
L.
blIi, Man Radiation Engineering and Analysis 1
SE-REA-074/94 e
1.0 INTRODUCTION
l Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons.
First, in order to interpret the radiation induced material property changes observed in the test specimens, the neutron environment (energ/ spectrum, flux, fluence) to which the test specimens were exposed must be known.
Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the pressure vessel and that experienced by the test specimens.
The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
The latter information is generally derived solely from analysis.
The use of fast neutron fluence, c (E > 1.0 MeV), to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for development of damage trend curves as well as f or the implementation of trend curve data to assess vessel condition.
In recent years, however, it has been suggested that an exposure model that accounts for differences in spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from the threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, '" Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with 4 (E > 1.0 MeV) to i
l provide a data base for future reference.
The energy dependent 2
SE-REA-074/94 dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Charactericing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom."
The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99,
" Radiation Damage to Reactor Vessel Materials."
This report provides the results of the neutron dosimetry evaluations perf ormed f or the surveillance capsule withdrawn at the end of the eighth fuel cycle at Brunswick Unit 1.
In this dosimetry evaluation, fast neutron exposure parameters in terms of 4 (E > 1.0 MeV), C (E > 0.1 MeV), and dpa are established for the capsule irradiation history.
The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and is then used to project the integrated exposure of the vessel wall.
Neutron exposure projections are provide for 8.7 EFPY (End of Cycle 8) as well as for 16 and 32 EFPY.
1 i
l 3
4
y SE-REA-074/94 2.0 DISCRETE ORDINATES ANALYSIS A plan view of the Brunswick Unit i reactor geometry at the core midplane is shown in Figure 2-1.
Since the reactor exhibits 1/8th core symmetry, only a 0 45 degree sector is depicted.
In addition to the core, reactor internals, pressure vessel, and sacrificial shield, the geometry also included representations of the jet pumps located internal to the the pressure vessel.
Development of the geometric model shown in Figure 2-1 made use of nominal design dimensions throughout.
The jet pumps were modelled as homogeneous zones characteristic of the pump geometry opposite the axial midplane of the reactor core.
A plan view of the surveillance capsule attached to the pressure vessel wall is shown in Figure 2-2.
From a neutronic standpoint, the surveillance capsule itself is a significant structure.
The presence of this material has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the shroud and the pressure vessel.
In order to accurately determine the neutron environment at the dosimetry location, the capsule itself was also included in the analytical model.
In performing the fast neutron evalur.tions for the surveillance capsules and reactor vessel geometry a series of three transport calculations were carried out.
The first, a two-dimensional R,0 calculation was used to provide distributions of neutron flux throughout the geometry depicted in Figures 2-1 and 2-2.
The additional computations, a two-dimensional analysis in R,Z geometry and a one-dimensional calculation in R geometry, were used to synthesize a three-dimensional distribution of the neutron environment from the following relationship:
L, ( R, Z )
(Equation 1)
$,(R,0,Z)
$(R,0)
=
where $,(R,0) is the group-g transport solution'in R,0 geometry and L, is a group-dependent axial shape f actor determined f rom the following equation.
{ $, (R, Z) ] / ($, (R) ]
(Equation 2)
L, ( R, Z )
=
4
SE REA-074/94 where 4,(R, Z) and 4,(R) are the group-g flux solutions from the R,2 and R transport calculations, respectively.
The use of this synthesis approach allows the axial variation in void fraction characteristic df BWR cores to be accounted for in the overall analysis.
All of the transport calculations for Brunswick Unit 1 were carried out using the DORT discrete ordinates code" and the l
SAILOR cross-section libraryCl.
The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications.
In the analyses, anisotropic scattering was treated with a P expansion of the cross-sections and the 3
angular discreti:ation was modelled with an Se order of angular quadrature.
Core power distributions including fuel assembly burnups and relative axial distributions of both fuel burnup and core void DI for fraction; were provided by Carolina Power and Light Company each of the eight fuel cycles utilized at Brunswick Unit 1.
For the current evaluations, these data sets were processed to provide a burnup weighted average of power distribution and core void fraction for use in the transport calculations.
Selected results from the neutron transport analyses are provided in Tables 2-1 and 2-2.
In Table 2 1, the calculated exposure parameters (&(E > 1.0 MeV), $(E > 0.1 MeV), and dpal are given at the geometric center of the surveillance capsule as well as at several azimuthal locations along the inner radius of the pressure vessel.
The tabulated data for both the capsule and vessel are representative of an axial elevation corresponding to the reactor core midplane.
Radial gradient information applicable to 4(E > 1.0 MeV), $(E >
0.1 MeV), and dpa/sec is given in Table 2-2.
The gradient data are presented on a relative basis for each exposure parameter at several azimuthal locations.
Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Table 2-2.
5 l
l
'1 1
SE-REA 074/94 4
Axial gradient information applicable to all three exposure parameters is provided in Table 2 4.
Again, the gradient data j
are presented on a relative basis indexed to the midplane of the reactor core; and may be used in a multiplicative fashion to translate the midplane calculations to any desired axial elevation, j
i i
I i
i 1
i i
i l
6 l
e
a L
4 A
-mB
-ar.&
a.2.aA%
w es a
a SE-REA-074/94 FIGURE 2-1 BRUNSWICK UNIT 1 R,O REACTOR GEOMETRY e
- e. g de a.
t#'
Insulation Pressure Yessel I--
v Jet Pumps e,s Shroud f'
i 3
A t t.. \\ %'
}
4.<a
=
et i
'hi O
J Core
/
j s
f,
~
\\
s o.
f%
4 l
/,.
8 50 100 150 200 250 350
- 400
~
1 Radius (cm) v 3<
iHs..f." -
7 l
t=r%g=pt;;_.
- &&lwn+s
' .Q: ;';'G1:,,l i
d,k,,w &, h ~.
J
- - - - - - - - - - - - - - - - - - ^ ^
SE-REA 074/94 i
FIGURE 2-2 SURVEILLANCE CAPSULE GEOMETRY 4--
108.75 In. to VESSEL CENTER e 0.88 in, f/
k l
/
/
i i
/ N
!+
0.7s in.
vssset O !."
ressitE cAesula WALL
/
./
l~ nux..
/
[/
CHARPY CAPSULE
/
\\
/
\\
/
/
/
~
LOWER BRACKET 8
I
i SE REA-074/94 TABLE 2-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER AND PRESSURE VESSEL MIDPLANE INNER RADIUS
$(E > 1.0 MeV) 4(E > 0.1 MeV)
LOCATION In/cm -sec)
In/cm -sec) doa/sec CAPSULE CENTER 1.18e+09 1.97e+09 1.81e-12 VESSEL IR 0 Degrees 7.46e+08 1.28e+09 1.19e-12 15 Degrees 9.75e+08 1.67e+09 1.55e-12 30 Degrees 9.00e+08 1.68e+09 1.43e-12 45 Degrees 1.47e+09 2.54e+09 2.33e-12 VESSEL 1/4T 0 Degrees 5.58e+08 1.25e+09 9.16e-13 15 Degrees 7.28e+08 1.59e+09 1.18e-12 30 Degrees 6.60e+08 1.65e+09 1.10e-12 45 Degrees 1.09e+09 2.41e+09 1.75e-12 VESSEL 3/4T 0 Degrees 2.54e+08 8.96e+08 4.76e-13 15 Decrees 3.02e+08 1.00e+09 5.43e-13 30 Degrees 2.97e+08 1.13e+09 4.86e-13 45 Degrees 4.56e+08 1.47e+09 8.16e-13 i
a l
9 i
P
SE-REA-074/94 TABLE 2-2 RELATIVE RADIAL DISTRIBUTION OF EXPOSURE PARAMETERS
'WITHIN THE PRESSURE VESSEL WALL c(E > 1.0 MeV)
&(E > 0.1 MeV)
LOCATION (n /cm: sec)
( n / cnd - s e c )
doa/sec CAPSULE CENTER VESSEL IR 0 Degrees 1.00 1.00 1.00 15 Degrees 1.00 1.00 1.00 30 Degrees 1.00 1.00 1.00 45 Degrees 1.00 1.00 1.00 VESSEL 1/4T 0 Degrees 0.76 0.98 0.77 15 Degrees 0.74 0.95 0.76 30 Degrees 0.75 0.98 0.77 45 Degrees 0.74 0.95 0.75 VESSEL 3/4T 0 Degrees 0.34 0.70 0.40 15 Degrees 0.31 0.60 0.35 30 Degrees 0.33 0.67 0.34 45 Degrees 0.31 0.58 0.35 10
SE-REA-074/94 TABLE 2-3 RELATIVE AXIAL DISTRIBUTION OF EXPOSURE PARAMETERS WITHIN THE PRESSURE VESSEL WALL HEIGHT (ft' F(2)
+5.5 0.54 I
+4.5 0.73 1
+3.5 1.00
+2.5 1.03
+1.5 1.04
+0.5 1.04 1
0.0 1.00
-0.5 0.96
-1.5 0.87 l
-2.5 0.80
-3.5 0.67
-4.5 0.55
-5.5 0.39 Note:
Z= 0.0 is referenced to the midplane of the reactor core.
i l
l 11
SE-REA-074/94 TABLE 2-4 CALCULATED NEUTRON SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER ENERGY NEUTRON FLUX ENERGY NEUTRON FLUX 2
(MeV)
(n/cm'-sec)
(MeV)
(n/cm -sec) 1.73e+01 1.93e+06 2.97e-01 1.38e+08 1.42e+01 6.70e+06 1.83e-01 1.18e+08 1.22e+01 2.85e+07 1.11e-01 9.19e+07 1.00e+01 2.99e+07 6.74e-02 7.93e+07 8.61e+00 4.33e+07 4.09e-02 3.11e+07 7.41e+00 9.12e+07 3.18e-02 1.42e+C7 6.07e+00 1.06e+08 2.61e-02 2.63e+07 4.97e+00 1.45e+08 2.42e-02 1.51e+07 3.68e+00 8.94e+07 2.19e-02 4.90e+07
- 3. Ole +00 5.83e+07 1.50e-02 8.88e+07 2.73e+00 6.19e+07 7.10e-03 9.04e+07 2.47e+00 2.90e+07 3.36e-03 8.44e+07 2.37e+00 7.24e+06 1.59e-03 1.39e+08 2.35e+00 3.52e+07 4.54e-04 7.76e+07 2.23e+00 8.71e+07 2.14e-04 8.40e+07 1.92e+00 8.82e+07 1.01e-04 1.10e+08 1.65e+00 1.13e+08 3.73e-05 1.35e+08 1.35e+00 1.59e+08 1.07e-05 7.91e+07 1.00e+00 9.96e+07 5.04e-06 1.02e+08 8.21e-01 4.94e+07 1.86e-06 7.18e+07 7.43e-01 1.13e+0'8 8.76e-07 6.75e+07 6.08e-01 9.29e+07 4.14e-07 1.86e+08 4.97e-01 1.02e+08 1.00e-07 1.07e+09 3.69e-01 8.82e+07 Note: The energy values listed represent the upper boundary i
of each energy group.
t 1
12 7
1 i
)
SE-REA 074/94 3.0 NEUTRON DOSIMETRY The passive neutron sensors included in the Brunswick Unit 1 surveillance program are listed in Table 3-1.
Also given in Table 3-1 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the dosimetry and the subsequent determination of the various exposure parameters of interest (4(E > 1.0 MeV), c(E > 0.1 MeV),
and dpa)
The use of passive monitors such as those listed in Table 3-1 does not yield a direct measure of the energy dependent neutron flux at the point of interest.
Rather, the activation process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period.
An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.
In particular, the following variables are of interest:
- The measured specific activity of each monitor.
- The physical characteristics of each monitor.
- The operating history of the reactor.
- The energy response of each monitor.
- The neutron energy spectrum at the monitor location.
The irradiation history of the Brunswick Unit i reactor during the first 8 cycles of operation is summarized in Table 3-2.
The irradiation history data were obtained from NU2EG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable l
operating period.
The measured specific activity of each of the neutron monitors corrected to the end of irradiation was provided by Carolina Power and Light Companym, Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation (2436 j
MWt) were determined from the following equation.
4 13
?,
4 SE REA-074/94
^
R=
(Equation 3)
NFYTP!-
'/] [e " '#]
p< [1 -e o
where:
Reaction rate averaged over the irrad;ation period and R
=
referenced to operation at a core power level of P,,.,
(rps/ nucleus).
Measured specific activity (dps/g).
A
=
Number of target element atoms per gram of sensor.
No
=
Weight fraction of target isotope in the sensor F
=
material.
Number of product atoms produced per reaction.
Y
=
P, = Average core power level during irradiation period j (MW).
P,,, =
Maximum or reference power level of the reactor (MW).
X=
Decay constant of the product isotope (1/sec).
Length of irradiation period j (sec).
t,
=
Decay time following irradiation period j (sec).
to and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
i A summary of tne measured end-of-irradiation (EOI) and saturated reaction product specific activities as well as the derived full power average reaction rates are listed in Table 3-3.
Having the measured reaction rates presented in Table 3-3 and the calculated neutron energy spectrum at the surveillance capsule 14
SE-REA-074/94 j
center from Table 2-4, the derivation of the pertinent neutron exposure parameters proceeded as follows.
For each dosimeter, spectrum averaged reaction cross-sections were defined relative to a 1.0 MeV threshold as
[ o,(E) &GD dE o, =
(Equation 4) l &UD dE
1.0 where
The spectrum average cross-section for the ith a,
=
dosimetry reaction.
The multigroup dosimetry reaction cross-sections a,(E) for the ith reaction from the SAILOR library.
The calculated multigroup neutron energy spectrum
$(E) at the dosimetry location.
These computed spectrum averaged cross-sections were then used with the measured reaction rates to determine the average
&(E > 1.0 MeV) from the following relationship:
$(E > 1.0 MeV) = 1 f b (Equation 5) 0 tel Cl where:
R, The measured reaction rate for the ith dosimetry
=
reaction.
The spectrum averaged reaction cross-section for ai the ith dosimetry reaction.
15
SE-REA-074/94 The number of foil reactions at the sensor set n
location.
Having the deriv'ed fast neutron flux, c(E > 1.0 MeV), at the center of the surveillance capsule, exposure rates in terms of
?(E > 0.1 Mev) and dpa/sec were determined using (c(E > 1.0 MeV)]/(c(E > 1.0 MeV)] and (dpa/sec] / (c (E 1.0 MeV) ratios obtained from the R,0 transport calculation.
Appropriate spectrum averaged reaction cross-sections and exposure parameter ratios applicable to the center of the Brunswick Unit 1 surveillance capsule are summarized in Table 3-4.
Results of the dosimetry evaluation f or the surveillance capsule are provided in Tables 3-5 and 3-6.
In Table 3-5 the measured time averaged exposure rate and integrated exposure experienced by the capsule are listed.
The data summarized in this table include evaluations of the dosimetry in terms of c(E
> 1.0 MeV),
4(E > 0.1 MeV), and dpa.
From the data listed in Table 3-5, it is noted that the surveillance capsule experienced a fast neutron fluence (E > 1.0 MeV) of 3.19e+17 n/cm The irradiation period for the capsule was S.7 effective full power years (EFPY)
Comparisons of the measured exposures with calculations are given in Table 3-6.
The data listed in Table 3-6 indicates that, for these fast neutron evaluations, the ratio of calculation to measurement (C/M] was 0.876 for copper, 1.054 for iron, and 1.113 for nickel.
The average calculation / measurement C/M bias factor was 1.014.
These results show excellent agreement between calculation and measurement and are consistent with results observed from dosimetry evaluations performed for other light water reactors with large downcomer water gaps adjacent to the l
I 16
SE-REA-074/94 TABLE 3-1 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS REACTION TARGET OF WEIGHT PRODUCT
!NTEREST FRACTION HALF-LIFE Cu-63 (n,a) Co-60 0.6917 5.271 yrs Fe-54 (n,p) Mn-54 0.0580 312.5 dys Ni-58 (n,p) Co-58 0.6827 70.78 dys i
i 17 n
SE-REA-074/94 TABLE 3-2 MONTHLY THERMAL GENERATION DURING THE FIRST EIGHT FUEL CYCLES
- OF THE BRUNSWICK UNIT 1 REACTOR THERMAL THERMAL THERMAL MONTH
'MW-bri MONTH (MW-hr)
MONTH
( MW - h r )
10/76-01/77 0705085 08/79 1439400 03/82 1299464 02/77 0790801 09/79 1029076 04/82 1095485 03/77 0724440 10/79 1239429 05/82 1035130 04/77 1087214 11/79 0485147 06/82 0931084 05/77 0000000 12/79 0939301 07/82 0654785 06/77 0000000 01/80 1646831 08/82 0000000 07/77 0668587 02/80 1625394 09/82 0000000 08/77 0822268 03/80 1331441 10/82 0359034 09/77 1428392 04/80 0689066 11/82 1201155 10/77 1070158 05/80 0945202 12/82 0376734 11/77 0877738 06/80 0000000 01/83 0000000 12/77 1103356 07/80 0000000 02/83 0000000 01/78 1340275 08/80 0190222 03/83 0000000 02/78 0574440 09/80 1347267 04/83 0000000 03/78 1642798 10/80 1542606 05/83 0000000 04/7B 1408546 11/80 1549226 06/93 0000000 05/78 1223266 12/80 1426980 07/83 0000000 06/78 1559814 01/81 0960641 08/83 0091490 07/78 1628292 02/81 1209905 09/83 1516965 08/78 1682857 03/81' 1349616 10/83 0939394 09/78 1249091 04/81 0321413 11/83 0448431 10/78 1629167 05/81 0000000 12/83 1534996 11/78 1744524 06/81 0000000 01/84 1725331 12/78 1035903 07/81 0000000 02/84 1447183
[
01/79 0631597 08/81 0000000 03/84 1319848 02/79 0000000 09/81 0198157 04/84 1228147 l
03/79 0000000 10/81 1432095 05/84 1425709 l
04/79 0477257 11/81 1265964 06/84 1398699 05/79 1190845 12/81 1556279 07/84 1770758 06/79 1133625 01/82 1688996 08/84 1554862 07/79 1345626 02/82 0836342 09/84 1156675 18
SE-REA-074/94 TABLE 3-2 (continued)
MONTHLY THERMAL GENERATION DURING THE FIRST EIGHT FUEL CYCLES OF THE BRUNSWICK UNIT 1 REACTOR f
THERMAL THERMAL THERMAL MQNTF{
(MW-hr)
MONTH (MW-hr)
MONTH (MW-hr) 10/84 1570009 05/87 0000000 12/89 1800831 11/84 0000000 06/87 0540780 01/90 1781223 12/84 0953893 07/87 1511206 02/90 1604048 01/85 1437180 08/87 1798609 03/90 1798008 02/85 1071300 09/87 1744963 04/90 1704875 03/85 1013117 10/87 1774236 05/90 1171250 04/85 0000000 11/87 1707675 06/90 1004786 05/85 0000000 12/87 1767758 07/90 1719812 06/85 0000000 01/88 1287050 08/90 1645943 07/85 0000000 02/88 0363722 09/90 1298187 08/85 0000000 03/88 1787981 10/90 0000000 09/85 0000000 04/88 1669936 11/90 0000000 10/85 0000000 05/98 1597827 12/90 0000000 11/85 0797431 06/88 1681863 01/91 0000000 12/85 1738439 07/88 1165287 02/91 0039153 01/86 1772271 08/88 1614296 03/91 1332400 02/86 1618301 09/88 1459337 04/91 0000000 03/86 1414507 10/88 1258449 05/91 1441798 04/86 1230649 11/88 0418661 06/91 1737176 05/86 1790910 12/88 0000000 07/91 1337545 06/86 1739648 01/89 0000000 08/91 1806228 07/86 1787764 02/89 0000000 09/91 1263705 08/86 1243447 03/89 0000000 10/91 1421371 09/86 1490305 04/89 0723404 11/91 1740497 10/86 1793856 05/89 1797814 12/91 1807719 11/86 1443670 06/89 0574134 01/92 1619441 12/86 1610145 07/89 1772605 02/92 1641131 01/87 1490832 08/89 1802176 03/92 1458987 02/87 0601349 09/89 1340181 04/92 1177149 03/87 0000000 10/89 1802094 04/87 0000000 11/89 1572424 i
19
.p
.e%w.L-.
SE-REA-074/94 TABLE 3-3 MEASURED SENSOR ACTIVITIES AND DERIVED REACTION RATES EOI SATURATED REACTION ACTIVITY ACTIVITY RATE SENSOR (dos /cm)
(dps /ctm)
(ros/ atom)
Cu-63 (n.a) Co-60 65428 1.46e+04 2.77e+04 4.23e-18 65429 1.40e+04 2.66e+04 4.06e-18 AVEPAGE 1.43e+04 2.72e+04 4.14e-18 Fe-54 (n.0) Mn-54 65428 1.02e+05 1.42e+05 2.26e-16 65429 1.00e+05 1.39e+05 2.22e-16 AVEPAGE 1.01e+05 1.40e+05 2.24e-16 Ni-58 (n.n) Co-58 65428 1.61e+06 1.83e+06 2.60e-16 65429 1.57e+06 1.78e+06 2.54e-16 AVERAGE 1.59e+06 1.80e+06 2.57e-16 E
i TABLE 3-4 CALCULATED SPECTRUM AVERAGED CROSS-SECTIONS a Avg.
REACTION (barns / atom)
Cu-63 (n,a) Co-60 3.10e-03 Fe-54 (n,p) Mn-54 2.02e-01 Ni-58 (n,p) Co-58 2.44e-01
($(E > 0.1 MeV)]/[$(E > 1.0 MeV)]
1.67e+00 (dpa/sec] / ($ (E > 1.0 MeV)]
1.54e-21 20
SE-REA-074/94 TABLE 3-5
SUMMARY
OF DOSIMETRY RESULTS FOR THE SURVEILLANCE CAPSULE WITHDRAWN AT THE END OF FUEL CYCLE 8 EXPOSURE RATES AT CAPSULE CENTER REACTION RATE
&(E > 1.O MeV)
(ros/ nucleus)
(n/cm - sec)
Cu-63 (n,a) Co-60 4.14e-18 1.34e+09 Fe-54 (n,p) Mn-54 2.24e-16 1.11e+09 Ni-58 (n,p) Co-58 2.57e-16 1.05e+09 AVERAGE 1.17e+09 INTEGRATED EXPOSURE AT CENTER OF CAPSULE c(E > 1.0 MeV) 3.19e+17 n/cd j
2 4(E > 0.1 MeV) 5.33e+17 n/cm dpa 4.91e-04 dpa l
l TABLE 3-6 COMPARISON OF MEASURED AND CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE POSITION CALULATED MEASURED C/M Cu-63 ( n, cr) Co-60 3.63e-18 3.99e-18 0.876 Fe-54 (n,p) Mn-54 2.36e-16 2.16e-16 1.054 l
Ni-58 (n,p) Co-58 2.86e-16 2.48e-16 1.113 AVERAGE 1.014 BIAS FACTOR = 1/[C/M}
0.986 21 s
m.
i SE REA-074/94 1
4.0 PROJECTIONS OF PRESSURE VESSEL EXPOSURE I
Neutron exposure projections at several azimuthal locations on the pressure vessel inner radius are given in Tables 4-1 and 4-2.
Along with the current (end of Cycle 8 exposure), projections are also provided for an exposure period of 32 EFPY.
The data in Table 4-1 are applicable to the maximum exposure locations on the vessel upper shell above core midplane.
The data in Table 4-2 apply to the beltline circumferential weld located 2.2 feet below the midplane of the active reactor core.
In computing these vessel exposures, the calculated values from Table 2-1 were scaled by the average M/C ratios observed from the evaluations of the capsule dosimetry.
This procedure resulted in a bias factor of 0.986 being applied to the calculated values for all exposure parameters.
Projection for future operation were based on the assumption that the exposure rates characteristic cf the Cycle 8 core loading pattern would continue to be applicable throughout plant life.
22
SE-REA-074/94 j
TABLE 4-1 NEUTRON EXPOSURE PROJECTIONS AT THE INNER RADIUS OF THE REACTOR PRESSURE VESSEL UPPER SHELL 4(E > 1.0 MeV) t(E > 0.1 MeV)
LOCATION (n/cm )
(n/cm )
dna 2
EOC 8 NEUTRON EXPOSURE (8.7 EFPY)
VESSEL IR 0 Degrees 2.09e+17 3.60e+17 3.32e-04 15 Degrees 2.73e+17 4.68e+17 4.35e-04 30 Degrees 2.52e+17 4.70e+17 4.01e-04 45 Degrees 4.12e+17 7.12e+17 6.53e-04 16 EFPY NEUTRON EXPOSURE VESSEL IR 0 Degrees 3.86e+17 6.65e+17 6.14e-04 15 Degrees 4.91e+17 8.43e+17 7.82e-04 30 Degrees 4.42e+17 8.25e+17 7.03e-04 45 Degrees 7.18e+17 1.24e+18 1.14e-03 32 EFPY NEUTRON EXPOSURE VESSEL IR 0 Degrees 7.73e+17 1.33e+18 1.23e-03 15 Degrees 9.64e+17 1.65e+18 1.54e-03 l
30 Degrees 8.55e+17 1.60e+18 1.36e-03 45 Degrees 1.39e+18 2.40e+18 2.20e-03 i
l Note: These exposures represent the maximum levels experienced by the vessel upper shell.
23 l
SE REA-074/94 TABLE 4-2 NEUTRON EXPOSURE PROJECTIONS AT THE INNER RADIUS OF THE REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELD J
AND PRESSURE VESSEL LOWER SHELL l
4(E > 1.0 MeV) c(E > 0.1 MeV) 2 LOCATION (n/cm )
(n/cm )
dp3 EOC 8 NEUTRON EXPOSURE (8.7 EFPY)
VESSEL IR 0 Degrees 1.67e+17 2.88e+17 2.65e-04 15 Degrees 2.18e+17 3.74e+17 3.47e-04 30 Degrees 2.01e+17 3.75e+17 3.20e-04 45 Degrees 3.29e+17 5.69e+17 5.21e-04 26 EFPY NEUTRON EXPOSURE VESSEL IR 0 Degrees 3.08e+17 5.30e+17 4.89e-04 15 Degrees 3.91e+17 6.71e+17 6.23e-04 30 Degrees 3.53e+17 6.59e+17 5.62e-04 45 Degrees 5.73e+17 9.90e+17 9.08e-04 32 EFPY NEUTRON EXPOSURE VESSEL IR 0 Degrees 6.17e+17 1.06e+18 9.80e-04 15 Degrees 7.70e+17 1.32e+18 1.23e-03 30 Degrees 6.82e+17 1.27e+18 1.09e-03 l
45 Degrees 1.11e+18 1.92e+18 1.76e-03 Note: These exposures represent the maximum levels experienced by the vessel cercumferential weld and lower shell.
f-24 l
l l
l SE-REA-074/94 i
j
5.0 REFERENCES
1.
ORNL RSIC COMPUTER CODE COLLECTION CCC-543, TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport, Version 2.7.3, May 1993.
2.
ORNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P,
Cross-Section 3
Library for Light Water Reactors, July 1987.
3.
CP&L Letter BNT-575-CA-A529, " Evaluation of Results of Dosimetry Measurements and Associated Work for BNP Unit 1 Reactor Vessel Surveillance Program",
W.
R.
Campbell to Gary S.
Weingarten, February 18, 1994.
t i
a 25 i