ML20210P641
| ML20210P641 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 02/04/1987 |
| From: | Papanic G YANKEE ATOMIC ELECTRIC CO. |
| To: | Mckenna E NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| FYR-87-013, FYR-87-13, NUDOCS 8702130428 | |
| Download: ML20210P641 (12) | |
Text
..
e y 3 Tslephone (617) 8724100 1
TWX 710 38G7619 YANKEE ATOMIC ELECTRIC COMPANY 2.C2.1
-$p.
1671 Worcester Road, Framingham, Massachusetts 01701
. YANKEE
~
February 4, 1987 FYR 87-013 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attention:
Ms. Eileen M. McKenna Project Directorate No. 1 Division of PWR Licensing - A
References:
(a) License No. DPR-3 (Docket No. 50-29)
(b) Letter, Westinghouse Electric Corporation to YAEC,
" Pressurized Thermal Shock Rule " dated July 26, 1985 (c)
S. L. Anderson, " Analysis of Fast Neutron Flux Levels and End-of-Life Exposure for the Yankee Rowe Reactor Pressure Vessel," Westinghouse Electric Corporation, March 1981 (d) " Fracture Toughness Requirements for Older Plants,"
Standard Review Plan, Branch Technical Position MTEB 5-2 (e)
"NRC Staff Evaluation of Pressurized Thermal Shock,"
November 1982 (Enclosure A to SECY-82-465)
(f) Letter, YAEC to USNRC, dated January 22, 1986 (g) Letter, YAEC to USNRC, dated August 12, 1986 Transmitting Copy of Reference (c)
(h) Letter, USNRC to YAEC, dated August 26, 1986 (i) Letter, YAEC to USNRC, dated October 28, 1986 (j) NUREG/CR-1861, " LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind Test,"
July 1981
Subject:
Pressurized Thermal Shock Rule - 10CFR50.61(b)
Dear Ms. McKenna:
Yankee is required under 10CFR50.61(b) to establish the toughness state of the Yankee Nuclear Power Station reactor vessel materials for pressurized thermal shock. Accordingly, a submittal concerning this issue was sent to the NRC on January 22, 1986 (Reference (f)).
Subsequently, the NRC staff requested a copy of Reference (c) and asked several questions concerning the submittal (Reference (h)).
The requested 8702130428 870204 PDR ADOCK 05000029 P
7 United States Nuclear Regulatory Commission February 4, 1987 Attention:
Ms. Eileen M. McKenna Page 2 FYR 87-013 report and answers to the questions were transmitted in References (g) and (i).
The purpose of this letter is to transmit, per your request, a revised version of our original submittal, which provides some clarifications to the analysis and incorporates our answers to the questions in Reference (h).
This revision is provided in Attachment A.
The revised analysis shows that the limiting reactor vessel material, based on the RTPTS correlation, is the lower reactor vessel plate. The is 2290F currently and 2530F at 32 Effective calculated maximum RTPTS Full-Power Years (EFPY). These values are well below the PTS screening criterion of 2700 for plate material. Based on a conservative 80% capacity factor, the vessel will not reach 32 EFPY until 2002, and it will not reach j
the screening criterion of 2700F until 2020.
We trust that you will find this information satisfactory; however, should you have any questions regarding this matter, please contact us.
Very truly yours, YANKER ATOMIC ELECTRIC COMPANY l
G George Papanic, J Senior Project Engineer - Licensing Yankee Project GP/dhm Attachment l
l cc: USNRC Region 1 USNRC Resident Inspector, YNPS l
l
O ATTACEMENT A Per 10CFR50.61(b), a determination of RT as been made for Yankee Nuclear PTS Power Station from current operation through 32 Effective Full Power Years (EFPY). The determination requires that the material chemistry, vessel fluence, and initial'RT be established.
NDT Material Chemistry Table 1 summarizes the chemical composition of the reactor vessel plate and weld materials.
This data was obtained from tests of material samples. Based on this information, the following values are used for the weight percent of copper and nickel:
Cu (%)
Ni (%)
Upper Plate 0.18 0.18 Lower Plate 0.20 0.63 Weld 0.12 0.07 Vessel Fluence The vessel fluence determination is based on a Westinghouse analysis, which is described in Reference (c). Key features of this analysis include:
1.
use of a methodology that is based on the P scattering approximation y
and ENDF/B-II cross sections, 2.
use of a radial power distribution which is a time average of distributions calculated for Cores 12 through 15, and 3.
use of an axial power distribution peak to average factor of 1.2.
Each of these features is discussed briefly, before using the analysis results to determine fluence values for the PTS calculation. 5212R
- o Methodology The methodology used in Reference (c) has been benchmarked against results of NRC sponsored experiments. These experiments and the benchmark comparisons are documented in Reference (j). Table 2 provides a summary of pertinent results from that document. These results provide a 1
comparison of:
(1) experimental data, (2) calculations done by Westinghouse using the Reference (c) methodology, and (3) calculations done by two other organizations that employed a methodology which used the P scattering approximation and ENDF/B-IV cross sections.
In Table 2, 3
Data Points A4, A5, and A6 lie within the simulated pressure vessel wall.
At those locations, all of the calculations are biased low from 9% to 14%,
with all of the Westinghouse calculations agreeing with the measurements to within about 10%. These comparisons indicate that the results of the Westinghouse calculations using the Reference (c) methodology are approximately equivalent to the results of the calculations by the other organizations using the P 8PProximation and ENDF/B-IV cross sections.
3 However, it also indicates that results of calculations using the Reference (c) methodology must be multiplied by a factor of 1.1 in order to match the experiments. Accordingly, this correction factor will be applied when using the Reference (c) results for the PTS calculation.
o Radial Power Distribution In Reference (c), the radial power distribution was assumed to be a time average of distributions calculated for Cores 12 through 15.
Power density gradients for the peripheral assemblies were similarly determined using a time average of the pin power distributions for these assemblies.
The justification for this assumption is that the average radial power distribution over these four cycles is approximately equivalent to the average experienced and expected over the zero to 32 EFPY range of plant operation being considered.
Power distributions for cycles previous to Core 12 were such that peripheral assembly powers were 2% to 8% lower than for Cores 12 though 15.
Power distributions for cycles subsequent to Core 15 have thus far been such that peripheral assembly powers are 3% to '
5212R
-5% higher than for Cores 12 through 15; and this ratio is expected to remain approximately constant in future cycles.
o Axial Power Distribution Use of an axial peak to average factor of 1.2 is considered to be conservative relative to the average value of this factor experienced to date or expected out to 32 EFPY of plant operation.
During Core 1, the plant was operated as a rodded core and thus the axial factor may have been greater than 1.2.
However, the power level was initially only 392 MWt and later only 485 MWt during that cycle. The plant was not operated at the present power level of 600 MWt until Cycle 2.
Also, the total of Core 1 operation was less than one EFPY and represents only a small fraction of the total accumulated fluence. Since Core 2, the plant has been operated with soluble boron as the main reactivity control. Thus, control rod insertion has been used only for fine reactivity control and the core has operated in a virtually unrodded condition., For these reasons, the average axial factor for Core 2 and all subsequent cores has been below 1.2.
Also, it should remain below this value for all future Cores.
o Fluence Calculations According to the analysis provided in Reference (c), the peak fluence accumulation rate at the inside surface of the upper plate near the core 19 2
midplane is 0.101 x 10 W per EFPY. The lower plate receives less than 89% of this peak value, because of its location relative to the core centerline. Therefore, if a methodology correction factor of 1.1 is applied in accordance with the benchmark comparisons discussed previously, the following values are obtained for the peak fluence accumulation rate at the inside surface of the upper and lower plates:
I9 2
l Upper Plate (1.1)(0.101x10 n/cm per EFPY) = 0.111x10 ' n/cm per EFPY or Weld 19 2
Lower Plate (0.89)(1.1)(0.101x10 n/cm per EFPY) = 0.099x10 n/cm per EFPY g 5212R
The total plant generation through 1986 is 97,581,945 MWHt or 18.57 EFPY.
Using'the above fluence accumulation rates and the actual value of EFPY gives the following values for the current peak fluence at the inside vessel surface:
19 19 Upper Plate (18.57 EFPY)(0.111x10 n/cm per EFPY) = 2.061x10 n/cm or Weld i
Lower Plate (18.57)(0.099x10 ' n/cm per EFPY) = 1.838x10 n/cm 1
19 2
For an arbitrary value of 32 EFPY, and using the above fluence 1ccumulation rates, the values of the peak fluence at the inside vessel surface would be:
19 2
19 2
Upper Plate (32 EFPY)(0.111x10 n/cm per EFPY) = 3.555 x 10 n/cm or Weld Lower Plate (32 EFPY)(0.099x10 ' n/cm2 19 2
per EFPY) = 3.168 x 10 n/cm Initial RTNDT Charpy tests were conducted by the manufacturer for both vessel plate and weld material samples. All tests were performed at 10 F using longitudinal e
specimens with none resulting in less than 30 ft-lbs.
Initial RT f r the NDT plate material was corrected per Reference (d) by adding 20 F to compensate for the lack of transverse-oriented specimens. The initial RT f r the NDT weld remains 10 F since these specimens are isotropic. A summary of RTNDT and Charpy test values is provided in Table 3.
Figure 1 shows the weld locations in relation to the reactor core.
RTPTS Values RT is calculated as prescribed in 10CFR50.61(b). RT as defined in PTS PTS l
10CFR50.61(b)(2) is the lower of the results given by Equations 1 and 2.
In i
all cases, Equation 1 resulted in lower values of RT values PTS
{
are contained in Table 4.
I 4
i 5212R 1
I
. _. _ _ _. _ _ - -, _. ~... _ - -
r The limiting reactor vessel material, based on the RTPTS " ##* *
" 8 lower vessel plate. This is due to its higher nickel content and resulting increased irradiation embrittlement. This is also noted in Reference (e).
The calculated maximum RT in the limiting plate is 229 F currently and PTS 253 F at 32 EFPY. These values are well below the PTS screening criterion of 270 F for plate material. Therefore, the Yankee reactor vessel can be operated for over 32 EFPY without significant risk resulting from pressurized thermal shock.
Based on an 80% capacity factor, the vessel will not reach 32 EFPY until 2002, and it will not reach the screening criterion of 270 F until 2020. These dates are obviously conservative, considering the fact that the cumulative, average capacity factor for the last 26 years is less than 74%. 5212R
n l.
TABLE 1 Reactor Vessel Plate and Weld Material ~ Properties Upper Plate Lower Plate
- Supplier: Lukens Supplier: Lukens Weld Chemical Composition (%)
Heat No.:
19281-2 Heat No.:
19244-3 Material C'
.20
.19-Mn 1.27 1.18
.68 P
.020
.016
.01 S
.028
.026 Si
.21
.20
.20 Mo
.48
.48
.02 Ni
.18
.63
.07 Cu
.18
.20
.12 Cr
.06 V
A1 S
5212R
y =.
4 r.i v.
TABLE 2 Results of Benchmark Comparisons of Reference-(c) Methodolony
{
Flux (E >1.0 MeV)
Data Calculations 1,ocation Measured Westinghouse A
Z
'A0 1.54 x 10 '
1.68 x 10-0 1.66 x 10
-6
-6 Al 3.71 x 10 3.68 x 10 3.39 x 10-6 3.55 x 10-6
-I A2 3.88 x 10 3.89 x 10-7 A3 1.33 x 10 1.29 x 10-7 1.21 x 10-7 1.21 x 10-7 A4 4.30 x 10-8 3.94 x 10 3.96 x 10-8 3.99 x 10-0
-8
-8
-0 AS 2.07 x 10 1.89 x 10 1.84 x 10-0 1.82 x 10-8
-9
-9 A6 9.11 x 10 8.33 x 10 7.88 x 10-9 7.87 x 10-9 l
Notes:
1.
Results are taken from Sections 6.3 and 7.1 of Reference (j).
2.
Data is for the 12/13 PCA configuration.
. 5212R
y 4
TABLE 3 Initial RT and Charpy Test Values NDT Upper Plate Lower Plate Heat No.
19281-2 19244-3 Charpy Test at +100F (ft-lbs)1 L52-50-46 L48-41.5-43 Initial RTNDT (Per MTEB 5-2)
+300F
+300F Weld To Weld To Upper Plate Lower Plate Heat No.
19281-2 19244-3 J
Charpy Test at +100F (ft-lbs)1,2 40-31-35 42-30-38 Initial RTNDT
+100F
+100F 1.
Preirradiation tests by Babcock and Wilcox at time of vessel fabrication.
Tests conducted at 1/4 thickness location.
2.
Weld material is Hi-Mang Moly wire with Linde No. 80 48XD flux.
, 5212R r
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TABLE 4 i
Fluence Initial
{
Material n/cm2 (E>1.0 MeV1 RTNDT (I)
RTPTS Upper Plate (Current) 2.061 x 10 "
30 F 182 F f"
Upper Plate (32 EFPY) 3.555 x 10 "
30 F
'199 F r
19 l
Lower Plate (Current) 1.838 x 10 30 F 229 F I9 Lower Plate (32 EFPY) 3.168 x 10 30 F 253 F l
19 Weld (current) 2.061 x 10 10 F 118 F 19 l
Weld (32 EFPY) 3.555 x 10 10 F 127 F l
l r
l Notet RT values are calculated using Equation 1 l-PTS 0.270 RTPTS = I + M + (-10 + 470 Cu + 350 CuN1) f l
wherer I = measured initial reference temperature in 0F l
M = 480F f
. 5212R l
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m-FIGURE 1 YANKEE REACTOR VESSEL WELD LOCATIONS 8
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