ML20072F250
| ML20072F250 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/16/1994 |
| From: | Grant S CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20072F244 | List: |
| References | |
| SR-BNP1-1005, SR-BNP1-1005-00, SR-BNP1-1005-001, SR-BNP1-1005-1, NUDOCS 9408230260 | |
| Download: ML20072F250 (9) | |
Text
CAROLINA POWER & LIGIIT COMPANY NUCLEAR ENGINEERING DEPARTMENT
SUMMARY
REPORT SR-BNPI-1005-001 BRUNSWICK UNIT 1 RPV SURVEILLANCE PROGRAM FIRST CAPSULE (300 ) REMOVAL AFTER 8 FUEL CYC.LES TEST IESULTS AND PROJECTIONS ORIGINATOR /DATE h
8 4\\94 REVIEW /DATE I2/M[
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APPROVAL /DATE M
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CONTENTS TITLE PAGE NO.
KEY WORDS 1
SUMNIARY AND CONCLUSION 1
INTRODUCTION.
1 CONIPARISON OF MEASUREMENTS TO PRIOR REPORTS AND PROJECTIONS 3
Fast Neutron Flux and Fluence 3
Chemistry 3
Tensile Tests.
4 Charpy V Tests 4
EFFECTS OF LOW TEMPER ATURE OPERATION.
5 REM AINING SURVEILLANCE PROGR AM....
5 REFERENCES..
5 Supplement 1 Brunswick Unit 1 RPV Surveillance Materials Testing Supplement 2 Neutron Dosimetry Results for the Surveillance Capsule Removed at the Conclusion of Fuel Cycle 8 Supplement 3 Information on Reactor Vessel Material Surveillanc. Program. NEDO-24161. Rev.1
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KEY WORDS Fast Neutron Flux and Fluence Radiation Surveillance Program Fracture Toughness 2
t SUMNIARY AND CONCLUSION i
The test results from the first reactor vessel surveillance capsule to be removed from Brunswick Unit 1 (300" capsule) demonstrate that the current heatup - cooldown curves are conservative; therefore, revision is not required.
INTRODUCTION The Brunswick (BSEP) 1 Surveillance Program for the reactor pressure vessel materials was designed for the requirements of ASTM E 185-66, except that the number of unirradiated tensile specimens for plate weld and heat-affected-zone was 2 for each rather than the specified 3 and the number of unirradiated Charpies was 3 rather than the specified 15.
Test methods and reporting for the 300 degree capsule, the first to be removed in the program, comply with the ASTM E 185-82 edition.
The results of mechanical properties, chemical analysis, and radiochemical testing of the capsule's contents are provided in Supplement 1 of this report. The tests were performed by General Electric at the Vallecitos Nuclear Center.
Westinghouse Electric used the flux wire activity measurements from Supplement I and power and void distributions as varied over time, as supplied by CP&L, for input to neutron transport calculations which produced the fast neutron exposure rates at the capsule and at key maximum i
locations on the vessel in the core region. The Westinghouse work is included as Supplement 2.
l Most of the surveillance program description and core region materials characterization are given in Supplement 3, "Information on Reactor Vessel Surveillance Program." Other information required by ASTM E 185-82 is provided in Supplement 1 or 2. Salient data for core region materials have been collected in Table 1 for ready reference. (4i93x94.nnr>
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TABLEI BRUNSWICK 1 EOL: 9/8/2016 CilEM.
EOL ID EOL 1/4T INITIAL RTm BELTLINE IIEAT NO.
FACTOR FLUENCE FLUENCE (Note 1)
ART,, _
y IDENT.
PLATE NO.
%CU
%NI (TABLE)
(27 EFPY)
(27 EFPY)
(*F)
I/4T Lower Shell C4535-2 0.12 0.58 82.6 9.4 E17 6.9 E17 10 29 201 Lower Shell C4550-1 0.11 0.60 74 9.4 E17 6.9 E17 10 26 251 Lower Int. Shell C4487-:
0.12 0.56 82.2 1.2 El8 8.6 E17 10 **
- 32 301 l
Lower Int. Shell B8496-1 0.19 0.58 139.8 1.2 E18 8.6 E17 10 54 351 Nonle N16A Q2Q1VW
-0.16 0.82 123 4.6 E17 3.4 El7 0 ** *)
29 Nonle N16B Q2QlVW 0.16 0.82 123 4.6 E17 3.4 E17 0 ** *'
29 Axial Welds S3986 0.05 0.96 68 6.5 E17 4.7 E17
-56 **
- 19 G1 & G2 (Note 2)
Axial Welds S3986 0.05
-0.96 68 8.2 E17 5.9 E17
-56 **
- 22 F1 & F2 (Note 2)
Cire. Welds IP4218 0.06 0.87 82 9.4 E17 6.9 E17
-56 **
- 28 NOTES:
I.
Values based on NDTT, unless otherwise specified.
2.
Represented by surveillance weld.
3.
CP&L testing of archive plate used to establish RTm based on Subsection NB-2331 of ASME Code Section III.
4.
Based on additional Lenape Forge drop weight test data for lleat Q2Q1VW (not included in Section 3.6 of NEDO-24161, Revision 1).
5.
Generic value. (4:93X94.BNP)
COMPARISON OF MEASUREMENTS TO PRIOR REPORTS AND PROJECTIONS Fast Neutron Flux and Fluence 5
2 The capsule fast neutron flux was determined to be 1.18 x 10' n/cm. - sec > 1 MeV (see Table 2-1 of Supplement 2). It had been measured at 1.4 x 10' after the first fuel cycle using dosimetry extractable from the capsule in situ. This higher initial flux was to be expected due to the low leakage feature of utilized core loading patterns. At 8.67 EFPY, the fluence for the capsule is 3.2 x 10" n/cm2> 1 MeV.
Whereas the higher flux (1.4 x 10') was used in the development of current heatup - cooldown curves, those curves remain conservative. A small additional conservatism was found in performance of S_upplement 2 calculations: the lead factor from the capsule to the maximum 1/4T position was found to be 1.04 compared to the previously assumed lead factor of 1.0. The lead factor from the capsule to the maximum ID azimuth of 45' is 0.80.
Chemistry Check chemical analyses were performed for the surveillance plate and the surveillance weld (see Table 6-1 of Supplement 1). The heat number and elemental chemistry of the surveillance weld had previously been unknown; however, General Electric recently was able to identify the weld heat as Adcom S3986 and the flux as Linde 124,' Lot 3876, Run No. 934. The chemistry and mechanical properties of the reactor vessel beltline fabricated materials are given in Section 3.0 of Supplement 3.
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Tensile Tests l
Room temperature test results for the surveillance plate and weld, unirradiated (see Section 3.5 of Supplement 3) and irradiated, as taken from Table 5-1 in Supplement 1, were:
i Yield Strength ksi Ultimate Strength ksi Total Elongation %
Unirradiated Plate 65 89 25 Irradiated Plate 69 91 20.4 Unirradiated Weld 71.8 86.5 30 Irradiated Weld 76 92 21.6 As demonstrated by the test results, the weld (vs. base metal) has the greater radiation embrittlement despite a lower copper content (.055 versus.099%). Higher phosphorus (.017 versus.009%) or higher nickel (.98 versus.53%) in the weld may account for the difference.
Charpy V Tests 1
As demonstrated below, the adjusted reference temperatures (ART), as given in Reference 1. are conservative compared to that obtained with the results of this capsule test at 8.7 EFPY.
l Measured ART 300* Capsule Calculated ART 8.67 EFPY ** "
8.67 EFPY ** '
Plate 301 30.6 63 Weld
-4.4 16 Plate 351 **
76 NOTES:
1.
Reference Table 4-3 of Supplement 1.
2.
Per Reference 2.
3.
Controlling Material (material specimens not included in surveillance capsule).
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The shelf energy after 8.7 EFPY is found in Table 4-3 of Supplement I for both surveillance plate and weld to be more than adequate. As converted to the transverse direction, the shelf energy for the plate is 96.6 ft. Ib. The shelf energy for the weld is 97.5 ft. Ib.
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EFFECTS OF LOW TEMPER ATURE OPER ATION i
The changes in mechanical properties with irradiation exposure to the first 8.7 EFPY are sufficiently small to indicate that operation below 525'F during that period is not a reason for concern. Future operation will entail much less low temperature operation, and the issue should be closed with respect to Reference 3.
REM AINING SURVEll. LANCE PROGR AM Brunswick I has two capsules remaining in place in the reactor at 30* and 120 azimuths. Removal dates have not yet been selected, because it is desirable to establish an integrated surveillance program for Units 1 and 2 in the future. The first Brunswick 2 capsule is scheduled for removal in 1996. It is importar.t to have the test results from that capsule in order to set the parameters for an integrated program. At that time, the remaining surveillance schedule can be established.
R EFERENCES i
1.
Amendment 172 to Brunswick Facility Operating License. February 15,1990. (Note: As noted to the USNRC in LER l-94-005, Amendment 172 was assigned to Brunswick Unit 2 at i
l that time incorrectly.)
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2.
USNRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials.
3.
Letter, R. B. Starkey, Jr. (CP&L) to USNRC Document Control Desk, NLS-92-180, j
Response to Generic Letter 92-01, Revision 1, July 6,1992.
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BRUNSWICK UNIT 1 RPV SURVEILLANCE MATERIALS TESTING (38 PAGES)
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